Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices 25761

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 April 8,

2000, through April 21, 2000. The last

biweekly notice was published on April

19, 2000 (65 FR 21034).

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 2, 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

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

25762 Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices

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).

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: April 7,

2000.

Description of amendment request:

The proposed amendment would revise

Harris Nuclear Plant (HNP) Technical

Specification (TS) 3/4.7.6, ‘‘Control

Room Emergency Filtration System,’’ TS

3/4.7.7, ‘‘Reactor Auxiliary Building

Emergency Exhaust System,’’ TS 3/

4.9.12, ‘‘Fuel Handling Building

Emergency Exhaust System,’’ and the

associated Bases. Specifically, the

licensee proposes to revise these TS to

provide an Action when the Control

Room Emergency Filtration System or

Reactor Auxiliary Building Emergency

Exhaust System ventilation boundary is

inoperable and a note that allows an

applicable ventilation boundary to be

open intermittently under

administrative controls. Additionally,

the licensee proposes to modify TS 3/

4.3.3.1, ‘‘Radiation Monitoring for Plant

Operations,’’ to provide consistency

between the applicability of the Control

Room Emergency Filtration System and

the radiation monitors that initiate a

Control Room Isolation signal.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Ventilation systems are not accident

initiating systems as described in the Final

Safety Analysis Report. The changes are

based on the low probability of a design basis

accident occurring during the 24 hour

completion time and compensatory measures

available to minimize dose consequences of

an event during this time. The proposed

change does not affect another Structure,

System, or Component.

Current HNP TS do not restrict fuel

movement in the fuel handling or loads over

spent fuel pools concurrent with an

inoperable Control Room Emergency

Filtration System. Providing restrictions for

fuel movement and loads over spent fuel

pools preserves assumptions made in the fuel

handling accident analysis. The addition of

applicability requirements for fuel movement

and movement of loads over spent fuel pools

is consistent with NUREG–1431, Revision 1,

and is more restrictive than current HNP TS.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Ventilation systems are not accident

initiating systems as described in the Final

Safety Analysis Report. As such, the failure

of the ventilation system to operate properly

or a premature actuation of the ventilation

system can not initiate an accident.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety.

The proposed change to ventilation

systems does not significantly affect any of

the parameters that relate to the margin of

safety as described in the Bases of the TS or

the FSAR [Final Safety Analysis Report].

Accordingly, NRC Acceptance Limits are not

affected by this change. The changes are

based on the low probability of a design basis

accident occurring during the 24 hour

completion time and compensatory measures

available to minimize dose consequences of

an event during this time.

The addition of applicability requirements

for Control Room Emergency Filtration

System during movement of irradiated fuel

assemblies and movement loads over spent

fuel pools provide additional margin not

currently provided in HNP TS.

Therefore, the proposed change does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: William D.

Johnson, Vice President and Corporate

Secretary, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Section Chief: Richard P.

Correia.

Carolina Power & Light Company, et

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

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: April 12,

2000.

Description of amendment request:

The proposed amendment would revise

Harris Nuclear Plant (HNP) Technical

Specification (TS) 3/4.4.9.2, ‘‘Pressure/

Temperature (P-T) Limits—Reactor

Coolant System,’’ and TS 3/4.4.9.4,

‘‘Overpressure Protection System,’’ and

the associated Bases. Specifically, the

licensee proposes to revise the

applicable TS to incorporate results of

the Reactor Vessel Surveillance Program

capsule analysis. A summary report was

previously submitted to the NRC (HNP–

99–157, dated 11/9/99) in accordance

Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices 25763

with Title 10 of the Code of Federal

Regulations, Part 50 (10 CFR 50),

Appendix H. Additionally, the

licensee’s submittal requested an

exemption to 10 CFR 50.60 (a), based on

American Society of Mechanical

Engineers (ASME) Code Case N–640 and

WCAP–15315. The exemption request

will be evaluated separate from the

proposed license amendment.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes affect operations of

the Reactor Coolant System (RCS)

components when the RCS temperature is

below 350° F. The revisions to P–T limits and

allowable heatup and cooldown rate limits

are consistent with ASME code cases which

have been authorized for other licensees by

the NRC. The proposed changes modify the

setpoint of the pressurizer PORVS [power

operated relief valves] for LTOPS [low

temperature overpressure setpoints]. Changes

to the LTOPS setpoints applicable below

350° F effectively increase the allowable

operating pressure for any given temperature

during shutdown. These changes do not

result in conditions which are outside of the

design basis for RCS Structures, Systems, and

Components (SSCs). Therefore, the proposed

changes do not alter the characteristics of the

RCS SSCs adversely, and therefore do not

impact the performance of the RCS SSCs

during power operations.

The revised P–T limits and heatup and

cooldown rate limits are within the design

capabilities of the RCS SSCs and pressure

control systems. While the proposed new P–

T limits are less restrictive than the current

Technical [Specification] requirements, they

assure that plant operation is within the

design capacity of the reactor vessel

materials. Therefore, the RCS capability as a

fission product barrier is not compromised.

The changes to the LTOPS setpoints do not

affect accident consequences since no credit

is assumed for operation of LTOPS to

mitigate accidents.

Therefore, the proposed amendment does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed changes do not involve new

plant components or procedures, but only

revise existing operational limits and

setpoints. These changes do not place SSCs

in conditions outside of their design basis,

and the revised operating setpoints and

conditions are within the capability of the

plant control systems.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety.

The proposed changes to the P–T limits

and LTOPS setpoints change the

calculational method from that described in

the bases to one based on ASME Code Case

N–640, and on WCAP–15315. The effect of

this change is to allow plant operation with

different limits, but still with adequate

margins to assure the integrity of the reactor

vessel and RCS.

Therefore, the proposed change does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: William D.

Johnson, Vice President and Corporate

Secretary, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Section Chief: Richard P.

Correia.

Commonwealth Edison Company,

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1

and 2, Will County, Illinois

Date of amendment request: March

15, 2000.

Description of amendment request:

The proposed amendments would

revise the ultimate heat sink

temperature in the technical

specifications from 98°F to 100°F.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of any accident previously evaluated?

Analyzed accidents are assumed to be

initiated by the failure of plant structures,

systems or components. An inoperable

Ultimate Heat Sink (UHS), which is the

source of water for the Essential Service

Water (ESW) System, is not considered as an

initiator of any analyzed events. The analyses

for Braidwood Station, Units 1 and 2, assume

a UHS temperature of 100°F. Therefore,

continued operation with a UHS temperature

less than or equal to 100°F will not increase

the probability of occurrence of any accident

previously evaluated in the Updated Final

Safety Analysis Report (UFSAR). The

proposed change does not involve any

physical alteration of plant systems,

structures or components. A UHS

temperature of up to 100°F does not increase

the failure rate of systems, structures or

components because the systems, structures

or components are rated and analyzed for

operation with ESW temperatures of 100°F

and the design allows for higher

temperatures than at which they presently

operate.

The basis provided in Regulatory Guide

1.27 ‘‘Ultimate Heat Sink for Nuclear Power

Plants,’’ Revision 2, dated January 1976, was

employed for the temperature analysis of the

Braidwood Station UHS to implement

General Design Criteria (GDC) 44, ‘‘Cooling

water,’’ and GDC 2, ‘‘Design bases for

protection against natural phenomena,’’ of

Appendix A to 10 CFR Part 50. This

Regulatory Guide was employed for both the

original design/licensing basis of the

Braidwood Station UHS and a subsequent

evaluation which investigated the potential

for increasing the average water temperature

of the UHS from ≤98°F to ≤100°F. The heat

loads selected for the UHS analysis

considered one Braidwood Station unit in a

Loss of Coolant Accident (LOCA) condition

concurrent with a Loss Of Offsite Power

(LOOP) event and the remaining Braidwood

Station unit undergoing a safe non-accident

shutdown. In the analysis, these heat loads

are removed by the UHS using only ESW

pumps. The main cooling pond is

conservatively assumed not to be available at

the start of the event. The analysis shows that

with an initial UHS temperature of 100°F, the

required heat loads can be met for 30 days

while maintaining ESW temperatures at

acceptable values.

