Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices 65337

amendment application are available

electronically through ADAMS and can

be accessed through the Public

Electronic Reading Room (PERR) link

<http://www.nrc.gov/NRC/ADAMS/

index.html> at the NRC Homepage.

A request for a hearing or petition for

leave to intervene should be filed within

30 days after publication of this notice

in the Federal Register, if possible. Any

request for hearing or petition for leave

to intervene shall be served by the

requestor or petitioner upon the

applicant, the Office of the General

Counsel, U.S. Nuclear Regulatory

Commission, Washington DC 20555; the

Secretary, U.S. Nuclear Regulatory

Commission, Washington, DC 20555;

and the Executive Secretary, U.S.

Department of State, Washington, DC

20520.

The information concerning the

amendment application follows.

NRC IMPORT LICENSE AMENDMENT APPLICATION

Name of applicant; Date of

application; Date received;

Application No.

Description of material

Country of origin

Material type Total qty End use

Starmet CMI; September

28, 2000; October 4,

2000; IW008/01.

Depleted uranium swart/

turnings; DU solid cylindrical pieces and.

Increase from 80,000 kgs

to 250,000 kgs DU.

DU will be recycled ........... United Kingdom.

Contaminated mineral oil .. Increase from 45,000 liters

to 240,000 liters mineral

oil.

Oil will be processed and

reused.

United Kingdom.

Dated this 25th day of October 2000 at

Rockville, Maryland.

For the Nuclear Regulatory Commission,

Ronald D. Hauber,

Deputy Director, Office of International

Programs.

[FR Doc. 00–28033 Filed 10–31–00; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–271]

Vermont Yankee Nuclear Power

Corporation; Correction

The October 18, 2000, Federal

Register contained a ‘‘Notice of

Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing,’’ (65 FR

62393) for the Vermont Yankee Nuclear

Power Station. This notice offered an

opportunity for comment or hearing

requests.

Inadvertently, this was the second

offering of such opportunity as a notice

had already been published in the

September 27, 2000, Federal Register

(65 FR 68111). The 30-day comment/

hearing request deadline is October 27,

2000, at 4:15 p.m. as stated in the

September 27, 2000, Federal Register.

Dated at Rockville, Maryland this 26th day

of October 2000.

For the Nuclear Regulatory Commission.

Richard P. Croteau, Sr.,

Project Manager, Section 2, Project

Directorate I, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–28034 Filed 10–31–00; 8:45 am]

BILLING CODE 7590–01–M

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from October 9,

2000, through October 20, 2000. The last

biweekly notice was published on

October 18, 2000.

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

Administration Services, Office of

Administration, U.S. Nuclear Regulatory

65338 Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices

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’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC

through September 22, 2000. The NRC

has relocated its Public Document Room

to the NRC’s headquarters building.

Effective September 26, 2000,

documents may be examined at the

NRC’s Public Document Room, located

at One White Flint North, 11555

Rockville Pike (first floor), Rockville,

Maryland 20852. The filing of requests

for a hearing and petitions for leave to

intervene is discussed below.

By December 1, 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, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852.

Publicly available records will be

accessible 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.

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, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852.

Publicly available records will be

accessible 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 Nuclear Power Plant,

Unit 1, Wake and Chatham Counties,

North Carolina

Date of amendment request: October

4, 2000.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) to

support the replacement of the current

Westinghouse Model D4 steam

generators (SGs) with Westinghouse

Model Delta 75 replacement steam

generators (RSGs), which is planned to

occur during the fall 2001 refueling

outage. The proposed changes to the TS

Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices 65339

are required in part as a result of the

physical differences between the

currently installed Model D4 SGs and

the Model Delta 75 RSGs. The Model

Delta 75 RSGs have thermally treated

alloy 690 tube material, which has been

proven through laboratory testing and

operational experience to provide

increased corrosion resistance compared

to the Inconel 600 tube material in the

Model D4 SGs. The licensee has

completed a comprehensive engineering

review program in support of the SG

replacement. This program evaluated

the difference between the current

Model D4 SGs and the Model Delta 75

RSGs in conjunction with restoration of

the original nominal reactor coolant

average temperature (Tavg) of 588.8°F.

The supporting analyses and

evaluations were performed to support

operations with the RSGs at the current

licensed core power of 2775 MWt and

also, where possible, at an uprated core

power level of 2900 MWt. This license

amendment application, however, does

not include a request for a power uprate.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change does not affect

accident initiators. The change in RCS

[reactor coolant system] volume, an input in

certain accident analyses, is acceptable since

applicable acceptance criteria continue to be

met. The change to the calculated residual

heat removal (RHR) cooldown time does not

adversely impact safe shutdown and meets

Standard Review Plan (SRP) requirements.

The analysis results for peak accident

pressure (Pa remain below the containment

design pressure limit. Revised Technical

Specification limits, such as SG minimum

water levels, have been analytically

determined based on the new SG design

features to preserve analysis assumptions.

Since the proposed change has no effect on

accident initiators, the probability of design

basis events is not impacted by these

analyses.

The leak rate assumed in accident analyses

is conservative and independent of the value

of Pa. Since the peak calculated accident

pressure remains below the design value, the

containment function as a barrier to the

release of radioactive material is not

adversely affected. Dose consequences have

been analyzed with respect to the proposed

change and applicable acceptance criteria

continue to be met.

HNP [Harris Nuclear Plant] structures,

systems and components, as well as the

RSGs, are designed to operate at the original

Tavg value of 588.8°F. The comprehensive

engineering review performed to support SG

replacement includes evaluation and analysis

results for transients and design basis

accidents that meet applicable acceptance

criteria. The proposed DNB [departure from

nucleate boiling] parameter values ensure the

core and reactor coolant system do not

exceed their safety limits. The trip setpoints

initiate safeguards actions that mitigate

postulated accident consequences. Since

plant systems will function as designed and

performance requirements are verified to be

acceptable with the proposed changes, there

are no impacts to accident initiators or

precursors. Therefore, these changes do not

involve a significant increase in the

probability of an accident previously

evaluated.

The RSGs are designed to minimize

potential for the types of tube degradation

experienced with the OSGs [original steam

generators]. The changes to reactor core

safety limits, RTS/ESFAS [reactor trip

system/engineered safety feature actuation

system] values and Tavg limit support

transient and accident analysis assumptions

relating to fuel clad failure. Since SG tube

and fuel clad integrity will continue to be

adequately maintained, these barriers to

release of radioactive material will perform

their required function. All dose

consequences continue to meet acceptance

criteria. As a result, the changes do not

involve a significant increase in

consequences of previously evaluated

accidents.

The proposed change reduces Technical

Specification limits on reactor coolant

activity to ensure that the dose consequences

of the proposed new SGTR [steam generator

tube rupture] analysis meet guidelines. The

Technical Specification limits are not

accident initiators and the proposed change

provides additional conservatism in the

limiting initial condition for RCS activity

level. The change in these limits functions

only to mitigate consequences of RCS liquid

and steam release events and does not impact

methods for plant operation. Since there are

no adverse impacts to initiators of accidents

previously evaluated, there is no significant

increase in the probability of such accidents.

The proposed change to the basis for

thyroid dose conversion factors is an

accepted update to that used in the analysis

of record. Dose consequences for all analyses,

including SGTR, remain within allowable

guidelines. Although there are increases to

previously reported dose consequences, the

increase is not significant since the results

remain under the SRP acceptance criteria.

Reducing the required time at hot standby

to maintain adequate CST [condensate

storage tank] minimum volume ensures

accident mitigation capability and is not

involved with accident initiation.

Conservative reactor trip setpoints during

power operation with inoperable main steam

line safety valves ensure accident analysis

assumptions for maximum secondary system

pressure. The proposed setpoint changes are

also unrelated to initiation of previously

evaluated accidents.

The removal of F* plugging criteria is

consistent with NUREG–0452, ‘‘Standard

Technical Specifications.’’ The remaining

Technical Specification plugging criterion is

based on a minimum wall thickness due to

wastage as determined by ASME [American

Society of Mechanical Engineers] Section XI.

The proposed change is conservative because

operation with F* degraded tubes is

eliminated.

The potential for a tube rupture is expected

to be acceptably low based on the

qualification analysis and testing for the

RSGs. The proposed program for periodic inservice inspection monitors the integrity of

the SG tubing to provide reasonable

assurance that there is sufficient time to take

proper and timely corrective action if any

tube degradation is detected. The deleted

OSG tube inspection requirements are not

applicable to the RSG due to design

differences.

SG tube inspections are not accident

initiators and the function of the inspection

program is to identify and eliminate, when

appropriate, accident precursor conditions.

Because of this, the probability of postulated

accidents (e.g., SG tube rupture) is not

increased.

Since the proposed changes are either

conservative (reduced RTS setpoints,

elimination of F* tube plugging criteria) or

related to mitigation of postulated accidents

(hot standby time prior to cooldown), there

is no significant increase in dose

consequences.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not adversely

impact methods of plant operation and new

methods of plant systems operation are not

created. The impacts of SG replacement on

NSSS [nuclear steam supply system] and

other plant systems, including the change in

RCS volume and other design differences,

have been evaluated and design and

performance requirements continue to be

met. Accordingly, no new accident scenarios,

failure mechanisms or limiting single failures

are introduced.

Changing Pa to reflect the results of current

analyses while maintaining its value below

the containment design pressure will not

introduce significant or adverse changes to

the plant design basis that would create a

new or different kind of accident.

Since HNP structures, systems and

components are designed for operation at the

proposed Tavg, returning to this value

following SG replacement does not introduce

a new method of plant operation. The

proposed reactor core safety limits, RTS/

ESFAS and Tavg limit changes do not

introduce new failure mechanisms or

limiting single failures since the values have

been shown to preserve transient and

accident analysis assumptions. The

acceptable evaluation results demonstrate

that the changes do not adversely affect or

challenge the performance or integrity of

safety related systems. Hence, no new

accident scenarios are created.

The proposed change to specific activity

limits does not modify plant systems,

65340 Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices

structures, or components or impact methods

of plant operation. The SGTR analysis that

provides the basis for the change

demonstrates adequate margin to overfill

based on revised emergency operating

procedures [EOPs]. Operator actions and

response times have been demonstrated to be

acceptable and to satisfy analysis

assumptions in simulator exercises. The

operator actions do not represent different

types of actions from those currently

included in the EOPs, and are not subject to

errors that would create new failure

mechanisms.

No new types of failures and no new or

different accident initiators are created. As a

result, the change will not introduce

significant or adverse changes to the plant

design basis that could lead to the creation

of a new or different kind of accident.

The revised thermal power restrictions

imposed by [the] Technical Specification[s]

when there are inoperable main steam line

safety valves are conservative and consistent

with the Technical Specification Bases

methodology. The reduction in hot standby

time to six hours continues to satisfy

requirements for safe shutdown. No new

failure mechanisms or limiting single failures

are introduced and there are no adverse

effects or challenges to the performance or

integrity of safety related systems.

Removing application of F* repair criteria

and sleeving methodologies from [the]

Technical Specification[s] with the RSGs

does not introduce significant or adverse

changes to the plant design basis that could

lead to a new or different kind of accident

being created. These requirements were

needed to address design and material

conditions of the original HNP SGs that are

not applicable to the RSGs. This change does

not alter the objective of SG surveillance

activities—maintaining the structural

integrity of this portion of the reactor coolant

system. The surveillance activities are

performed during outages and the proposed

surveillance program is consistent with

regulatory guidance. No new failures are

created.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

Analyses supporting the change reflect the

RSG design, including changes to primary

and secondary volume and minimum water

level for adequate heat sink. Mass and energy

and containment response analyses model

ESF [engineered safety features] systems and

safety analysis limits. The analysis results

demonstrate that applicable acceptance

criteria, including DNB, RCS pressure, peak

cladding temperature, containment pressure

and temperature, and dose, continue to be

met.

The calculated containment peak accident

pressures result in a value of Pa that remains

below the containment design pressure. The

margin of safety during design basis events

is preserved if Pa does not exceed the design

pressure. In addition, performance of the

containment leakage rate testing program

ensures the validity of the assumptions used

in the accident analyses.

All evaluations and analyses performed to

support SG replacement reflect the proposed

values. The results demonstrate that

applicable acceptance criteria (including

DNB, RCS pressure and temperature, LOCA

[loss-of-coolant accident] peak clad

temperature, containment pressure and

temperature, and dose) continue to be met.

Calculations confirm that the proposed RTS/

ESFAS values provide margin to the safety

analysis limits, thereby ensuring that the

margin of safety inherent in the safety

analysis limit itself is preserved.

The specific activity limit has no effect on

safety limits or limiting safety system

settings. The only impact of the specific

activity limits is on dose consequences for

certain design basis accidents such as SGTR.

The proposed change is in the conservative

direction for dose. The SGTR analysis is the

most limiting of these accident dose

evaluations. The proposed new analysis

demonstrates that the SGTR consequences

remain within allowable SRP guideline

limits, which establish the margin of safety.

Since the analysis results remain below the

SRP guidelines, the margin of safety is not

reduced.

The thermal power restrictions imposed by

[the] Technical Specification[s] when there

are inoperable main steam line safety valves

[ensure] sufficient relieving capacity during

all postulated transients, thereby maintaining

margin of safety for secondary system

pressure.

The proposed in-service inspection

program for the RSGs monitors the integrity

of the SG tubing and is consistent with

NUREG–0452. Since the F* repair criteria

and sleeving methods specifically addressed

OSG degradation issues, their removal does

not involve a reduction in a margin of safety.

Analyses demonstrate that the RSG tubing

retains adequate structural and leakage

integrity using the proposed plugging criteria

during normal, transient, and postulated

accident conditions. The RSG tubing design

and proposed plugging criteria are consistent

with applicable ASME requirements and do

not allow for operation with indications

identified by F* criteria.

The proposed plugging limit maintains the

margin of safety by providing for leakage

detection and shutdown in the event of an

unexpected tube leak while minimizing the

potential for excessive leakage or tube burst

in the event of [an] MSLB [main steam line

break] or LOCA.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

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.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of amendment request:

September 20, 2000.

Description of amendment request:

The proposed amendment would

change Technical Specification (TS)

5.5.7.d to decrease the maximum

allowed pressure drops across control

room emergency filtration (CREF) makeup and recirculation train filters and

charcoal adsorbers. Specifically, the

pressure drops would be changed from

‘‘6.0 inches WG [water gauge]’’ to ‘‘3.0

inches WG’’ for the CREF makeup train,

and from ‘‘8.0 inches WG’’ to ‘‘4.2

inches WG’’ for the CREF recirculation

train. Additionally, the words ‘‘(CREF

only)’’ would be removed to clarify that

standby gas treatment system (SGTS)

prefilter is included in the Ventilation

Filter Testing Program (VFTP).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes revise the pressure

drop acceptance criteria of TS Section 5.5.7.d

from ‘‘6.0 inches WG’’ to ‘‘3.0 inches WG’’ for

the CREF makeup train, and from ‘‘8.0 inches

WG’’ to ‘‘4.2 inches WG’’ for the CREF

recirculation train. The change in pressure

drop reflects the impact of reduced fan speed

on system characteristics. The removal of

‘‘(CREF only)’’ is editorial and clarifies the

inclusion of the SGTS prefilters in the VFTP.

These changes assure that the Surveillance

procedure appropriately demonstrate[s] the

ability of the CREF trains to perform its

specified function.

No changes in either system design or

operating strategies will be made as a result

of these changes. Thus, no opportunity exists

to increase the probability or consequences of

a previously analyzed accident. Therefore,

the proposed change does not involve a

significant increase in the probability or

consequences of a previously evaluated

accident.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not involve a

change to the plant operation. As a result, the

proposed change does not affect any of the

parameters or conditions that could

contribute to the initiation of any accidents.

This change involves the pressure drop

across the filters and adsorbers of the CREF

make-up and recirculation trains. The

Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices 65341

surveillance requirements for performing the

actual test have not changed. The VFTP tests

are performed such that the pressure drop

across the combined HEPA [high efficiency

particulate adsorber] filters, the prefilters,

and the charcoal adsorbers is less than the

value specified in the acceptance criteria in

accordance with Regulatory Guide 1.52,

Revision 2, and ANSI/ASME N510–1980 at

the system flowrate of plus or minus 10

percent. The change involves restrictive

changes in test acceptance criteria and test

methodology and no change in system

operation.

No new accident scenarios are created by

the lower value of filter and adsorber

pressure drop. No safety-related equipment

or safety functions are altered as a result of

this change. Additionally, the removal of

‘‘(CREF only)’’ is editorial and clarifies the

inclusion of the SGTS prefilters in the VFTP.

Therefore, the changes do not create the

possibility of a new or different kind of

accident or malfunction from those

previously analyzed.

3. The change does not involve a

significant reduction in the margin of safety.

The proposed change in combination with

existing restrictions within the TS provides

assurance that there are no credible

mechanisms to prevent the CCHVAC [control

center heating, ventilation, and air

conditioning] system from performing its

specified function. The maximum allowable

differential pressure is more restrictive and

corresponds to the reduced fan speed across

the HEPA filters, prefilters, and charcoal

adsorbers as a result of this change. There

will be no changes in system operating

strategies because of the inclusion of the

make-up moisture separator/prefilter and test

acceptance criteria for the CREF trains.

Additionally, the removal of ‘‘(CREF only)’’

is editorial and clarifies the inclusion of the

SGTS prefilters in the VFTP. 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: Peter

Marquardt, Legal Department, 688 WCB,

Detroit Edison Company, 2000 2nd

Avenue, Detroit, Michigan 48226–1279.

NRC Section Chief: Claudia M. Craig.

Duke Energy Corporation, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: June 13,

2000.

Description of amendment request:

The proposed amendments would make

administrative changes to the McGuire

Nuclear Station, Units 1 and 2, Facility

Operating Licenses (FOLs). Specifically,

they would: (1) Delete existing License

Conditions which have been met by

completed licensee’s actions or are

imposed by other regulatory

requirements, or (2) make other

miscellaneous changes to the McGuire

Nuclear Station, Units 1 and 2, FOLs.

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 the change involve a significant

increase in the probability or consequence of

an accident previously evaluated?

No. The proposed amendment to the FOLs

are either administrative, eliminate

duplication of other regulatory requirements

or delete License Conditions fully met by

Duke Energy Corporation. No actual plant

equipment, operating practices, or accident

analyses are affected by this proposed

amendment. Therefore, the proposed

amendment has no impact on the possibility

of any type of accident: new, different, or

previously evaluated.

2. Will the change create the possibility of

a new or different kind of accident from any

previously evaluated?

No. The proposed amendment to the FOLs

[are] either administrative, eliminate

duplication of other regulatory requirements

or delete License Conditions fully met by

Duke Energy Corporation. No actual plant

equipment, operating practices, or accident

analyses are affected by this proposed

amendment and no failure modes not

bounded by previously evaluated accidents

are created. Therefore, the proposed

amendment has no impact on the possibility

of any type of accident: new, different, or

previously evaluated.

3. Will the change involve a significant

reduction in a margin of safety?

No. Margin of safety is associated with

confidence in the ability of the fission

product barriers (i.e., fuel and fuel cladding,

reactor coolant system pressure boundary,

and containment structure) to limit the level

of radiation dose to the public. The proposed

amendment to the FOLs are either

administrative, eliminate duplication of other

regulatory requirements or delete License

Conditions fully met by Duke Energy

Corporation. Therefore, no reduction in any

existing margin of safety is involved.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Ms. Lisa F.

Vaughn, Duke Energy Corporation, 422

South Church Street, Charlotte, North

Carolina 28201–1006.

NRC Section Chief: Richard L. Emch,

Jr.

Duke Energy Corporation, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of amendment request: August

22, 2000.

Description of amendment request:

The proposed amendments would

revise the Technical Specifications and

associated Bases to limit the amount of

time a Refueling Water Storage Tank

level channel can be in a trip condition.

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 of consequences of

an accident previously evaluated?