Based on the above, it has been

demonstrated that the operation at an initial

UHS temperature of TM100° F at the start of

the design basis event will result in the

continued ability of the equipment and

components supplied by the ESW system to

perform their intended safety functions.

Therefore, increasing the average water

temperature limit of the UHS from TM98° F to

TM100° F does not increase the consequences

of any accident previously evaluated. Raising

this limit does not introduce any new

equipment, equipment modifications, or any

new or different modes of plant operation,

nor does it affect the operational

characteristics of any equipment or systems.

Therefore, this proposed change does not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change does not involve a

physical alteration of the units. There is no

change being made to the parameters within

which the units are operated that is not

bounded by the analyses. There are no

setpoints at which protective or mitigative

actions are initiated that are affected by this

proposed change. This proposed change will

not alter the manner in which equipment

operation is initiated, nor will the function

demands on credited equipment be changed.

No alteration in the procedures that ensure

the units remain within analyzed limits is

proposed, and no change is being made to

25764 Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices

procedures relied upon to respond to an offnormal event. As such, no new failure modes

are being introduced. The proposed change

does not alter assumptions made in the safety

analysis.

Increasing the allowed average water

temperature of the UHS in Technical

Specification (TS) 3.7.9, ‘‘Ultimate Heat Sink

(UHS),’’ has no impact on plant operation.

Operating at the proposed higher temperature

limit does not introduce new failure

mechanisms for systems, structures or

components. The engineering analyses

performed to support the change to UHS

temperature limit provides the basis to

conclude that the equipment is designed for

operation at elevated temperatures. The

current analyses and calculations assume a

UHS temperature of 100° F, which is within

the design limits of the affected equipment.

In addition, design and construction codes

applied to the affected structures, systems

and components provided sufficient margin

to accommodate the proposed temperature

change.

Therefore, this proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The proposed change allows operation

with the UHS temperature TM100° F. The

margin of safety is determined by the design

and qualification of the plant equipment, the

operation of the plant within analyzed limits,

and the point at which protective or

mitigative actions are initiated. The proposed

change does not impact these factors. The

existing analyses already assume an initial

UHS temperature of 100° F for design basis

accident conditions. There are no required

design changes or equipment performance

parameter changes associated with this

change. No protection setpoints are affected

as a result of this change. This temperature

increase has been confirmed to not change

the operational characteristics of the design

of any equipment or system. All accident

analysis assumptions and conditions will

continue to be met. Thus, the proposed

increase in UHS temperature does not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Ms. Pamela B.

Stroebel, Senior Vice President and

General Counsel, Commonwealth

Edison Company, P.O. Box 767,

Chicago, Illinois 60690≤980767.

NRC Section Chief: Anthony J.

Mendiola.

Detroit Edison Company, Docket No.

50–16, Enrico Fermi Atomic Power

Plant, Unit 1, Monroe County, Michigan

Date of amendment request: February

11, 1999 (Reference NRC–00–0023).

Description of amendment request:

The proposed amendment will revise

the Technical Specifications by: (1)

Deleting Specification A.8, the

definition of ‘‘Primary System’’ which

will no longer be necessary if the

specifications related to the Primary

System cover gas system are deleted; (2)

deleting Specification D, which

specifies the requirements for the

Primary System cover gas system; (3)

deleting the portion of Specification H.1

that specifies the surveillance

requirements for the Primary System

pressure alarms; (4) deleting Table H.1

item a, the Primary System pressure

alarm points; (5) deleting Specification

H.3.b, the requirement to perform

surveillances of the door and seals

around the machinery dome; (6)

deleting Specification I.7.b, which

requires procedures for maintaining

cover gas supply; and (7) deleting

Specification I.9.d, which requires

keeping records of CO2 cover gas usage.

The above-listed changes would allow

the licensee to remove the Primary

System cover gas system from service,

an action that would allow the licensee

to begin work on removing the

remaining residual sodium from the

Primary System. The licensee also

requested an editorial change in Table

H.1 item b.1, to change ‘‘Bldg.’’ to

‘‘Building’’.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration using the standards in 10

CFR 50.92(c). The licensee’s analysis is

presented below:

(1) The proposed change does not involve

a significant increase in the probability or

consequences of an accident.

Removing the cover gas from the [P]rimary

[S]ystem, opening the [P]rimary [S]ystem,

and cleaning out sodium residues will not

significantly increase the probability of an

accident occurring as long as the probability

of an uncontrolled water reaction with the

sodium is not significantly increased. This is

done by conducting cutting operations and

sodium reactions under control conditions.

Removing the cover gas or opening the

system will not take place until the current

asbestos abatement project in the Reactor

Building is complete since water is being

used. The abatement is expected to be

completed this winter before the license

amendment will be approved. Note that EPA

approval for dry removal has been obtained

for where there is a risk of water coming into

contact with sodium. The successful

dismantling of the secondary sodium system

piping in the Steam Generator building

demonstrates that sodium systems can be cut

open safely. The sodium residue processing

in the secondary sodium storage tanks

demonstrates sodium cleanup can be

conducted safely. The consequences of an

accident will not be increased by removing

the cover gas, opening the [P]rimary

[S]ystem, or reacting the sodium residues

because the previously analyzed accidents

already involve the release of all the

radioactive material in the [P]rimary [S]ystem

and all the radioactive material in the liquid

waste system. The maximum postulated dose

to the public was analyzed to be within the

10 CFR [Part] 20 limit of 100 mrem/year. This

change will not increase the amount of

radioactive material available to be released.

(2) The proposed change does not create

the possibility of a new or different accident

from any previously evaluated.

Removing the cover gas from the [P]rimary

[S]ystem, opening the [P]rimary [S]ystem,

and cleaning out the sodium residues will

not create a new or different type of accident.

A sodium accident has been previously

evaluated. The only other type of accident

which could possibly be caused by removing

the [P]rimary [S]ystem cover gas, opening the

[P]rimary [S]ystem, or processing primary

sodium residues is a liquid waste release,

which is highly unlikely. A liquid waste

accident has also been previously evaluated.

Only the [P]rimary [S]ystem and other

equipment or piping containing primary

sodium is expected to be affected by this

change.

(3) The proposed change does not involve

a significant reduction in a margin of safety.

Only a relatively small amount of sodium

remains in the [P]rimary [S]ystem and other

equipment containing primary sodium. Some

of this residual may have been converted to

sodium carbonate, leaving even less sodium

remaining. The cover gas was a good

precaution, especially for systems sitting

unattended for many years. It prevented

moisture from intruding into the systems and

reacting with the sodium residues. It

prevented oxygen from entering and reacting

with any hydrogen formed from reactions of

water and sodium. Discontinuing the use of

cover gas slightly reduces the margin of

safety, but not significantly. Removing the

cover gas does not, in itself, introduce water

into the system in an uncontrolled manner.

Even if slight amounts of moisture from

humidity in the air enter over the next year

or two until the sodium is removed while the

system is opened or unsealed, the system

volume is large enough that the system will

be able to dissipate any small reactions that

occur. In addition, the calculated

consequence[s] of releasing the radioactive

material in the primary sodium is small and

well within 10 CFR [Part] 20 and Technical

Specification limits.

The planned processing of sodium residues

is evaluated as releasing the radioactive

material to the atmosphere, as planned

release using controls specified in the

Technical Specifications for gaseous

effluents. For these reasons, the proposed

change does not involve a significant

reduction in the margin of safety.

Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices 25765

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,

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: John Flynn,

Esquire, Detroit Edison Company, 2000

Second Avenue, Detroit, Michigan

48226.

NRC Branch Chief: Larry W. Camper.

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant, Units 3 and 4, Dade

County, Florida

Date of amendment request:

November 30, 1999, as supplemented

March 8, 2000.

Description of amendment request:

The proposed amendments would

revise the Technical Specifications to

allow the use of credit for soluble boron

in the spent fuel pool criticality

analyses. In addition, a revised

criticality analysis for the fresh fuel

storage racks will be used to update the

licensing bases. Criticality analyses

were performed using the methodology

developed by the Westinghouse Owners

Group and described in WCAP–14416–

NP–A, Revision 1, Westinghouse Spent

Fuel Rack Criticality Analysis

Methodology.