No. The proposed change modifies the

allowed outage time that a channel of the

Refueling Water Storage Tank (RWST) can be

in the tripped condition from an indefinite

period of time to a more conservative

maximum of 48 hours. The Engineered Safety

Features Actuation System (ESFAS) is an

accident mitigating system, and not an

accident initiator. Therefore, the proposed

change will have no impact on any accident

probabilities. Accident consequences will not

be affected, as no changes are being made to

the plant involving a reduction in reliability

or effectiveness of the Emergency Core

Cooling System (ECCS). Consequently, any

previous evaluations associated with

accidents will not be affected by this change.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

No. The proposed change does not modify

the design or configuration of the plant. The

proposed change provides a more

conservative time limit for a channel to be in

the tripped condition. No physical changes

are being made to plant systems, structures

or components nor will the proposed change

reduce the ability of any of the safety related

equipment required for accident mitigation.

Consequently, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Does this change involve a significant

reduction in a margin of safety?

No. Margin of safety is related to the

confidence in the ability of the fission

product barriers to perform their design

functions during and following accident

conditions. These barriers include the fuel

cladding, the reactor coolant system, and the

containment system. The proposed change

provides a more restrictive time limit for a

channel of the RWST to be in a tripped

condition than is currently allowed by the

ITS. The performance of the fission product

barriers will not be degraded by the proposed

changes. Consequently, plant safety analyses

will not be affected.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

65342 Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Ms. Lisa F.

Vaughn, Duke Energy Corporation, 422

South Church Street, Charlotte, North

Carolina 28201–1006.

NRC Section Chief: Richard L. Emch,

Jr.

Entergy Nuclear Generation Company,

Docket No. 50–293, Pilgrim Nuclear

Power Station, Plymouth County,

Massachusetts

Date of amendment request:

September 1, 2000.

Description of amendment request:

The proposed amendment would make

a change to Pilgrim Technical

Specification Table 4.6–3. The proposed

change substitutes ‘‘21 (approx)’’ under

the column ‘‘Effective Full Power Years

(EFPY)’’ for the current ‘‘18 (approx).’’

This proposed change would defer the

second reactor vessel surveillance

capsule pull to allow Pilgrim to pursue

participation in the Boiling Water

Reactor Integrated Surveillance Program

Plan.

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 analysis is presented below:

The proposed changes do not involve

a significant increase in the probability

or consequences of an accident

previously evaluated.

Pressure-temperature (P/T) limits

(Pilgrim Technical Specifications

Figures 3.6.1, 3.6.2, and 3.6.3) are

imposed on the reactor coolant system

to ensure that adequate safety margins

against nonductile or rapidly

propagating failure exist during normal

operation, anticipated operational

occurrences, and system hydrostatic

tests. The P/T limits are related to the

nil-ductility reference temperature,

RTndt, as described in American

Society of Mechanical Engineers

(ASME) Section III, Appendix G.

Changes in the fracture toughness

properties of reactor pressure vessel

(RPV) beltline materials, resulting from

the neutron irradiation and the thermal

environment, are monitored by a

surveillance program in compliance

with the requirements of 10 CFR Part

50, Appendix H. The effect of neutron

fluence on the shift in the nil-ductility

reference temperature of pressure vessel

steel is predicted by methods given in

Regulatory Guide (RG) 1.99, Revision 2.

The proposed change is a revision of

the second surveillance capsule

withdrawal time, identified in

Technical Specification Table 4.6–3,

from approximately 18 effective full

power years (EFPY) to approximately 21

EFPY. This change will not affect P/T

limits as given in Pilgrim Technical

Specifications Figures 3.6.1, 3.6.2, and

3.6.3. This change is not related to any

accidents previously evaluated. This

change will not affect any plant safety

limits or limiting conditions of

operation. The proposed change will not

affect reactor pressure vessel

performance because no physical

changes are involved and Pilgrim vessel

P/T limits will remain in accordance

with RG 1.99, Revision 2 requirements.

The proposed change will not cause the

reactor pressure vessel or interfacing

systems to be operated outside of their

design or testing limits. Also, the

proposed change will not alter any

assumptions previously made in

evaluating the radiological

consequences of accidents.

Therefore, the probability or

consequences of accidents previously

evaluated will not be increased by the

proposed change.

The proposed changes do not create

the possibility of a new or different kind

of accident from any accident

previously evaluated.

The proposed change revises the

second RPV material surveillance

capsule withdrawal time in Pilgrim

Technical Specification Table 4.6–3

from approximately 18 EFPY to

approximately 21 EFPY. This proposed

change does not involve a modification

of the design of plant structures,

systems, or components (SSCs). The

proposed change will not impact the

manner in which the plant is operated

as plant operating and testing

procedures will not be affected by the

change. The proposed change will not

degrade the reliability of SSCs

important to safety because equipment

protection features will not be deleted

or modified, equipment redundancy or

independence will not be reduced,

supporting system performance will not

be downgraded, the frequency of

operation of equipment important to

safety will not be increased, and

increased or more severe testing of

equipment important to safety will not

be imposed. No new accident types or

failure modes will be introduced as a

result of the proposed change.

Therefore, the proposed change does not

create the possibility of a new or

different kind of accident from that

previously evaluated.

The proposed changes do not involve

a significant reduction in a margin of

safety.

Operation in accordance with the

existing P–T curves ensures reactor

vessel and cooling system integrity. The

capsule pull is a surveillance technique

that provides data for modification of

the curves. The margins will not change

without new data from the capsule pull.

The NRC-approved methods used to

develop the temperatures associated

with the current curves reside in

Pilgrim’s TSs and have been determined

to be conservative. The proposed change

will not affect any safety limits, limiting

safety system settings, or limiting

conditions of operation. The proposed

change does not represent a change in

initial conditions, or in a system

response time, or in any other parameter

affecting the course of an accident

analysis supporting the Bases of any

Technical Specification. Therefore, the

proposed changes do not involve a

significant reduction in any margins of

safety.

Based on the staff’s analysis, 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: J. M. Fulton,

Esquire, Assistant General Counsel,

Pilgrim Nuclear Power Station, 600

Rocky Hill Road, Plymouth,

Massachusetts 02360–5599

NRC Section Chief: James W. Clifford.

FirstEnergy Nuclear Operating

Company, Docket No. 50–346, DavisBesse Nuclear Power Station, Unit 1,

Ottawa County, Ohio

Date of amendment request: August 7,

2000.

Description of amendment request:

The proposed changes would revise

Technical Specification (TS) Bases 3/

4.3.1 and 3/4.3.2 to clarify the actions

that must be performed when Steam and

Feedwater Rupture Control System

(SFRCS) components and SFRCSactuated components are inoperable.

Specifically, the proposed changes

would provide guidance on which TS

actions are applicable for SFRCSactuated components. Also, the

proposed changes would introduce new

TS 3/4.7.1.8 which would provide

appropriate requirements for the Main

Feedwater Control Valves and the

Startup Feedwater Control Valves.

Additionally, the proposed changes

would introduce new TS 3/4.7.1.9

which would provide requirements for

the Turbine Stop Valves.

Basis for proposed no significant

hazards consideration determination:

Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices 65343

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

The Davis-Besse Nuclear Power

Station (DBNPS) has reviewed the

proposed changes and determined that

a significant hazards consideration does

not exist because operation of the DavisBesse Nuclear Power Station, Unit No.

1, in accordance with these changes

would:

1a. Not involve a significant increase

in the probability of an accident

previously evaluated because no such

accidents are affected by the proposed

changes. The amendment application

proposes to revise Technical

Specification (TS) Bases 3/4.3.1 and 3/

4.3.2, Reactor Protection System and

Safety System Instrumentation, to

clarify Steam and Feedwater Rupture

Control System (SFRCS) requirements;

to add new TS 3/4.7.1.8, Main

Feedwater Control Valves (MFCVs) and

Startup Feedwater Control Valves

(SFCVs); to add new TS 3/4.7.1.9,

Turbine Stop Valves (TSVs); and to

make associated changes to the Bases

and TS Index.

The proposed change would provide

clear guidance for the actions required

when SFRCS logic and SFRCS-actuated

components are inoperable. The new

TSs for MFCVs, SFCVs, and TSVs

provide appropriate operational

requirements for these SFRCS-actuated

components. The proposed change will

not change any system hardware or

testing requirements. Initiating

conditions and assumptions remain as

previously analyzed for accidents in the

DBNPS Updated Safety Analysis Report

(USAR). The proposed changes to the

TS Bases and Index are consistent with

the changes described above.

The proposed changes are consistent

with the intent of NUREG–1430,

‘‘Standard Technical Specifications—

Babcock and Wilcox Plants,’’ Revision

1, April 1995. The proposed change

does not involve a change to any plant

hardware and does not affect the

probability of any equipment

malfunction or accident-initiating event.

1b. Not involve a significant increase

in the consequences of an accident

previously evaluated because the source

term, containment isolation, or

radiological releases are not affected by

the proposed changes. The proposed

changes ensure that excessive

unavailability of SFRCS-actuated

equipment which is used to mitigate

accident consequences does not occur.

The reliability of the plant hardware

and the ability of the SFRCS-actuated

equipment to perform its safety function

are not affected. Existing system and

component redundancy is not affected

by the proposed changes. The existing

system and component operation is not

affected by the proposed changes, and

the assumptions used in evaluating the

radiological consequences in the DBNPS

USAR are not invalidated. Therefore, for

each postulated accident the

consequences remain bounded by the

consequences from the previously

evaluated accidents.

2. Not create the possibility of a new

or different kind of accident from any

accident previously evaluated because

these proposed changes do not involve

any physical changes to systems or

components, nor do they alter the

manner in which the systems or

components are operated.

3. Not involve a significant reduction

in a margin of safety because, for the

proposed changes, there are no new or

significant changes to the initial

conditions contributing to accident

severity or consequences. Accordingly,

there are no significant reductions in a

margin of safety.

On the basis of the above, the DBNPS

has determined that the License

Amendment Request does not involve a

significant hazards consideration. As

this License Amendment Request

concerns a proposed change to the

Technical Specifications that must be

reviewed by the Nuclear Regulatory

Commission, this License Amendment

Request does not constitute an

unreviewed safety question.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Mary E.

O’Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street,

Akron, OH 44308.

NRC Section Chief: Anthony J.

Mendiola.

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

Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request:

September 19, 2000.