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.

There is no increase in the probability of

a fuel assembly drop accident in the Spent

Fuel Pool (SFP) when considering the

presence of soluble boron in the SFP water

for criticality control. The handling of the

fuel assemblies in the SFP has always been

performed in borated water. The

consequences of a fuel assembly drop

accident in the SFP are not affected when

considering the presence of soluble boron.

There is no increase in the probability of

the accidental misloading of spent fuel

assemblies into the SFP racks when

considering the presence of soluble boron in

the pool water for criticality control. Fuel

assembly placement will continue to be

controlled pursuant to approved fuel

handling procedures and will be in

accordance with the Technical Specification

(TS) spent fuel rack storage limitations. There

is no increase in the consequences of the

accidental misloading of spent fuel

assemblies into the SFP racks because

criticality analyses demonstrate that the pool

will remain subcritical following an

accidental misloading if the pool contains an

adequate boron concentration. The proposed

TS ensure that an adequate SFP boron

concentration will be maintained. There is no

increase in the probability of the loss of

normal cooling to the SFP water when

considering the presence of soluble boron in

the pool water for subcriticality control since

a high concentration of soluble boron has

always been maintained in the SFP water.

A loss of normal cooling to the SFP water

causes an increase in the temperature of the

water passing through the stored fuel

assemblies. This causes a decrease in water

density, which would result in a net increase

in reactivity when soluble boron is present in

the water and Boraflex neutron absorber

panels are present in the racks. However, the

additional negative reactivity provided by the

1950 ppm boron concentration limit, above

that provided by the concentration required

(650 ppm) to maintain Keff less than or equal

to 0.95, will compensate for the increased

reactivity which could result from a loss of

SFP cooling event. Because adequate soluble

boron will be maintained in the SFP water,

the consequences of a loss of normal cooling

to the SFP will not be increased.

The Fresh Fuel racks are analyzed by

employing the ‘‘Westinghouse Spent Fuel

Rack Criticality Analysis Methodology’’

approved by the NRC and described in

WCAP–14416, NP–A, Revision 1. Only the

method for Fresh Fuel storage racks

criticality calculations has changed. The

method of handling fuel, the maximum fuel

enrichment, and the limiting values for

criticality have not changed. Therefore, there

is no change in the margin of safety for the

Fresh Fuel storage racks.

Therefore, based on the conclusions of the

above analysis, the proposed changes will

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any previously

evaluated.

Spent fuel handling accidents are not new

or different types of accidents, they have

been analyzed in Section 14.2.1 of the

Updated Final Safety Analysis Report

(UFSAR). Criticality accidents in the SFP are

not new or different types of accidents, they

have been analyzed in the UFSAR and in the

spent fuel storage criticality analysis. Current

TS 3/4.9.14 already contains a limit on the

SFP boron concentration. The boron

concentration in the SFP has always been

maintained near the limit of the RWST boron

concentration for refueling purposes. The

current TS boron concentration requirement

for the SFP water conservatively bounds the

boration assumptions of the revised

criticality analyses. Since soluble boron has

always been maintained in the SFP water, the

implementation of this requirement for

criticality purposes will have no effect on

normal pool operations and maintenance.

Since soluble boron has always been

present in the SFP, a dilution of the SFP

soluble boron has always been a possibility.

However, it was shown in the SFP dilution

analysis that a dilution of the Turkey Point

SFP which could increase the spent fuel

storage rack Keff to greater than 0.95 is not a

credible event. Therefore, the

implementation of limitations on the SFP

boron concentration for criticality purposes

will not result in the possibility of a new or

different kind of accident.

Proposed TS 3/4.9.14 Table 3.9–1 specifies

the requirements for the spent fuel rack

storage, which is currently contained in the

TS. These proposed new SFP storage

limitations are consistent with the

assumptions made in the spent fuel rack

criticality analysis, and will not have any

significant effect on normal SFP operations

and maintenance, and will not create any

possibility of a new or different kind of

accident. Verifications will continue to be

performed to ensure that the SFP loading

configuration meets specified requirements.

The Fresh Fuel racks are analyzed by

employing the ‘‘Westinghouse Spent Fuel

Rack Criticality Analysis Methodology’’

approved by the NRC and described in

WCAP–14416, NP–A, Revision 1. Only the

method for Fresh Fuel storage racks

criticality calculations has changed. The

method of handling fuel, the maximum fuel

enrichment, and the limiting values for

criticality have not changed. Therefore, there

is no change in the margin of safety for the

Fresh Fuel storage racks.

As discussed above, the proposed changes

will not create the possibility of a new or

different kind of accident from any

previously evaluated. There is no significant

change in plant configuration, equipment

design or equipment.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The proposed TS changes will provide

adequate safety margin to ensure that the

stored fuel assembly array will always

remain subcritical. Those limits are based on

a plant specific criticality analyses performed

in accordance with the NRC approved

Westinghouse Spent Fuel Rack criticality

analysis methodology.

The criticality analysis takes credit for

soluble boron to ensure that Keff will be less

than or equal to 0.95 under normal

circumstances. Storage configurations have

been defined using a 95/95 Keff calculation to

ensure that the spent fuel rack Keff will be

less than 1.0 with no soluble boron. Soluble

boron credit is used to provide safety margin

by maintaining Keff less than or equal to 0.95,

including uncertainties, tolerances, and

accident conditions in the presence of SFP

soluble boron.

The loss of substantial amounts of soluble

boron from the SFP that could lead to

exceeding a Keff of 0.95 has been evaluated

in the SFP Dilution analysis and shown to be

not credible.

The analysis shows that the dilution of the

SFP boron concentration from 1950 ppm to

650 ppm is not credible. When this result is

combined with the results from the 95/95

criticality analyses, which show that the

spent fuel rack Keff will remain less than 1.0

when flooded with unborated water, it

provides a level of safety comparable to the

25766 Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices

conservative criticality analysis methodology

required by ANSI 57.2–1983, NUREG–0800,

and Regulatory Guide 1.13.

The Fresh Fuel racks are analyzed by

employing the ‘‘Westinghouse Spent Fuel

Rack Criticality Analysis Methodology’’

approved by the NRC and described in

WCAP–14416, NP–A, Revision 1. Only the

method for Fresh Fuel storage racks

criticality calculations has changed. The

method of handling fuel, the maximum fuel

enrichment, and the limiting values for

criticality have not changed. Therefore, there

is no change in the margin of safety for the

Fresh Fuel storage racks.

Therefore, the proposed changes in these

license amendments will not result in a

significant reduction in the plant’s margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

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.

GPU Nuclear, Inc. et al., Docket No. 50–

219, Oyster Creek Nuclear Generating

Station, Ocean County, New Jersey

Date of amendment request: June 3,

1999, as supplemented on December 22,

1999

Description of amendment request:

The proposed amendment would permit

continued plant operation with a

maximum of two inoperable

recirculation loops, provided certain

conditions are met. Oyster Creek’s

Technical Specifications (TSs), Section

3.3.F.2 currently permit operation with

4 of the 5 recirculation loops with

certain constraints. If only 3 loops are

operable, however, the TSs require plant

shutdown within 12 hours. Analysis

indicates that the plant may be safely

operated at 90 percent power with three

operable recirculation loops.

Two definitions are added to Section

1 of the TSs to specify the difference

between an idle recirculation loop and

an isolated recirculation loop. These

definitions have been incorporated into

the specification to provide an explicit

description of acceptable valve

configurations. In addition, several

paragraphs have been added to the

Bases of Section 3.3 and one paragraph

in the Bases of Section 3.10 has been

modified. In each case the Bases section

has been segmented from the

specification, which affects the

pagination of the Bases.

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:

1. The proposed TS change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

When operating with two inoperable

recirculation loops, the proposed section

3.3.F.2.b requires that the reactor core

thermal power not exceed 90% of rated

power. This is a physical limitation of the

plant conditions because maximum power is

about 90% of rated power at the maximum

recirculation flow with only three

recirculation pumps operating. As such, the

90% of rated power becomes a limiting

condition for three-loop operation. The

licensee states that the results of this analysis

conform to all the requirements of 10 CFR

50.46 and Appendix K.

The licensee analyzed recirculation pump

trip transients for single and multiple pump

trips. Although the transient in general is

very mild, the licensee considers the case of

simultaneous trip of all five pumps to be the

limiting event among all possible

recirculation pump trip events. For threeloop operation, given the requirement that

the power level be maintained at or below

90% of rated power, the transient resulting

from the loss of all three pumps would be

bounded by the five-pump-trip event.

The proposed change, which permits three

loop operation with a maximum of two idle

or one idle and one fully isolated loop, will

provide adequate safety margins during

transient and accident conditions. The

proposed changes do not affect any accident

precursors because the accident occurrence is

not dependent on the number of operating

recirculation loops. Therefore, the probability

of an accident previously evaluated is not

increased. The proposed TS change will

assure the ability of systems to perform their

intended function. Therefore, the proposed

changes will not introduce a significant

increase in the consequences of an accident

previously evaluated. Therefore, the

probability of occurrence or the

consequences of an accident previously

evaluated in the Safety Analysis Report

(SAR) will not increase as a result of these

changes.

2. The proposed TS change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes permit three-loop

operation with a maximum of two idle or one

idle and one fully isolated loop. The licensee

considers the case of simultaneous trip of all

the five pumps to be the limiting event

among all possible recirculation pump trip

events. For three-loop operation, given the

requirement that the power level be

maintained at or below 90% of rated power,

the transient resulting from the loss of all

three pumps would be bounded by the fivepump-trip event.

The proposed changes will not create a

possibility for an accident or transient of a

different type than any previously identified

in the SAR.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety.

The proposed changes will not decrease

the margin of safety as defined in the basis

of any Technical Specification. All relevant

transient and accident scenarios have been

analyzed for the conditions of three-loop

operation and have demonstrated adequate

margin to safety limits. Therefore, the

proposed changes do not involve a

significant reduction in the margin of safety.

They neither adversely affect the

performance characteristics of systems nor do

they affect the ability of systems to perform

their intended function. Therefore, the

proposed changes do not involve a

significant reduction in the margin of safety.

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: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: M. Gamberoni,

Acting.

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: March

17, 2000.

Description of amendment request:

This amendment request proposes to

revise the Cooper Nuclear Station

Technical Specifications to incorporate

the recommended Generic Letter 99–02,

‘‘Laboratory Testing of Nuclear-Grade

Activated Charcoal,’’ laboratory testing

protocol of American Society for Testing

and Materials (ASTM) D3803–1989 for

Engineered Safety Feature ventilation

system charcoal samples.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed charcoal testing changes and

explicit reference to American Society for

Testing and Materials (ASTM) D3803–1989

nuclear-grade activated charcoal test protocol

do not affect Engineered Safety Feature (ESF)

ventilation system operation or performance,

reliability, actuation setpoints, or accident

mitigation capabilities. The proposed

Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices 25767

changes also do not affect the operation and

performance of any other equipment

important to safety at Cooper Nuclear Station

(CNS). ASTM D3803–1989 is a more accurate

and demanding test which ensures that the

charcoal filter efficiencies assumed in the

CNS accident dose analysis are maintained.

The proposed changes involve ESF

ventilation system charcoal testing only and

do not affect accident initiators. Therefore

the proposed changes do not significantly

increase the probability or consequences of

an accident previously evaluated, as revised

by the design basis accident radiological

assessment calculational methodology

revision submitted to the NRC under

Reference 3 [in the March 17, 2000,

amendment request].

2. Does not create the possibility for a new

or different kind of accident from any

accident previously evaluated.

The charcoal testing changes, and explicit

reference to ASTM D3803–1989 nucleargrade activated charcoal test protocol, do not

affect ESF ventilation system operation or

performance, or the operation and

performance of any other equipment

important to safety at CNS. The proposed

changes clarify and explicitly identify the

testing of the ESF ventilation system charcoal

samples. No new or different accident

scenarios, transient precursors, failure

mechanisms, plant operating modes, or

limiting single failures are introduced as a

result of these changes. Therefore, the

possibility of a new or different kind of

accident from that previously evaluated, as

revised by the design basis accident

radiological assessment calculational

methodology revision submitted to the NRC

under Reference 3, is not created by this

change.

3. Does not create a significant reduction

in the margin of safety.

The required performance of the ESF

ventilation systems following a design basis

accident is not impacted by utilizing a more

demanding protocol for charcoal testing.

Thus, the margin of safety assumed in the

CNS accident analysis, as revised by the

design basis accident radiological assessment

calculational methodology revision

submitted to the NRC under Reference 3, is

maintained. Revising the Technical

Specifications to clarify charcoal testing

methodology and explicitly referencing the

charcoal [adsorber] testing being performed

does not affect ESF ventilation system

performance or operation, or the operation

and performance of any other equipment

important to safety at CNS. Therefore, these

changes do not result in a significant

reduction in any 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: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus,

NE 68602–0499.

NRC Section Chief: Robert A. Gramm.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of amendment requests:

December 23, 1999.

Description of amendment requests:

The proposed amendment would revise

improved TS (ITS) 5.5.9.d.1.j)(iv) to

change the tube support plate (TSP)

intersections that are excluded from

application of steam generator (SG) tube

voltage based repair criteria for outside

diameter stress corrosion cracking

indications at TSPs.

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.

Application of a smaller wedge region

exclusion zone [due to loss of coolant

accident (LOCA) plus safe shutdown

earthquake (SSE)] and a new seventh tube

support plate (TSP) bending stress exclusion

zone (due to feedline break (FLB)/steamline

break (SLB) plus SSE) with respect to

alternate repair criteria (ARC), does not

increase the probability of tube burst or

leakage following a postulated main steam

line break (MSLB). Exclusion zones tubes

will be inspected by bobbin every outage and

by rotating pancake coil (RPC) if bobbin

detects degradation. Tubes containing RPCconfirmed crack-like degradation at wedge

region exclusion zone intersections and at

the seventh TSP bending exclusion zone

intersections will be plugged.

Tube burst criteria are inherently satisfied

during normal operating conditions because

of the proximity of the TSP. It is

conservatively assumed that the entire

crevice region is uncovered because of TSP

displacement during the secondary side

blowdown of a MSLB. Therefore, during a

postulated MSLB accident, tube burst

capability must exceed the Regulatory Guide

1.121 criterion requiring a margin of 1.43

times the SLB pressure differential on tube

burst.

Relative to the expected leakage during

accident condition loadings, a postulated

MSLB outside of containment, but upstream

of the main steam isolation valve, represents

the most limiting radiological condition. The

steam generator (SG) tubes are subjected to

an increase in differential pressure following

a MSLB, resulting in a postulated increase in

leakage and associated offsite doses. Leakage

following a MSLB bypasses containment.

Following each inspection, condition

monitoring will be performed to verify that

tube burst and leakage performance criteria

were satisfied for all degradation.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Implementation of revised ARC exclusion

zones does not introduce any significant

change to the plant design basis. Use of new

exclusion zones does not create a mechanism

which could result in an accident in the free

span. It is expected that for all plant

conditions, neither a single nor multiple tube

rupture event would likely occur in a SG

where ARC exclusion zones have been

applied.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

Revised wedge region exclusion zones are

based on a DCPP-specific analysis under

locked tube conditions for the combined

effects of a LOCA and SSE. The number of

wedge region tubes that are predicted to

collapse has been decreased when compared

to the prior analysis, which used highly

conservative assumptions. The revised

analysis incorporates DCPP-specific LOCA

and seismic loads that were not available

when the prior analysis was performed.

However, the revised analysis also yields

conservative results, such that the number of

tubes in the exclusion zone (244 per SG)

bound the number of tubes calculated to

collapse (144 per SG). Tubes located in the

revised wedge region exclusion zone will

continue to be subject to enhanced eddy

current inspection requirements and will be

excluded from application of ARC. Thus,

existing tube integrity requirements apply to

these tubes and the margin of safety is not

reduced.