Description of amendment request:

The proposed amendment revises the

Standby Liquid Control (SLC) boron

solution requirements in TS Figure

3.1.7–1 to ensure a minimum boron

concentration of 660 parts per million

(ppm) in the reactor.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The current TS SRs ensure acceptable SLC

boron solution volume and concentration

values which produce a minimum boron

concentration of 600 ppm in the reactor. This

proposed amendment revises the boron

solution requirements to ensure a minimum

boron concentration of 660 ppm in the

reactor. A minimum boron concentration of

660 ppm in the reactor is sufficient to initiate

and maintain reactor subcriticality as the

nuclear system cools. The DAEC is currently

pursuing the use of GE–14 fuel, power uprate

to 1912 MWth and extended cycle length.

Increasing the minimum boron concentration

to 660 ppm in the reactor will provide

adequate shutdown margin for these future

core designs. A General Electric analysis,

using the approved methods described in

Revision 14 of General Electric Standard

Application for Reactor Fuel (GESTAR II),

NEDE 24011–P–A, provides the basis for the

increase in minimum boron concentration.

This analysis assumes DAEC operation with

an equilibrium core of GE–14 fuel, operating

at 1912 MWth with 24 month cycles and

bounds all planned future core designs.

The concentration requirement of the SLC

boron solution will continue to comply with

the requirements of 10 CFR 50.62(c)(4).

The proposed editorial change to the title

of TS Figure 3.1.7–1 provides clarity only

and does not alter performance of TS SR

3.1.7.1 and SR 3.1.7.5.

Therefore, this proposed amendment will

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Ensuring a minimum boron concentration

of 660 ppm in the reactor will result in an

increase in the acceptable minimum volume

of boron solution in the SLC tank. The tank

can accept the increased volume of boron

solution. The tank low and high level alarms

will be altered; these alarms alert the

operator to take action in the event boron

solution volume is decreasing or increasing,

respectively. Increasing the volume in the

SLC tank and increases to the low and high

level alarm setpoints will have no effect on

operator actions to maintain system

operability. Revising the SLC boron solution

requirements to ensure a minimum boron

concentration of 660 ppm in the reactor

cannot create a new or different kind of

accident.

The proposed editorial change to the title

of TS Figure 3.1.7–1 provides consistent

wording between the Figure, TS SR 3.1.7.1

and the TS BASES to avoid potential

inconsistencies when determining the tank

volume acceptability.

Therefore this proposed amendment will

not create the possibility of a new or different

65344 Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices

kind of accident from any accident

previously evaluated.

3. The proposed amendment will not

involve a significant reduction in a margin of

safety.

DAEC TS SRs ensure SLC boron solution

volume and concentration values which

produce a minimum boron concentration of

600 ppm in the reactor. This proposed

amendment revises SLC boron solution

requirements to ensure a minimum boron

concentration of 660 ppm in the reactor. A

boron concentration of 660 ppm in the

reactor is sufficient to initiate and maintain

reactor subcriticality as the nuclear system

cools. The DAEC is currently pursuing the

use of GE–14 fuel, power uprate to 1912

MWth and extended cycle length. A General

Electric (GE) analysis, using the approved

methods described in Revision 14 of General

Electric Standard Application for Reactor

Fuel (GESTAR II), NEDE 24011–P–A,

provides the basis for the increase in

minimum boron concentration. GE’s analysis

demonstrated that at a minimum boron

concentration of 600 ppm in the reactor, the

SLC shutdown margin was marginal with

respect to the TS requirement for shutdown

margin. At a minimum boron concentration

of 660 ppm in the reactor, sufficient margin

is maintained such that TS Section 3.1.1 will

be met under all anticipated conditions. The

GE analysis assumes DAEC operation with an

equilibrium core of GE–14 fuel, operating at

1912 MWth with 24 month cycles and

bounds all planned future core designs.

The concentration requirement of the SLC

boron solution will continue to comply with

the requirements of 10 CFR 50.62(c)(4).

The SLC system is designed to inject a

quantity of boron that produces a minimum

boron concentration of 600 ppm in the

reactor. An additional 25% of that quantity

of boron is also injected to compensate for

imperfect mixing, leakage and volume in

other small piping connected to the reactor.

This margin will be maintained such that an

additional 25% of the quantity of boron

required to achieve a minimum boron

concentration of 660 ppm in the reactor will

also be injected. Therefore, this proposed

amendment will not involve a significant

reduction in a margin of safety.

The proposed editorial change to the title

of TS Figure 3.1.7–1 provides consistent

wording between the Figure, TS SR 3.1.7.1

and the TS BASES to avoid potential

inconsistencies when determining tank

volume acceptability.

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: Al Gutterman,

Morgan, Lewis & Bockius, 1800 M

Street, NW., Washington, DC 20036–

5869.

NRC Section Chief: Claudia M. Craig.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York

Date of amendment request:

September 26, 2000.

Description of amendment request:

The licensee proposed to amend the

unit’s Technical Specifications (TSs),

Sections 3.3.0, 3.3.2 thru 3.3.5 and 3.3.7

to allow performance of reactor vessel

hydrostatic or leakage tests, and scram

time and excess flow check valve tests,

when the reactor system temperature is

above 215 degrees Fahrenheit and the

reactor is not critical, without having to

maintain primary containment integrity.

These proposed changes are consistent

with the guidelines of NUREG–1433,

Revision 1, ‘‘Standard Technical

Specifications, General Electric Plants,

BWR/4.’’ Concurrently, the licensee

proposed to amend TS Sections 3.4.0

thru 3.4.5 to require reactor building

(secondary containment) integrity when

the reactor system temperature is above

215 degrees Fahrenheit. This will

ensure that, should accidents occur

during the above tests, the reactor

building will perform its function to

contain radiological releases. The

licensee also proposed to impose a doseequivalent iodine level of 1.5

microcuries per gram (as opposed to

9.47 microcuries per gram allowed

during plant operation) during the

above conditions to limit radiological

consequences of possible accidents to

limits bound by the previously analyzed

main steam line break accident.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below:

The operation of the unit in

accordance with the proposed

amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS changes will result

in not requiring containment integrity,

but requiring reactor building integrity

(secondary containment) during

performance of the specified tests.

These are changes in configuration (no

hardware design change is involved) to

permit testing of various systems or

components under different conditions.

The changed conditions are not

considered precursor of accidents.

Accordingly, the probability of an

accident previously evaluated is not

increased.

Accidents could occur during a test.

As part of the proposed TS change, the

licensee proposed to limit the doseequivalent iodine level to ensure that

the consequences of possible accidents

will continue to be bounded by a

previously analyzed accident, the main

steam line break accident. Thus, there

will not be any increase in the

consequences of an accident previously

evaluated.

The operation of the unit in

accordance with the proposed

amendment will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. No hardware design change

is involved with the proposed TS

changes. Other than the changes in the

configuration of the containment, the

reactor building, and the permissible

dose-equivalent iodine level, there are

no other procedural changes. Therefore,

the proposed amendment will not create

the possibility of a new or different kind

of accident from any previously

evaluated.

The operation of the unit in

accordance with the proposed

amendment will not involve a

significant reduction in a margin of

safety.

The proposed TS changes will not

adversely affect the performance

characteristics and intended functions

of systems and components covered by

the above-listed TS sections. The tests,

not affected by the proposed

amendment, will ensure the ability of

the subject systems and components to

perform their intended function.

Therefore, the proposed changes do not

involve a significant reduction in a

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: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Section Chief: Marsha

Gamberoni.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of amendment request: August

25, 2000.

Description of amendment request:

Millstone Unit No. 1 (MP1) is being

decommissioned. To support this

activity, several modifications are

Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices 65345

required to modify/eliminate MP1

systems that support the operation of

structures, systems, and components

that are shared or common to Millstone

Unit No. 2 (MP2) and Millstone Unit

No. 3 (MP3). One of the separation

projects entails the replacement of the

existing MP1 to MP2 4160-volt cross-tie

with a new MP3 to MP2 4160-volt crosstie. Northeast Nuclear Energy Company

(NNECO) has proposed a one time

extension to the allowed outage time

(AOT) for Action a.2 of Technical

Specification (TS) 3.8.1.1, ‘‘Electrical

Power Systems—A. C. Sources—

Operating’’ in order to complete this

modification. The proposed change

would extend the AOT from 72 hours to

14 days, provided the MP3 station

blackout (SBO) diesel generator (DG) is

available to supply MP2; otherwise the

AOT would only be extended to 7 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. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The offsite circuits supply power to

equipment required to support the safe

shutdown and post-accident operations of

MP2. The preferred off-site power supply is

from the 345 kV [kilovolt] switchyard,

through the reserve station service

transformer (RSST). The alternate (delayed)

source of offsite power is the 4.16 kV crosstie from MP1 via bus 14H.

To ensure that the probability of a

complete loss of offsite power is not

significantly increased, the MP1 4.16 kV

cross-tie will only be removed from service

when the weather conditions and forecast are

favorable. Additionally, during the time that

the alternate offsite source is inoperable,

actions will be taken to protect the operable

offsite circuit (i.e., no work will be conducted

that could challenge the operability of the

offsite circuit).

Although the offsite circuits provide power

to components that help mitigate the

consequences of accidents previously

evaluated, the extension in the AOT does not

affect any of the assumptions used in the

deterministic evaluations of these accidents.

Thus, this change will not increase the

consequences of any accident previously

analyzed.

Based on the above, the proposed change

does not involve a significant increase in the

probability or consequences of an accident

previously analyzed.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change is an extension to a

TS AOT. It does not alter the physical design,

configuration, or method of operation of the

plant. Therefore, the proposed change will

not create the possibility of a new or different

kind of accident from any previously

analyzed.

3. Involve a significant reduction in a

margin of safety.

During the implementation of the

modification to provide MP2 with a 4.16 kV

cross-tie with MP3, NNECO will:

a. Appropriately consider the 7 day and 14

day weather forecasts prior to removing the

MP2 4.16 kV cross-tie with MP1 from service

to minimize the potential for loss of offsite

power due to severe weather or salt spray.

b. Protect the equipment redundant to the

systems removed from service or whose

power supply is affected by the modification.

This includes limiting work on the 345 kV

lines, the switchyard, the RSST, the diesel

generators, the service water system, the high

pressure safety injection system, and the

reactor building closed cooling water system.

This restriction will ensure that MP2 will

remain capable of mitigating any potential

design basis accident during the

implementation of the modification.

c. Within 7 days of entering Action a. of

TS 3.8.1.1, establish the capability to supply

MP2 with power from the MP3 SBO DG via

operator actions within one hour of an event

resulting in a loss of the remaining offsite

source of power. The capability to utilize the

MP3 SBO DG is a contingency measure (i.e.,

will be able to serve as a temporary diesel).

Additionally, NNECO has evaluated the

dominant sequences affecting plant risk

using probabilistic safety analysis

techniques. The analysis determined that the

Delta Core Damage Probability associated

with the extended allowed outage time was

small.

There will be no significant reduction in a

margin of safety because the increased risk is

acceptable, focus on maintaining the

operability of the redundant equipment and

systems will be increased, the probable

weather conditions will be appropriately

considered, and the capability to supply MP2

with power from the MP3 SBO DG via

operator action will be established within 7

days of entering Action a. of TS 3.8.1.1. The

capability to utilize the MP3 SBO DG is a

contingency measure (i.e., will be able to

serve as a temporary diesel).

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Section Chief: James W. Clifford.

Northeast Nuclear Energy Company, et

al., Docket Nos. 50–336 and 50–423,

Millstone Nuclear Power Station, Unit

Nos. 2 and 3, New London County,

Connecticut

Date of amendment request: August

25, 2000.

Description of amendment request:

Millstone Unit No. 1 (MP1) is being

decommissioned. To support this

activity, several modifications are

required to modify/eliminate MP1

systems that support the operation of

structures, systems, and components

that are shared or common to Millstone

Unit No. 2 (MP2) and Millstone Unit

No. 3 (MP3). One of the separation

projects entails the replacement of the

existing MP1 to MP2 4160-volt cross-tie

with a new MP3 to MP2 4160-volt crosstie. Northeast Nuclear Energy Company

(licensee) has evaluated this proposed

new cross-tie utilizing the criteria of 10

CFR 50.59. The modification involves

four unreviewed safety questions

(USQs). One USQ pertains to MP2 and

three USQs pertain to MP3. The licensee

is requesting approval of the USQs.

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:

Basis for No Significant Hazards

Consideration—Millstone Unit No. 2

Unreviewed Safety Question

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The test that will be performed to verify

the ability of the cross-tie to supply

approximately 3 MVA of power from MP3 to

MP2 will require parallel operation of the

MP3 SBO [station blackout] DG [diesel

generator] with the MP2 4160 Volt system.

To ensure that the MP3 SBO DG and

associated connection will be isolated

immediately should any abnormal event or

failure occur at MP2, a special Class 1E

instantaneous overcurrent relaying scheme

will be used. This will provide additional

assurance that if a fault occurs in the MP3

SBO DG or the associated cross-tie cabling,

it will be isolated immediately and should

not affect operability of the MP2 safety

related 4160 Volt system. In addition, if a

65346 Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices

loss of normal power occurs at MP2 during

the test, the undervoltage condition will be

detected by the credited undervoltage

scheme, and MP2 will respond as designed

to the loss of power.

The performance of this test does not

significantly alter the manner in which the

plant is operated. There will be no adverse

effect on plant operation or accident

mitigation equipment. The response of the

plant and the operators following a design

basis event will not be significantly different.

Therefore, the proposed activity does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The installation of the Class 1E

instantaneous overcurrent relay scheme to be

used during performance of this test will

provide additional assurance that the MP2

safety related/busses will remain operable

during the test to parallel the MP3 SBO DG

to the MP2 4160 Volt system. The failure of

the temporary overcurrent protection scheme

would have the same effect as failure of the

existing overcurrent protection scheme.

Therefore, the proposed activity 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 installation of the Class 1E

instantaneous overcurrent relay scheme to be

used during performance of this test will

provide additional assurance that the MP2

safety related busses will remain operable

during the test to parallel the MP3 SBO DG

to the MP2 4160 Volt system. This will

provide additional assurance that if a fault

occurs in the MP3 SBO DG or the associated

cross-tie cabling, it will be isolated

immediately and should not affect

operability of the MP2 safety related 4160

Volt system. In addition, if a loss of normal

power occurs at MP2 during the test, the

undervoltage condition will be detected by

the credited undervoltage scheme, and MP2

will respond as designed to the loss of power.

The proposed activity will have no adverse

effect on plant operation or equipment

important to safety. The plant response to the

design basis accidents will not change and

the accident mitigation equipment will

continue to function as assumed in the

design basis accident analysis. Therefore, the

proposed activity does not involve a

significant reduction in a margin of safety.

Basis for No Significant Hazards

Consideration—Millstone Unit No. 3

Unreviewed Safety Questions

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

MP3 USQ 1

The MP3 to MP2 4160 Volt cross-tie will

impose up to an additional 3 MVA load on

MP3. As a result, the minimum acceptable

switchyard voltage for MP3 will increase

from 334 kV to 337 kV. The analysis

performed to support the new design shows

that, with a minimum voltage of 337 kV in

the switchyard, MP3 will remain connected

to offsite power. The minimum voltage

protection assures that acceptable starting

and running voltages are present for the

safety systems that may be required to

operate should a loss of coolant accident

(limiting design basis event) occur at MP3

while MP2 is being supplied with the safe

shutdown load of 3 MVA from either the

MP3 RSST [reserve station service

transformer] or the MP3 NSST [normal

station service transformer]. Thus, the MP3

offsite connection will not be tripped when

offsite power is available.

The extra loading imposed by the MP2 safe

shutdown loads on the MP3 power supplies

is within the MP3 capability to

simultaneously supply the worst case MP3

normal and accident loads. The ability of

MP3 to mitigate design basis accidents and

events is not adversely affected by the new

design.

This activity does not significantly alter the

manner in which the plant is operated. There

will be no adverse effect on plant operation

or accident mitigation equipment. Also, the

response of the plant and the operators

following a design basis event will not be

significantly different. Therefore, this activity

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

MP3 USQ2

The proposed design change will allow the

MP3 RSST (or the MP3 NSST, if the MP3

RSST is unavailable) to provide the alternate

offsite source for MP2 to meet GDC [General

Gesign Criterion] 17 requirements. If the MP3

RSST is not available, MP2 will need to

credit the MP3 NSST as the alternate offsite

power supply for GDC 17 compliance. The

connection for the MP3 NSST is between

breakers 13T and 14T. This connection point

only provides 1 breaker separation (13T)

between the MP2 RSST and the MP3 NSST.

If breaker 13T is closed, this arrangement

does not provide adequate separation

between the two offsite sources as required

by GDC 17. Opening breaker 13T when MP3

is shutdown will provide the required

separation for MP2, but it will reduce the

reliability of the offsite supply for MP3.

Therefore, opening the 13T breaker to allow

MP2 to meet GDC 17 requirements increases

the probability of a loss of offsite power at

MP3 when shutdown.

A loss of offsite power at MP3 due to a

fault in the offsite distribution network is a

low probability event. When combined with

the expected frequency of removing the MP3

RSST from service of once per MP3 refueling

outage, the probability of a loss of offsite

power at MP3 when shutdown as a result of

breaker 13T being open is low. In addition,

the MP3 shutdown risk program will

evaluate the impact of removing the MP3

RSST from service, and plan accordingly.

This will ensure the MP3 RSST will be

removed from service when the shutdown

risk is determined to be acceptably low. Also,

at least one MP3 EDG will be available, as

required by Technical Specifications, to

supply power to necessary loads if offsite

power is lost.

This activity does not significantly alter the

manner in which the plant is operated. There

will be no adverse effect on accident

mitigation equipment.

Also, the response of the plant and the

operators following a design basis event will

not be significantly different. Therefore, this

activity does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

MP3 USQ 3

The connection of an additional 3 MVA

load for MP2 onto the MP3 electrical

distribution system will increase the short

circuit fault levels on the MP3 electrical

distribution system. If worst case conditions

are established, the additional contribution

from the MP2 loads would increase the fault

level to above the switchgear ratings. The

worst case short circuit conditions occur

with MP3 at full power, the switchyard at its

maximum voltage, an MP3 EDG paralleled to

the bus for surveillance testing, and the MP3

to MP2 cross-tie supplying 3 MVA of power

to MP2.

Although the fault levels could exceed the

MP3 switchgear and circuit breaker ratings,

it is not expected that the worst case

conditions will be frequently established. For

the majority of scenarios, this situation can

be avoided because the new design allows

the 3 MVA of power for MP2 to be supplied

by either MP3 bus 34A or 34B. Procedural

guidance will be provided to select the MP3

bus (34A or 34B) to supply power to MP2

that powers the opposite train from the train

associated with the MP3 EDG to be tested.

Even if the worst case plant configuration

cannot be avoided, there is a low probability

of event occurrence given the low probability

of a bus fault coincident with the short time

that the MP3 EDG would be paralleled to the

system. In addition, the event would only

affect one MP3 train of safety related

equipment. The other train will be available

to function as assumed to mitigate any

accidents that may occur.

The use of the MP3 to MP2 cross-tie does

not significantly alter the manner in which

the plant is operated. Considering the low

likelihood of a fault occurring concurrently

with worst case conditions, and that only one

MP3 train of safety related equipment should

be affected, sufficient accident mitigation

equipment will be available to mitigate a

design basis accident. In addition, the

response of the plant and the operators

following a design basis event will not be

significantly different. Therefore, this activity

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

MP3 USQ 1

The ability to supply 3 MVA of power from

MP3 to MP2 will reduce the margin between

the minimum switchyard voltage and the

minimum acceptable switchyard voltage.