New seventh TSP bending exclusion zones

are also based on a DCPP-specific analysis

under locked tube conditions for the

combined effects of a FLB/SLB and SSE. The

analysis yields conservative results, such that

914 tubes per SG at the seventh TSP are

assumed to exceed the Westinghouse lower

tolerance limit yield stress of the tubing.

Tubes located in the seventh TSP bending

exclusion zone will be subject to enhanced

eddy current inspection requirements and

will be excluded from application of ARC.

Thus, existing tube integrity requirements

apply to these tubes and the margin of safety

is not reduced.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Christopher J.

Warner, Esq., Pacific Gas and Electric

25768 Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices

Company, P.O. Box 7442, San

Francisco, California 94120.

NRC Section Chief: Stephen Dembek.

South Carolina Electric & Gas Company

(SCE&G), South Carolina Public Service

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

Summer Nuclear Station, Unit No. 1,

Fairfield County, South Carolina

Date of amendment request: April 6,

2000.

Description of amendment request:

The Virgil C. Summer Nuclear Station

Technical Specifications are being

revised to change the definitions and

surveillance requirements for response

time testing of the Engineered Safety

Feature Actuation System (ESFAS) and

the Reactor Trip System (RTS). These

changes will permit the verification of

response time, whereas the current

definitions imply the response time

must be measured.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

This change to the Technical specifications

(TS) does not result in a condition where the

design, material, and construction standards

that were applicable prior to the change are

altered. The same RTS and ESFAS

instrumentation is being used; the time

response allocations/modeling assumptions

in the Final Safety Analysis Report (FSAR)

Chapter 15 analyses are still the same; only

the method of verifying the time response is

changed. The proposed change will not

modify any system interface and could not

increase the likelihood of an accident since

these events are independent of this change.

The proposed change will not change,

degrade or prevent actions or alter any

assumptions previously made in evaluating

the radiological consequences of an accident

described in the FSAR.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

This change does not alter the performance

of process protection racks, Nuclear

Instrumentation, and logic systems used in

the plant protection systems. These systems

will still have response time verified by test

before being placed in operational service.

Changing the method of periodically

verifying instrument[ation] for these systems

(assuring equipment operability) from

response time testing to calibration and

channel checks will not create any new

accident initiators or scenarios. Periodic

surveillance of these systems will continue

and may be used to detect degradation that

could cause the response time to exceed the

total allowance. The total time response

allowance for each function bounds all

degradation that cannot be detected by

periodic surveillance. Implementation of the

proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Does this change involve a significant

reduction in margin of safety?

This change does not affect the total system

response time assumed in the safety analysis.

The periodic system response time

verification method for the process

protection racks, Nuclear Instrumentation,

and logic systems is modified to allow the

use of actual test data or engineering data.

The method of verification still provides

assurance that the total system response is

within that defined in the safety analysis,

since calibration tests will continue to be

performed and may be used to detect any

degradation which might cause the system

response time to exceed the total allowance.

The total response time allowance for each

function bounds all degradation that cannot

be detected by periodic surveillance. Based

on the above, it is concluded that the

proposed change does not result in a

significant reduction in margin with respect

to plant safety.

Pursuant to 10 CFR 50.91, the preceding

analyses provides a determination that the

proposed Technical Specifications change

poses no significant hazard as delineated by

10 CFR 50.92.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Randolph R.

Mahan, South Carolina Electric & Gas

Company, Post Office Box 764,

Columbia, South Carolina 29218.

NRC Section Chief: Richard L. Emch,

Jr.

South Carolina Electric & Gas Company

(SCE&G), South Carolina Public Service

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

Summer Nuclear Station (VCSNS), Unit

No. 1, Fairfield County, South Carolina

Date of amendment request: April 6,

2000.

Description of amendment request:

The proposed Technical Specifications

change request (TSCR) seeks to remove

the prescriptive testing requirements of

TS 4.8.1.1.2.i.2 to allow the ASME Code

Class 3 portions of the diesel fuel oil

system to be pressure tested in

accordance with Section XI of the

ASME Boiler and Pressure Vessel Code

as required by TS 4.0.5. This will permit

the use of Code Case N–498–1 as

accepted by Regulatory Guide 1.147,

Revision 12, for assessment of the diesel

fuel oil system pressure boundary

integrity.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

Industry experience has shown that an

inservice leak test conducted at normal

operating temperature and pressure is just as

effective at finding leakage as a hydrostatic

test conducted at 110% of the design

pressure.

Therefore, there is no increase in the

probability or consequences of previously

evaluated accidents.

Also note that the Diesel Generator Fuel

Oil System is not specifically modeled in the

VCSNS Probability Risk Assessment. It is

contained in the diesel generator fail to run

event that has a probability of 5.8E–2. If the

diesel generator fuel oil system had been

modeled, pipe ruptures would not have been

included because they would be dominated

by failure of other components such as check

valves which have failure probabilities

several orders of magnitude higher.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed TSCR provides an

alternative method of leak detection for the

required 10-year inservice inspection. It does

not result in an operational condition

different from that which has already been

considered by TS. Therefore, the change does

not create the possibility of a new or different

kind of accident or malfunction.

3. Does this change involve a significant

reduction in margin of safety?

The alternative method of leak detection

has no impact on the consequences of any

analyzed accident and does not significantly

change the failure probability of equipment

which provides protection for the health and

safety of the public. Therefore, there is no

significant decrease in the margin of safety.

Pursuant to 10 CFR 50.91, the preceding

analyses provides a determination that the

proposed Technical Specifications change

poses no significant hazard as delineated by

10 CFR 50.92.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Randolph R.

Mahan, South Carolina Electric & Gas

Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices 25769

Company, Post Office Box 764,

Columbia, South Carolina 29218.

NRC Section Chief: Richard L. Emch,

Jr.

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: March

30, 2000 (PCN–515).

Description of amendment requests:

The amendment application proposes to

revise the San Onofre Nuclear

Generating Station, Units 2 and 3,

Technical Specification (TS) 3.6.6.1,

‘‘Containment Spray and Cooling

Systems,’’ and the associated Bases. The

proposed change would revise the

Allowed Outage Time (AOT) for a single

inoperable train of the containment

spray system from 72 hours to 7 days

and revise the combined AOT of 10

days which appears in both Conditions

A and C of Limiting Condition for

Operation 3.6.6.1 from 10 days to 14

days.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Will 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 3.6.6.1,

‘‘Containment Spray and Cooling Systems’’

and the associated Bases. The proposed

change revises the Allowed Outage Time

(AOT) for a single inoperable train of the

Containment Spray System (CSS) from 72

hours to 7 days. The following changes are

proposed for the Containment Spray System

as described in Technical Specification (TS)

3.6.6.1:

a. The Allowed Outage Time (AOT) for a

single train of Containment Spray (Condition

A of LCO 3.6.6.1) is extended from 72 hours

to 7 days.

b. The Combined AOT of 10 days which

appears in both Conditions A and C of LCO

3.6.6.1 is extended from 10 days to 14 days.

c. The Bases of TS 3.6.6.1 are revised to

reflect the changes described above.

The Containment Spray System is an

Engineered Safety Feature (ESF) system.

Inoperable Containment Spray components

are not considered to be accident initiators.

Therefore, this change does not involve an

increase in the probability of an accident

previously evaluated.

The proposed AOT for the Containment

Spray System does impact the ability to

mitigate accident sequences. Therefore, to

fully evaluate the effects of the proposed CSS

AOT extension, Probabilistic Safety Analysis

(PSA) methods were utilized. The results of

these analyses show no significant increase

in core damage frequency. As a result, there

would be no significant increase in the

consequences of an accident previously

evaluated. Therefore, this change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

(2) 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, configuration, or method of operation

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

The proposed change does not affect the

limiting conditions for operation or their

bases that are used in the deterministic

analyses to establish the margin of safety.

PSA evaluations were used to evaluate these

changes.

Therefore, there will be no significant

reduction in a margin of safety as a result of

this change.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment 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.

Virginia Electric and Power Company,

Docket Nos. 50–280 and 50–281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request: March

17, 2000.

Description of amendment request:

The proposed changes would modify

the voltage setting limits specified in

Technical Specification (TS) Table 3.7–

4, page 3.7–26, item 7 for the emergency

bus degraded voltage, and revise the

loss of voltage setpoints from a

percentage of nominal bus voltage to an

actual bus voltage value. The degraded

voltage setting limit is being changed to

increase the minimum allowable bus

voltage to improve long-term motor

performance in the event of operation

with bus voltage less than nominal. The

emergency bus loss of voltage setting

limit is being revised to better address

expected relay performance over time

(i.e., setting drift). Section 3.6.B, page

3.6–1, of the TS would be changed to

revise the required reactor coolant

system conditions from the existing

wording of ‘‘350 degrees F or 450 psig’’

to ‘‘350 degrees F and 450 psig.’’

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:

We have reviewed the proposed change

against the criteria of 10 CFR 50.92 and have

concluded that the change does not pose a

significant safety hazards consideration as

defined therein. Specifically, operation of

Surry Power Station with the proposed

change will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

No increase in the probability of

occurrence or consequences of an accident

previously evaluated will result from the

proposed change in the setting limits for the

emergency bus degraded voltage and loss of

voltage relay setpoints. The proposed change

only affects actuation limits and therefore has

no bearing on the probability of an accident.

Neither the logic nor the function of the

undervoltage protection circuits is being

changed, nor is circuit or equipment

reliability being reduced. The higher

degraded voltage relay setpoint limit will

improve motor terminal voltage, and thus

promote longer motor life. Changing the

setpoint limit for the loss of voltage relays

will better characterize the relays’

capabilities and facilitate calibration.

Further, the performance characteristics of

the electrical distribution system and

components supplied (motors, etc.) are not

being altered, and compliance with GDC–17

[General Design Criterion] is being

maintained. The electrical distribution

system remains capable of performing its

safety function without spurious separation

of the emergency buses from offsite power. If

offsite power is lost, the capability of the

EDG’s [emergency diesel generators] to

perform their safety function is not altered.

Therefore, the probability of an accident

previously evaluated is not increased.

The consequences of an accident do not

increase since the proposed change

implements setting limits that will continue

to ensure that adequate voltages will be

available for the continuous operation of

safety-related equipment required to function

to mitigate a design basis accident. The

proposed setting limits for the emergency bus

degraded voltage and loss of voltage bound

the setpoints and initial conditions assumed

in the accident analyses and ensure that

appropriate protection is maintained.

The editorial change is administrative in

nature and consequently does not affect the

probability or consequences of an accident in

any way.

25770 Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Implementing the proposed Technical

Specifications emergency bus degraded

voltage and loss of voltage relay setting limits

cannot create the possibility of a new or

different kind of accident than any accident

previously evaluated. Revising the setpoint

setting limits does not introduce any new

accident precursors, and operation of the

electrical distribution system and the

undervoltage relaying schemes is unchanged.

Raising the setting limit for emergency bus

degraded voltage and decreasing the setting

limit for emergency bus loss of voltage do not

introduce any new accident precursors or

modes of operation. The relays will continue

to detect undervoltage conditions and

transfer safety loads to the emergency diesel

generators at a voltage level adequate to

ensure proper safety equipment performance

and to prevent long-term equipment

degradation due to undervoltage conditions.

The proposed setting limits include adequate

tolerances to calibrate the undervoltage

relays while ensuring that emergency bus

voltages remain above analytical limits. As

noted above, the performance characteristics

of the electrical distribution system and the

components being supplied are not being

altered, and compliance with GDC–17 is

being maintained. The proposed Technical

Specifications change will ensure that

appropriate electrical protection is available

as assumed in the safety analysis.

The editorial change is administrative in

nature and consequently does 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 change continues to ensure

that adequate voltage is available for safetyrelated equipment relied upon to respond to

a design basis accident. The proposed setting

limit for degraded bus voltage is conservative

with respect to the existing Technical

Specifications and ensures an adequate safety

margin is being maintained. Further, the

setting limit is maintained low enough to

prevent spurious actuations given expected

offsite grid voltages. The setting limit for the

emergency bus loss of voltage relays is being

changed to better characterize the relays’

capabilities and to facilitate calibration.

While the loss of bus voltage setting limit is

being reduced, sustained bus voltage in this

range is not credible. Furthermore, there is

no safety limit associated with the loss of

voltage setting limit.

The proposed change continues to ensure

that the setting limits for the emergency bus

degraded voltage and loss of voltage relays

bound the setpoints and initial conditions

assumed in the accident analyses and

ensures that appropriate electrical protection

is maintained. The editorial change is

administrative in nature and consequently

does not affect the safety analysis in any way.

Consequently, the margin of safety is not

being reduced by the proposed Technical

Specifications change.

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: Donald P. Irwin,

Esq., Hunton and Williams, Riverfront

Plaza, East Tower, 951 E. Byrd Street,

Richmond, Virginia 23219.

NRC Section Chief: Richard L. Emch,

Jr.

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Carolina Power & Light Company, et

al., Docket No. 50–325, Brunswick

Steam Electric Plant, Unit 1, Brunswick

County, North Carolina

Date of amendment request: April 14,

2000.

Description of amendment request:

The proposed amendment would

modify Surveillance Requirement

3.1.3.3 to allow partial insertion of

control rod 26–47 instead of insertion of

one complete notch. This revised

acceptance criterion would be limited to

the current Unit No. 1 operating cycle,

after which the current one-notch

requirement will be re-established.

Date of publication of individual

notice in Federal Register: April 21, 2000

(65 FR 21481).

Expiration date of individual notice:

May 22, 2000.

GPU Nuclear, Inc., Docket No. 50–320,

Three Mile Island Nuclear Station, Unit

2, Middletown, Pennsylvania

Date of amendment request: April 6,

2000.

Brief description of amendment

request: The proposed amendment

would reflect an administrative name

change from GPU Nuclear Corporation

to GPU Nuclear, Inc. Furthermore, the

proposed license amendment makes an

editorial change to better describe TMI–

2’s use of site physical security, guard

training and qualification, and safeguard

contingency plans that are maintained

by the Three Mile Island Nuclear

Station, Unit 1, licensee, AmerGen

Energy Company, LLC. In addition, the

licensee requests that minor changes

(mainly in titles) be made in Section 6.0

of the Technical Specifications to reflect

the TMI–2 organizational and

administrative controls that will exist

following the sale of the Oyster Creek

Nuclear Generating Station.

Date of publication of individual

notice in Federal Register: April 21, 2000

(65 FR 21484).

Expiration date of individual notice:

May 21, 2000.

Notice of Issance 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

Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices 25771

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).

Baltimore Gas and Electric Company,

Docket No. 50–317, Calvert Cliffs

Nuclear Power Plant, Unit No. 1,

Calvert County, Maryland

Date of application for amendment:

February 18, 2000, as supplemented

March 3, 2000.

Brief description of amendment: The

amendment approved resolution of an

issue involving the Societie Alsacienne

Construction Mechaniques Del

Melhouse (SACM) diesel generator (DG)

that constitutes an unreviewed safety

question. Specifically, a new failure

mode has been identified for DG 1A

SACM that is not adequately described

in the Updated Final Safety Analysis

Report. The manufacturer has indicated

that operating the engine in a light load

condition may degrade engine

performance and ultimately result in

engine failure.

Date of issuance: April 20, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 235.

Renewed Facility Operating License

No. DPR–53: Amendment revised the

Updated Final Safety Analysis Report.

Date of initial notice in Federal

Register: March 7, 2000 (65 FR 12038).

The March 3, 2000, submittal did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 20, 2000.

No significant hazards consideration

comments received: No.

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

May 27, 1999, as supplemented by

letters of February 25, 2000, March 30,

2000, and eMail of March 13, 2000.

Brief description of amendment: This

amendment revises the maximum

allowable service water temperature

permitted by Surveillance Requirement

3.7.8.2 for the ultimate heat sink (UHS)

from the currently permitted limit of 95

°F to 97 °F while it restores the original

Technical Specifications provisions for

required action and completion times of

6/36 hours to be in mode 3/5,

respectively, in the event the UHS

temperature were to exceed 97 °F.