However, the extra loading imposed by the

MP2 safe shutdown loads on the MP3 power

supplies is within the MP3 capability to

simultaneously supply the worst case MP3

normal and accident loads. As a result, the

response of the plant and the operators

following an accident will not be

Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices 65347

significantly different. There will be no

adverse effect on accident mitigation

equipment and no new failure modes will be

introduced. Therefore, the activity will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

MP3 USQ 2

Maintaining breaker 13T and its associated

disconnect switches open to ensure

separation and independence of the two MP2

offsite sources is a change to the normal

configuration of MP3. The new configuration

involves the opening of a breaker. It does not

involve any other physical changes to the

plant or the operating methodology. This

activity does reduce the diversity of offsite

power supplies for the MP3 NSST, which

increases the potential for a loss of offsite

power at MP3. However, the plant response

to a loss of offsite power because breaker 13T

is open will not be significantly different

from any previously analyzed loss of offsite

power. Therefore, opening breaker 13T and

its associated disconnect switches will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

MP3 USQ 3

The connection of an additional 3 MVA

load for MP2 onto the MP3 electrical

distribution system will increase the short

circuit fault levels on the MP3 electrical

distribution system. In the worst case, the

short circuit fault could lead to a loss of one

MP3 train of safety related equipment.

However, the safety related equipment in the

other train would remain capable of

mitigating the event. The loss of a train of

safety-related equipment is an analyzed

event. Therefore, this activity will not

introduce the possibility of a new or different

kind of accident than any previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

MP3 USQ 1

The MP3 to MP2 4160 Volt cross-tie will

impose up to an additional 3 MVA load on

MP3. As a result, the minimum acceptable

switchyard voltage for MP3 will increase

from 334 kV to 337 kV. The margin between

the minimum switchyard voltage (345 kV)

and the minimum acceptable switchyard

voltage will decrease from 11 kV to 8 kV.

This is an improvement for MP2. However,

it is a reduction in the operating margin for

MP3.

The analysis performed to support the new

design shows that, with a minimum voltage

of 337 kV in the switchyard, MP3 will remain

connected to offsite power. The minimum

voltage protection assures that acceptable

starting and running voltages are present for

the safety systems that may be required to

operate should a loss of coolant accident

(limiting design basis event) occur at MP3

while MP2 is being supplied with the safe

shutdown load of 3 MVA from either the

MP3 RSST or the MP3 NSST. Thus, the MP3

offsite connection will not be tripped when

offsite power is available.

The extra loading imposed by the MP2 safe

shutdown loads on the MP3 power supplies

is within the MP3 capability to

simultaneously supply the worst case MP3

normal and accident loads. As a result, the

response of the plant and the operators

following an accident will not be

significantly different. There will be no

adverse effect on plant operation or

equipment important to safety. The plant

response to the design basis accidents will

not change and the accident mitigation

equipment will continue to function as

assumed in the design basis accident

analysis. Therefore, this activity does not

involve a significant reduction in a margin of

safety.

MP3 USQ 2

The proposed design change will allow the

MP3 RSST (or the MP3 NSST, if the MP3

RSST is unavailable) to provide the alternate

offsite source for MP2 to meet GDC 17

requirements. If the MP3 RSST is not

available, MP2 will need to credit the MP3

NSST as the alternate offsite power supply

for GDC 17 compliance. This will require the

13T breaker to be open to meet GDC 17

requirements. As a result, the probability of

a loss of offsite power at MP3 will increase.

A loss of offsite power at MP3 due to a

fault in the offsite distribution network is a

low probability event. When combined with

the expected frequency of removing the MP3

RSST from service of once per MP3 refueling

outage, the probability of a loss of offsite

power at MP3 when shut down as a result

of breaker 13T being open is low. In addition,

the MP3 shutdown risk program will

evaluate the impact of removing the MP3

RSST from service, and plan accordingly.

This will ensure the MP3 RSST will be

removed from service when the shutdown

risk is determined to be acceptably low. Also,

at least one MP3 EDG will be available, as

required by Technical Specifications, to

supply power to necessary loads if offsite

power is lost.

The plant response to the design basis

accidents will not change with breaker 13T

open, and the accident mitigation equipment

will continue to function as assumed in the

design basis accident analysis. Therefore, this

activity does not involve a significant

reduction in a margin of safety.

MP3 USQ 3

The connection of an additional 3 MVA

load for MP2 onto the MP3 electrical

distribution system will increase the short

circuit fault levels on the MP3 electrical

distribution system. If worst case conditions

are established, the additional contribution

from the MP2 loads would increase the fault

level to above the switchgear ratings. The

worst case short circuit conditions occur

with MP3 at full power, the switchyard at its

maximum voltage, an MP3 EDG paralleled to

the bus for surveillance testing, and the MP3

to MP2 cross-tie supplying 3 MVA of power

to MP2.

Although the fault levels could exceed the

MP3 switchgear and circuit breaker ratings,

it is not expected that the worst case

conditions will be frequently established. For

the majority of scenarios, this situation can

be avoided because the new design allows

the 3 MVA of power for MP2 to be supplied

by either MP3 bus 34A or 34B. Procedural

guidance will be provided to select the MP3

bus (34A or 34B) to supply power to MP2

that powers the opposite train from the train

associated with the MP3 EDG to be tested.

Even if the worst case plant configuration

cannot be avoided, there is a low probability

of event occurrence given the low probability

of a bus fault coincident with the short time

that the MP3 EDG would be paralleled to the

system. In addition, the event would only

affect one MP3 train of safety related

equipment. The other train will be available

to function as assumed to mitigate any

accidents that may occur.

The response of the plant and the operators

following a design basis event will not be

significantly different when using the MP3 to

MP2 4160 Volt cross-tie. The accident

mitigation equipment will continue to

function as assumed in the design basis

accident analysis. Therefore, this activity

does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut

NRC Section Chief: James W. Clifford.

Nuclear Management Company, LLC,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request:

November 18, 1999, as supplemented

August 7, 2000.

Description of amendment request:

The proposed amendment to the

Kewaunee Nuclear Power Plant

Technical Specifications requests

approval to increase the allowable

number of spent fuel assemblies stored

in the spent fuel pools.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated. In the analysis of the

safety issues concerning the expanded pool

canal storage capacity, the following

previously postulated accident scenarios

have been considered:

a. A spent fuel assembly drop in the Spent

Fuel Pool

b. Loss of Spent Fuel Pool cooling flow

c. A seismic event

The probability that any of the accidents in

the above list can occur is not significantly

65348 Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices

increased by the modification. The

probabilities of a seismic event or loss of SFP

cooling flow are not influenced by the

proposed changes. The probability of an

accidental fuel assembly drop is primarily

influenced by the methods used to lift and

move the fuel. The method of handling fuel

during the loading of the canal racks will be

the same as current fuel handling methods,

since the same equipment (i.e., Spent Fuel

Handling Crane) and the same procedural

guidance will be used. Since the methods

used to move fuel during normal operations

remain nearly the same as those used

previously, there is no significant increase in

the probability of an accident.

Accordingly, the proposed modification

does not involve a significant increase in the

probability of an accident previously

evaluated.

The consequences of the previously

postulated scenarios for an accidental drop of

a fuel assembly in the SFP have been reevaluated for the proposed change and were

determined to be bounded by the existing

analysis. The results show that the postulated

accident of a fuel assembly striking the top

of the storage racks will not distort the racks

sufficiently to impair their functionality. The

resulting structural damage to a falling

assembly and/or a stored assembly has been

determined to remain unchanged. The

minimum subcriticality margin, Keff less than

or equal to 0.95, will be maintained. The

structural damage to the SFP structure, pool

liner, and fuel assembly resulting from a fuel

assembly drop striking the pool floor or

another assembly located within the racks

remains unchanged. The resulting structural

damage to these items subsequent to this

event is not influenced by the proposed

changes.

The consequences of a loss of SFP cooling

have been evaluated and found to have no

increase. The concern with this accident is a

reduction of SFP water inventory from bulk

pool boiling resulting in uncovering fuel

assemblies. This situation would lead to fuel

failure and subsequent significant increase in

offsite dose. Loss of SFP cooling at Kewaunee

is mitigated by ensuring that a sufficient time

period exists between the loss of forced

cooling and uncovering fuel. This period of

time is compared against a reasonable period

to re-establish cooling or supply an

alternative water source (such as service

water, makeup water, or fire protection

system water). This evaluation included the

determination of the time to boil. The time

to boil represents the onset of loss of pool

water inventory and is commonly used as a

gage for establishing the comparison of

consequences before and after a refueling

project. The heat up rate in the SFP is a

nearly linear function of the fuel decay heat

load. The fuel decay heat load will increase

subsequent to the proposed changes because

of the increase in the number of stored

assemblies. The heat up rate established for

the limiting heat load conditions prior to the

addition of canal racking was 10.8°F per

hour. This would result in the pool

temperature increasing from the maximum

design temperature of 150°F to boiling in a

period of 5.7 hours. The heat up rate

established for the limiting heat load

conditions subsequent to the proposed

changes has been determined as 7.5°F per

hour. This would result in the pool

temperature increasing from the maximum

design temperature of 150°F to boiling in a

period of 8.4 hours.

This time to boil comparison was made for

limiting heat load conditions. However, the

end of this period of time does not represent

the onset of any significant increase in offsite

doses. As stated above, this consequence

would result after fuel is uncovered through

unchecked boiling and the resulting water

level drop of approximately 25 feet to the top

of the fuel storage racks. This depth is

conservative, since the top of active fuel is

below this level. Subsequent to the proposed

changes under limiting heat loads the

maximum boil-off rate will decrease to 40.9

gpm. This will result in the time lapse

between loss of SFP cooling and the

uncovering of the racks to increase to 48.5

hours.

As stated above, subsequent to racking the

canal, the time to boil after loss of forced

cooling in the most severe scenario was 8.4

hours. The ensuing rate of evaporative loss

would not result in the fuel being uncovered

until after an additional 40.1 hours. The

margin concerning time to boil will increase

due to the proposed modification. This

increased margin is due to the conservatism

of the assumptions previously used to

determine the heat load for this condition

(i.e. the assumption that all fuel assemblies

in the core were transferred to the pool

instantaneously). Also, another factor is that

the required incore hold time prior to fuel

movement will increase from 100 hours to

148 hours. Therefore, the calculated time to

boil in this most severe scenario will increase

subsequent to the proposed modification. In

the unlikely event that all pool cooling is

lost, sufficient time will continue to be

available subsequent to the proposed changes

for the operators to provide alternate means

of cooling (i.e., service water, makeup water,

or fire protection system water) before fuel is

uncovered. Therefore, the proposed changes

represent no increase in the consequences of

a loss of pool cooling.