Date of issuance: April 18, 2000.

Effective date: April 18, 2000.

Amendment No.: 187.

Facility Operating License No. DPR–

23: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR 9001).

The supplements of February 25, March

13, and March 30, 2000, provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 18, 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:

May 5, 1999, as supplemented on

October 8, 1999.

Brief description of amendments: The

amendments resolved an Unreviewed

Safety Question (USQ) related to an

evaluation of the reactor building

ventilation system exhaust plenum

masonry walls. The amendments

approved the use of different

methodology and acceptance criteria for

the reassessment of certain masonry

walls subjected to transient

pressurization loads resulting from a

high energy line break. This change to

the licensing basis, when evaluated by

the licensee in accordance with 10 CFR

50.59, resulted in an USQ that required

prior approval by the NRC staff in

accordance with the provisions of 10

CFR 50.90.

Date of issuance: April 11, 2000.

Effective date: Immediately, to be

implemented during the next scheduled

Final Safety Analysis Report update.

Amendment Nos.: 139 and 124.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the Updated Final Safety

Analysis Report.

Date of initial notice in Federal

Register: June 16, 1999 (64 FR 32286).

The October 8, 1999, 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 April 11, 2000.

No significant hazards consideration

comments received: No.

Consolidated Edison Company of New

York, Docket No. 50–247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of application for amendment:

September 23, 1999.

Brief description of amendment: The

amendment relocates items associated

with instrumentation for toxic gas

monitoring from Technical

Specifications to the Updated Final

Safety Analysis Report.

Date of issuance: April 20, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 208

Facility Operating License No. DPR–

26: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 1, 1999 (64 FR

67332).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 20, 2000.

No significant hazards consideration

comments received: No.

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request:

December 22, 1999, as supplemented by

letters dated March 20, March 24 (2),

March 29, and April 5, 2000.

Brief description of amendment: The

amendment authorized revisions to the

radiological assessment calculational

methodology for the loss-of-coolant

accident (LOCA) and the control rod

drop accident (CRDA). The amendment

request was submitted to address

potential unreviewed safety questions

resulting from these revisions due to

instances of increased dose

consequences. Because of outstanding

issues involving various assumptions

used in these calculational

methodologies, the staff is deferring the

review of implementing this change on

a permanent basis. Subsequently, this

amendment is to be effective

immediately and remain effective until

Cooper Nuclear Station enters mode 4 in

preparation for refueling outage 20

(effectively, one operating cycle). Also,

the staff has deferred review of the

radiological assessment methodology

revisions for the fuel handling accident

(FHA) and the main steamline break

(MSLB) accident. It is anticipated that

Nebraska Public Power District (NPPD)

will resolve any outstanding issues

concerning these calculational

methodology revisions in a timely

manner in support of a permanent

change that is acceptable to the staff.

25772 Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices

Date of issuance: April 7, 2000.

Effective date: April 7, 2000, to be

implemented within 30 days and

remain effective until Cooper Nuclear

Station enters mode 4 in preparation for

refueling outage 20.

Amendment No.: 183.

Facility Operating License No. DPR–

46: The amendment authorizes changes

to the licensing basis and changes to the

operating license.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4280).

The March 20 and 24 (2), March 29, and

April 5, 2000, letters provided

additional clarifying information that

was within the scope of the original

application and Federal Register notice

and did not change the staff’s initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 7, 2000.

No significant hazards consideration

comments received: No.

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request:

December 15, 1999, as supplemented by

letters dated February 15 and April 8,

2000.

Brief description of amendment: The

technical specification change revises

the average power range monitors

(APRMs) neutron flux-high (flowbiased) allowable value based on a

revised power-to-flow map. The revised

power-to-flow map extends the current

plant operating domain to above the

rated rod line, to within an envelope

referred to as the maximum extended

load line limit (MELLL) and adds the

increased core flow (105 percent)

region.

Date of issuance: April 11, 2000.

Effective date: April 11, 2000, to be

implemented within 30 days.

Amendment No.: 184.

Facility Operating License No. DPR–

46: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4279).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated April 11, 2000.

No significant hazards consideration

comments received: No.

North Atlantic Energy Service

Corporation, et al., Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request:

November 19, 1999.

Description of amendment request:

The amendment revised Technical

Specification (TS) 6.4.3, ‘‘Nuclear Safety

Audit Review Committee (NSARC),’’ by

relocating the specific requirements of

this TS to the Quality Assurance

Program located in the Updated Final

Safety Analysis Report (UFSAR).

Date of issuance: April 11, 2000.

Effective date: As of its date of

issuance, and shall be implemented

within 90 days.

Amendment No.: 67.

Facility Operating License No. NPF–

86: Amendment revised the Technical

Specifications and authorized changes

to the UFSAR.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4281).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 11, 2000.

No significant hazards consideration

comments received: No.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

February 1, 2000.

Brief description of amendment: The

amendment revises limiting conditions

for operation (LCO) 3.0.1 and 3.0.2 and

adds LCO 3.0.5 to the Technical

Specifications (TSs) for Milstone 3. LCO

3.0.5 establishes allowances for

restoring equipment to service under

administrative controls when the

equipment has been removed from

service or declared inoperable to

comply with actions in the TSs.

Date of issuance: April 17, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 179.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 1, 2000 (65 FR 11092).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 17, 2000.

No significant hazards consideration

comments received: No.

Northern States Power Company,

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

Island Nuclear Generating Plant, Units

1 and 2, Goodhue County, Minnesota

Date of application for amendments:

November 10, 1999, as supplemented

February 25, 2000.

Brief description of amendments: The

amendments revise the elevated F-star

(EF*) distance for the steam generator

tubes specified in Technical

Specification 4.12.D.1.(l) following a

correction to a minor error in the

calculations supporting the current EF*

distance.

Date of issuance: April 19, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days of the date of issuance.

Amendment Nos.: 149 and 140.

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR 9010).

The February 25, 2000, supplemental

letter provided clarifying information

that did not change the staff’s initial

proposed no significant hazards

consideration determination and did not

expand it beyond the scope of the

original Federal Register notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 19, 2000.

No significant hazards consideration

comments received: No.

PECO Energy Company, Docket Nos.

50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

June 7, 1999, as supplemented

September 27, 1999.

Brief description of amendments:

Revised Technical Specifications

Section 3/4.4.3 to clarify the action

statement concerning inoperative

reactor coolant leakage detection

systems.

Date of issuance: April 5, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment Nos.: 140 and 103.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: July 14, 1999 (64 FR 38034).

The September 27, 1999, letter

provided 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 April 5, 2000.

No significant hazards consideration

comments received: No.

PECO Energy Company, Docket No. 50–

352, Limerick Generating Station, Unit

1, Montgomery County, Pennsylvania

Date of application for amendment:

October 14, 1999, as supplemented

February 11, 2000.

Brief description of amendment: This

amendment revised Technical

Specification (TS) Section 2.2, ‘‘Safety

Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices 25773

Limits and Limiting Safety Systems

Settings,’’ and TS Section 3.0/4.0,

‘‘Limiting Conditions for Operation and

Surveillance Requirements.’’

Date of issuance: April 12, 2000.

Effective date: Effective as of the date

of issuance and; Unit 1 shall be

implemented during the LGS Unit 1

refueling outage scheduled to begin

March 29, 2000.

Amendment No.: 141.

Facility Operating License No. NPF–

39. This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 1, 1999 (64 FR

67337).

The February 11, 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 amendment is contained in a

Safety Evaluation dated April 12, 2000.

No significant hazards consideration

comments received: No.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

February 26, 1998, as supplemented

October 14, 1999.

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) by changing the

value of the allowable containment

leakage rate to 1.5 percent per day and

correcting conflicting information in TS

Section 4.6.C, ‘‘Coolant Chemistry.’’

Date of issuance: April 14, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 261.

Facility Operating License No. DPR–

59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 22, 1998 (63 FR 19977).

The October 14, 1999, supplemental

letter provided clarifying information

that did not change the initial proposed

no significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 14, 2000.