The consequences of a design basis seismic

event are not increased. The consequences of

this accident are evaluated on the basis of

subsequent fuel damage or compromise of

the fuel storage or building configurations

leading to radiological or criticality concerns.

The canal racks have been analyzed and

found to properly function during seismic

motion. The stored fuel has been determined

to remain intact and the storage racks

maintain the fuel and fixed poison

configurations subsequent to a seismic event.

The structural capability of the pool and liner

will not be exceeded under the appropriate

combinations of dead weight, thermal, and

seismic loads. The SFP structure will remain

intact during a seismic event and will

continue to adequately support and protect

the fuel racks, stored fuel, and pool coolant.

Thus, the consequences of a seismic event

are not increased.

Therefore, it is concluded that the

proposed changes do not significantly

increase the probability or consequences of

any accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

During and following project

implementation, the spent fuel must be safely

stored, meeting the requirements 10 CFR 20

and 10 CFR 100. The installation activities as

well as the use of the canal racks must be

evaluated for the possibility of creating a new

or different kind of accident from any

previously analyzed.

This evaluation was done by reviewing the

differences between the analyses and

assumptions associated with the proposed

storage configuration (existing racks plus

canal racks) with those of the existing storage

configuration. The accident scenarios

associated with the existing racks were then

reviewed to determine if the possibility of a

new type of accident would be introduced by

the implementation of the proposed

modification. Also, the installation activities

were evaluated.

Due to the proposed changes, the following

events were considered as the only events

which might represent a new or different

kind of accident:

a. An accidental drop of a rack module into

the pool during construction activity.

b. Fuel assembly mispositioning accident

in the canal.

A construction accident resulting in a rack

drop is an unlikely event. A rack-lifting rig

will be used to lift and suspend the racks

using the existing Fuel Handling Building

crane. The crane, hoists and lifting rig have

been or will be designed, tested and

inspected to ensure that they will properly

lift and transport the new racks. The

postulated rack drop event is commonly

referred to as a ‘‘heavy load drop’’ over the

pools. Heavy loads will not be allowed to

travel over any racks containing fuel

assemblies. The area of concern represented

by this event is that the pool structure will

be compromised leading to loss of

moderator/coolant. The question of a new or

different type of event is answered by

determining whether heavy load drops of this

type have been considered previously. The

last phase of the re-racking of the SFPs at

KNP was completed in 1987. The rack

modules transported during installation of

the pool re-racking project were significantly

larger than those associated with the

proposed canal rack project. Also, the KNP

Technical Specifications state that

‘‘Placement of additional fuel storage racks is

permitted, however, these racks must not

traverse directly above spent fuel stored in

the pools’’. All movements of heavy loads

will be strictly controlled by procedure, will

follow approved safe load paths, and will be

performed by qualified personnel. Therefore,

the rack drop does not represent a new or

different kind of accident.

Fuel assembly mispositioning in the canal

is an unlikely event since assembly

placement will be administratively

controlled. The administrative controls will

include positive, visual identification of each

fuel assembly to be stored in the canal racks.

The mispositioning event for the canal racks

represents a change from the previously

analyzed condition, since Kewaunee

currently has no restriction on fuel storage.

Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices 65349

However, the mispositioning event in the

canal does not represent a new or different

kind of accident. A fuel assembly

mispositioning event has been previously

analyzed for the existing racks and was found

to be acceptable. The new mispositioning

event for the canal racks was evaluated using

similar techniques with similar acceptance

criteria and was also shown to be acceptable.

Therefore, since both events represent the

mispositioning of a fuel assembly, the

mispositioning of an assembly in the canal

racks is not considered to represent a new or

different kind of accident.

Under the extremely unlikely conditions of

a mispositioning event involving the

placement of a fresh (un-burned) fuel

assembly in the canal racks, the boron in the

pool water will maintain the required

subcriticality margin. Only a small fraction of

the minimum pool boron concentration

required by the KNPP Technical

Specifications would be required for this

condition. Under the postulated accident

condition of a total loss of boron from the

pool water, proper subcriticality margin

would exist. The multiple accident scenario

of the extremely unlikely mispositioning

event combined with a postulated total loss

of pool boron is a non-credible condition.

Therefore, although credit is taken for a small

amount of boron in the pool water for one

accident condition, a decrease of pool boron

concentration is not a new or different kind

of accident from any previously analyzed.

The proposed change does not involve the

modification of any equipment credited in

the mitigation of the design basis accidents.

The proposed change meets all applicable

requirements for the safe storage of spent fuel

at KNP. Therefore, the potential for a new or

previously unanalyzed accident is not

created by the implementation of this

modification.

The refueling canal does not have floor

drains. Therefore, draining the refueling

canal through a floor drain is not a postulated

event.

3. Involve a significant reduction in the

margin of safety.

The function of the SFP is to store the fuel

assemblies in a subcritical and coolable

configuration through all environmental and

abnormal loadings, such as an earthquake or

fuel assembly drop. The new rack design

must meet all applicable requirements for

safe storage and be functionally compatible

with the SFP.

WPSC has addressed the safety issues

related to the expanded pool storage capacity

in the following areas:

a. Material, mechanical and structural

considerations

b. Nuclear criticality

c. Thermal-hydraulic and pool cooling

The mechanical, material, and structural

designs of the new racks have been reviewed

in accordance with the applicable provisions

of the NRC Guidance entitled, ‘‘Review and

Acceptance of Spent Fuel Storage and

Handling Applications’’. The rack materials

used are compatible with the spent fuel

assemblies and the SFP environment. The

design of the new racks preserves the proper

margin of safety during abnormal loads such

as a dropped assembly and tensile loads from

a stuck assembly. It has been shown that

such loads will not invalidate the mechanical

design and material selection to safely store

fuel in a coolable and subcritical

configuration.

The methodology used in the criticality

analysis of the expanded SFP meets the

appropriate NRC guidelines and ANSI

standards. The margin of safety for

subcriticality is maintained by having the

neutron multiplication factor equal to, or less

than, 0.95 under all accident conditions,

including uncertainties. This criterion is the

same as that used previously to establish

criticality safety evaluation acceptance and

remains satisfied for all analyzed accidents.

Therefore, the accepted margin of safety

remains the same.

The thermal-hydraulic and cooling

evaluation of the pool demonstrated that the

pool can be maintained below the specified

thermal limits under all required design

conditions. The pool temperature will not

exceed 140°F during the failure of a cooling

pump under normal offload conditions. The

pool temperature will remain below the

maximum design temperature of 150°F under

maximum heat load conditions. The

maximum local water temperature in the hot

channel will remain below the boiling point.

The fuel will not undergo any significant

heat up after an accidental drop of a fuel

assembly on top of the rack blocking the flow

path. Following a loss of cooling to the pool

sufficient time exists (48.5 hours for the

limiting heat load) for the operators to

intervene and line up alternate cooling paths

and the means of inventory makeup before

the onset of pool boiling. The thermal limit

specified for the evaluations performed to

support the proposed change is the same as

was used in the previous evaluation.

Therefore, the accepted margin of safety

remains the same.

Thus, it is concluded that the changes do

not involve a significant reduction in the

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: Bradley D.

Jackson, Esq., Foley and Lardner, P.O.

Box 1497, Madison, WI 53701–1497.

NRC Section Chief: Claudia M. Craig.

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:

September 14, 2000.

Description of amendment request:

The proposed changes will delete

Technical Specification (TS) Section

6.2.3 and revise TS Sections 6.3.1 and

6.5.2.8 to remove the references to the

Independent Safety Engineering Group

(ISEG).

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:

South Carolina Electric & Gas Company

(SCE&G) has evaluated the proposed changes

to the VCSNS TS described above against the

significant Hazards Criteria of 10 CFR 50.92

and has determined that the changes do not

involve any significant hazard. The following

is provided in support of this conclusion.

1. Does the proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed amendment is a

programmatic and administrative change to

delete the ISEG. The requirement for an ISEG

is located within the Administrative Section

of TS (Section 6.0). The ISEG functions in an

oversight role and, as such, performs no

actions that would affect any precursors to

any accident previously evaluated. This

change does not physically alter safetyrelated systems, nor does it affect the way in

which safety-related systems perform their

functions. However, the independent

oversight functions and qualification

requirements stated in TS are being

performed by qualified personnel in other

departments of the VCSNS organization. This

assures that the functions and qualification

requirements delineated for the ISEG are

being met. Because the design of the facility

and system operating parameters are not

being changed, and the oversight functions

are performed by other departments, the

proposed amendment does not involve an

increase in the probability or consequences

of any accident previously evaluated.

2. Does the proposed change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed amendment to delete the

ISEG is a programmatic and administrative

change. There are no physical alterations to

safety-related systems and no changes to the

functions of any safety-related systems. The

independent oversight functions assigned to

the ISEG are addressed by other departments

within the VCSNS organization. Personnel in

those departments meet the qualifications

required of the ISEG. This assures that the

oversight functions and qualification

requirements delineated for the ISEG are

being met. Oversight functions that are

performed do not in themselves lead to

activities that are considered as accident

precursors or initiators. Because this change

does not alter the design of the facility and

system operating parameters are not being

changed, and the oversight functions and

qualification requirements of the ISEG are

being met by other departments, the

proposed amendment 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?

The proposed amendment is a

programmatic and administrative change to

65350 Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices

delete the ISEG. This proposed change

involves utilizing appropriately trained and

qualified personnel in other departments to

perform the oversight functions previously

assigned to the ISEG. No physical alterations

to safety-related systems and no changes to

the functions of any safety-related systems

are being made. The use of personnel in other

departments to perform the oversight

functions of the ISEG will provide assurance

that plant operations continue to be

conducted in a safe manner. These oversight

functions will be performed by personnel

that maintain their independence from line

operations. Because the design of the facility

and system operating parameters are not

being changed, and the oversight functions of

the ISEG are appropriately addressed by

other departments and will continue to be

performed, the proposed amendment 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: Thomas G.

Eppink, South Carolina Electric & Gas

Company, Post Office Box 764,

Columbia, South Carolina 29218

NRC Section Chief: Richard L. Emch,

Jr.

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Units 1 and 2, Somervell

County, Texas

Date of amendment request:

September 6, 2000.