No significant hazards consideration

comments received: No.

Rochester Gas and Electric

Corporation, Docket No. 50–244, R. E.

Ginna Nuclear Power Plant, Wayne

County, New York

Date of application for amendment:

November 30, 1999.

Brief description of amendment: The

amendment revises Technical

Specification 5.5.10, ‘‘Ventilation Filter

Testing Program’’ to meet the actions

requested by Generic Letter 99–02.

Date of issuance: April 12, 2000.

Effective date: April 12, 2000.

Amendment No.: 77.

Facility Operating License No. DPR–

18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 26, 2000 (65 FR 4290)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 12, 2000.

No significant hazards consideration

comments received: No.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

February 26, 1999.

Brief description of amendments:

These amendments revised the

Technical Specifications (TSs) to

eliminate inconsistencies and

redundancies in Section 3.8.1.1, action

statements involving inoperable offsite

AC circuits and combinations of

inoperable offsite power supplies and

emergency diesel generators.

Date of issuance: April 14, 2000.

Effective date: April 14, 2000.

Amendment Nos.: 255 and 246.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revised

the TSs.

Date of initial notice in Federal

Register: March 24, 1999 (64 FR 14287).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 14, 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

25774 Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices

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

2, 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

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).

Tennessee Valley Authority, Docket

Nos. 50–259, 50–260, and 50–296,

Browns Ferry Nuclear Plant, Units 1, 2,

and 3, Limestone County, Alabama

Date of application for amendments:

March 29, 2000 (TS–402).

Brief Description of amendments: The

amendments revised the Technical

Specifications (TS) requirements

applicable to opening of secondary

containment access doors.

Date of issuance: April 21, 2000.

Effective date: April 21, 2000.

Amendment Nos.: 238, 264, and 224.

Facility Operating License No. DPR–

33, DPR–52 and DPR–68: Amendments

revise the TS.

Public comments requested as to

proposed no significant hazards

consideration (NSHC): Yes (65 FR 18141

dated April 6, 2000). The notice

provided an opportunity to submit

comments on the Commission’s

proposed NSHC determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by April 20, 2000,

but indicated that if the Commission

makes a final NSHC determination, any

such hearing would take place after

issuance of the amendment.

Federal Register / Vol. 65, No. 86 / Wednesday, May 3, 2000 / Notices 25775

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, and final determination

of NSHC are contained in a Safety

Evaluation dated April 21, 2000.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

Dated at Rockville, Maryland, this 26th day

of April 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–10743 Filed 5–2–00; 8:45 am]

BILLING CODE 7590–01–P

OFFICE OF PERSONNEL

MANAGEMENT

Privacy Act of 1974; Amendment to a

System of Records

AGENCY: Office of Personnel

Management.

ACTION: Technical amendment of

existing routine use.

SUMMARY: This notice serves as a

technical amendment to an existing

routine use contained in OPM’s

CENTRAL–1 system of records.

DATES: The change will be effected

without further notice on June 12, 2000

unless comments are received that

would result in a contrary

determination.

ADDRESSES: Send written comments to

the Office of Personnel Management,

ATTN: Mary Beth Smith-Toomey, Office

of the Chief Information Officer, 1900 E

Street NW., Room 5415, Washington,

DC 20415–7900.

FOR FURTHER INFORMATION CONTACT:

Mary Beth Smith-Toomey, (202) 606–

8358.

SUPPLEMENTARY INFORMATION: In OPM’s

CENTRAL–1 system of records, routine

use(s) has been amended to move

‘‘requesting’’ in front of the word

‘‘States’’ to clarify that OPM can

disclose information to Federal agencies

regardless of whether they specifically

requested the information.

(s) To disclose information contained

in the Retirement Annuity Master File;

including the name, Social Security

Number, date of birth, sex, OPM’s claim

number, health benefit enrollment code,

retirement date, retirement code (type of

retirement), annuity rate, pay status of

case, correspondence address, and ZIP

code, of all Federal retirees and their

survivors to Federal agencies and

requesting States to help eliminate fraud

and abuse in the benefit programs

administered by the Federal agencies

and States (and those States to local

governments) and to collect debts and

overpayments owed to the Federal

Government, and to State governments

and their components.

Office of Personnel Management.

Janice R. Lachance,

Director.

[FR Doc. 00–10989 Filed 5–2–00; 8:45 am]

BILLING CODE 6325–01–P

SECURITIES AND EXCHANGE

COMMISSION

Proposed Collection; Comment

Request

Upon Written Request, Copies Available

From: Securities and Exchange Commission,

Office of Filings and Information Services,

Washington, DC 20549.

Extension:

Rule 17f–1(b)—SEC File No. 270–28—OMB

Control No. 3235–0032

Rule 17f–1(c) and Form X–17F–1A—SEC File

No. 270–29—OMB Control No. 3235–

0037

Rule 17h–1T and 17h–2T—SEC File No. 270–

359—OMB Control No. 3235–0410

Notice is hereby given that pursuant

to the Paperwork Reduction Act of 1995

(44 U.S.C. 3501 et seq.) the Securities

and Exchange Commission

(‘‘Commission’’) is soliciting comments

on the collections of information

summarized below. The Commission

plans to submit these existing

collections of information to the Office

of Management and Budget for

extension and approval.

Rule 17f–1(b) requires approximately

1,150 entities in the securities industry

to register in the Lost and Stolen

Securities Program. Registration fulfills

a statutory requirement that entities

report and inquire about missing, lost,

counterfeit, or stolen securities.

Registration also allows entities in the

securities industry to gain access to a

confidential database that stores

information for the program.

It is estimated that 1,150 entities will

register in the Lost and Stolen Securities

Program annually. It is also estimated

that each respondent will register one

time. The staff estimates that the average

number of hours necessary to comply

with the Rule 17f–1(b) is one-half hour.

The total burden in 575 hours annually

for respondents, based upon past

submissions. The average cost per hour

is approximately $50. Therefore, the

total cost of compliance for respondents

is $28,750.

Rule 17f–1(b) is a reporting rule and

does not specify a retention period. The

rule requires a one-time registration for

reporting institutions. Registering under

Rule 17f–1(b) is mandatory to obtain the

benefit of a central database that stores

information about missing, lost,

counterfeit, or stolen securities for the

Lost and Stolen Securities Program.

Reporting institutions required to

registered under Rule 17f–1(b) will not

be kept confidential, however, the Lost

and Stolen Securities Program database

will be kept confidential.

Rule 17f–1(c) and Form X–17F–1A

requires approximately 23,000 entities

in the securities industry to report lost,

stolen, missing, or counterfeit securities

to a central database. Form X–17F–1A

facilitates the accurate reporting and

precise and immediate data entry into

the central database. Reporting to the

central database fulfills a statutory

requirement that reporting institutions

report and inquire about missing, lost,

counterfeit, or stolen securities.

Reporting to the central database also

allows reporting institutions to gain

access to the database that stores

information for the Lost and Stolen

Securities Program.

It is estimated that 23,000 reporting

institutions will report that securities

are either missing, lost, counterfeit, or

stolen annually. It is also estimated that

each reporting institution will submit

this report 56 times each year. The staff

estimates that the average amount of

time necessary to comply with Rule

17f–1(c) and Form X–17F–1A is five

minutes. The total burden is 107,333

hours annually for respondents, based

upon past submissions. The average cost

per hour is approximately $50.

Therefore, the total cost of compliance

for respondent is $5,366,666.

Rule 17f–1(c) is a reporting rule and

does not specify a retention period. The

rule requires an incident-based

reporting requirement by the reporting

institutions when securities are

discovered missing, lost, counterfeit, or

stolen. Registering under Rule 17f–1(c)

is mandatory to obtain the benefit of a

central database that stores information

about missing, lost, counterfeit, or

stolen securities for the Lost and Stolen

Securities Program. Reporting

institutions required to register under

Rule 17f–1(c) will not be kept

confidential, however, the Lost and

Stolen Securities Program database will

be kept confidential.

Rules 17h–1T requires a broker-dealer

to maintain and preserve records and

other information concerning certain

entities that are associated with the