Brief description of amendments: The

proposed amendment would change

Comanche Peak Steam Electric Station

(CPSES), Units 1 and 2, Technical

Specification (TS) 5.5.9, ‘‘Steam

Generator Tube Surveillance Program,’’

to permit installation of a laser welded

tube sleeve as an alternative to plugging

defective steam generator tubes. TS

5.5.10, ‘‘Steam Generator Tube

Inspection Report,’’ would be revised to

address reporting requirements for

repaired tubes. Also, an editorial

correction is proposed to Table 5.5–2.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

The tubesheet and/or tube support plate

intersection laser welded sleeve

configurations [were] designed and analyzed

in accordance with the requirements of the

ASME Code [American Society of

Mechanical Engineers Boiler and Pressure

Vessel Code]. Fatigue and stress analyses of

the sleeved tube assemblies produced

acceptable results. Additionally, mechanical

testing for the full length tubesheet sleeves

has shown that the structural strength of

Alloy 690 sleeves under normal, faulted, and

upset conditions is within acceptable limits.

Leakage testing for these same 3⁄4 inch tube

sleeves has demonstrated that primary to

secondary leakage is not expected during any

plant conditions. Similar results are

anticipated for the lower joints of elevated

tubesheet sleeves. Confirmatory mechanical

and leak testing will be conducted

supporting the installation of elevated

tubesheet sleeves at CPSES, Unit 1.

The hypothetical consequences of failure

of a sleeve would be bounded by the current

steam generator tube rupture analysis

included in the Comanche Peak Steam

Electric Station (CPSES) Final Safety

Analysis Report (FSAR). Due to the slight

reduction diameter caused by the sleeve wall

thickness, it is expected that primary coolant

release rates would be slightly less than

assumed for the steam generator tube rupture

analysis (depending on the break location),

and therefore, would result in lower total

primary fluid mass release to the secondary

system. Combinations of tubesheet sleeves

and tube support plate sleeves would reduce

the primary fluid flow through the sleeved

tube assembly due to the series of diameter

reductions the fluid would have to pass on

its way to the break area. The overall effect

would be reduced steam generator tube

rupture release rates. The proposed

Technical Specification change to support

the installation of Alloy 690 laser welded

sleeves does not adversely impact any other

previously evaluated design basis accident or

the results of (LOCA) [loss-of-coolant

accident] and non-LOCA accident analyses

for the current Technical Specification

minimum RCS [reactor coolant system] flow

rate.

Conformance of the sleeve design with the

applicable sections of the ASME Code and

the successful completion of the leakage and

mechanical tests (for the lower sleeve joint

for the elevated tubesheet sleeves (ETS),

support the conclusion that the installation

of laser welded tube sleeves will not increase

the probability or consequences of an

accident previously evaluated. Depending

upon the break location for a postulated

steam generator tube rupture event,

implementation of tube sleeving could act to

reduce the radiological consequences to the

public due to reduced flow rate through a

sleeved tube compared [to] a non-sleeved

tube based on the restriction afforded by the

sleeve wall thickness.

The editorial correction [to] Technical

Specification (TS) Table 5.5–2 is

typographical in nature and does not require

additional evaluation. Therefore, the

proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

Implementation of the laser welded

sleeving (LWS) will not introduce significant

or adverse changes to the plant design basis.

Stress and fatigue analysis of the repair has

shown the ASME Code minimum stress

values are not exceeded. Implementation of

laser welded sleeving restores the overall

tube bundle structural and leakage integrity

to a level consistent to that of the originally

supplied tubing during all plant conditions.

Any hypothetical accident as a result of

potential tube or sleeve degradation in the

repaired portion of the tube is bounded by

the existing tube rupture accident analysis.

Finally, through the results obtained from the

extensive testing and qualification program,

the possibility of a common-mode failure,

such as multiple simultaneous steam

generator tube failures, is not credible.

Therefore, it is concluded that the proposed

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

The editorial correction [to] TS Table 5.5–

2 is typographical in nature and does not

require additional evaluation. Therefore, the

proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

Response: No.

The laser welded sleeving repair of

degraded steam generated tubes as identified

in References 1 and 2 was shown by analysis

to restore the integrity of the tube bundle

consistent with its original design basis

condition. The safety factors used in the

design of sleeves for the repair of degraded

tubes are consistent with the safety factors in

the ASME Boiler and Pressure Vessel Code

used in steam generator design. The design

of the full length tubesheet sleeve lower

joints for the 3⁄4 inch tube sleeves (roll-first

installation sequence) were verified by

testing to preclude pullout and primary-tosecondary leakage during normal and

postulated accident conditions. The

qualification of the lower joint of the TSS

[tube support sleeve], ETS and the full length

tubesheet sleeves (FLTS) (roll-last

installation sequence) will be confirmed at

the time of the sleeving outage. Since the

installed sleeve represents a portion of the

pressure boundary, a baseline inspection of

these areas is required prior to operation with

the sleeves installed. The portions of the

installed sleeves assembly which represent

the reactor coolant pressure boundary can be

monitored for the initiation and progression

of sleeve/tube wall degradation, thus

satisfying the recommendations of Regulatory

Guide 1.83, Rev. 1. The portion of the tube

bridged by the sleeve joints is effectively

removed from the pressure boundary, and the

sleeve then forms the new pressure

boundary. The areas of the sleeved tube

assembly which require inspection are

defined in WCAP–13698, Rev. 3.

EPRI [Electric Power Research Institute]

qualified eddy current techniques will be

used for the detection of tube degradation in

3⁄4 inch laser welded sleeved tubes. Alternate

Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices 65351

inspection techniques may be used as they

become available, as long as it can be

demonstrated that the technique used

provides the same degree or greater degree of

inspection rigor.

The effect of sleeving on the design

transients and accident analyses were

reviewed and found to remain valid up to the

level of steam generator tube plugging

consistent with the minimum reactor flow

rate as specified in Technical Specification

3.4.1. Continued compliance with the RCS

flow limits of Technical Specification 3.4.1 is

assured through precision flow

measurements.

Because all relevant safety analyses were

reviewed and found to remain valid, and

because the appropriate design margins are

maintained through compliance with the

relevant ASME Code requirements, it is

concluded that the proposed change does not

involve a significant reduction in a margin of

safety.

The editorial correction [to] TS Table 5.5–

2 is typographical in nature and does not

require additional evaluation. The

confirming modifications to the reporting

requirements of TS 5.6.10 are administrative

only. Therefore, these proposed changes do

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Units 1 and 2, Somervell

County, Texas

Date of amendment request:

September 15, 2000.

Brief description of amendments: The

proposed change replaces the general

references currently provided in

Technical Specification 5.6.6 for

determining the reactor coolant system

pressure and temperature limits with

the requirement that the Pressure/

Temperature Limits and Low

Temperature Overpressure Protection

System Setpoints shall not be revised

without prior U.S. Nuclear Regulatory

Commission approval.

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. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

The addition of this requirement to the

Technical Specifications is administrative

and has no impact on accident initiation or

mitigation. Therefore, the proposed changes

do not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

The addition of this requirement to the

Technical Specifications is administrative

and cannot initiate an accident. Therefore,

the proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

Response: No.

The addition of this requirement to the

Technical Specifications is administrative

and has no impact on accident initiation or

mitigation. Therefore the proposed change

does not involve a reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

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:

September 22, 2000.

Description of amendment request:

The proposed changes would remove

obsolete license conditions from the

Operating Licenses (OLs) and

implement associated changes to the

Technical Specifications (TS) and Bases.

The proposed changes include removal

of license conditions associated with

completed facility modifications

(including the Steam Generator (SG)

Repair Program, as well as support

modifications related to Leak-BeforeBreak Technology); removal of

superseded license conditions

(addressing security); relocation of

secondary water chemistry monitoring

program requirements into the TS;

removal of expired license conditions

and TS (addressing service water piping

restoration); and editorial changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Virginia Electric and Power Company has

reviewed the requirements of 10 CFR 50.92

as they relate to the proposed administrative

change to the Operating Licenses, DPR–32

and DPR–37, and the associated Technical

Specifications for Surry Units 1 and 2 and

determined that a significant hazards

consideration is not involved. The proposed

administrative change to the Surry Operating

Licenses and associated Technical

Specifications makes minor editorial

corrections, relocates one license condition

to Appendix A of the license, and removes

outdated, superceded or otherwise nonapplicable license conditions and Technical

Specifications requirements and provides a

license document that is directly applicable

to the current plant licensing and design

bases. There is no safety significance

associated with this proposed change since

the change does not alter any currently

applicable Operating License requirements.

Accordingly, the current Surry licensing and

design bases are unchanged. In support of

this conclusion, the following evaluation is

provided.

Criterion 1—The proposed license

amendment does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed change is administrative

(and in part editorial) in nature and neither

station operations nor design are affected by

the change. The removal of license

conditions and associated Technical

Specifications regarding superceded (OL

Sections 3.H and 3.L) or expired (OL Section

3.0, TS Table 3.7–2, and TS 3.14)

requirements has no impact on plant

operations since the requirements no longer

have a legitimate means of being applied.

The relocation within the Operating License

of the requirement to have a secondary water

chemistry monitoring program (OL Section

3.K) to new Section 6.4.P of the Technical

Specifications does not alter the program or

its implementation. The impact of the

replacement SGs at Surry and the re-design

of the SG and RCP [reactor coolant pump]

supports on the previously evaluated

accidents was performed, approved and

documented by the issuance of the license

amendments for OL Sections 3. G and 3.M.

Removal of these license conditions which

refer to completed work and a design and

licensing bases that is documented in the

UFSAR [updated final safety analysis report]

also does not alter station operation or the

design of the affected components. The

proposed change is within the current design

and licensing bases of the facility. This

change does not affect the initiators of

analyzed events nor the assumed mitigation

of accident or transient events. Analyzed

events are initiated by the failure of plant

structures, systems, or components. This

change does not impact the condition or

performance of these structures, systems or

components. Consequences of analyzed

events are the result of the plant being

65352 Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices

operated within assumed parameters at the

onset of any event, and the successful

functioning of at least one train or division

of the equipment credited with mitigating the

event. These changes do not impact the

capability of the credited equipment to

perform, nor is there any change in the

likelihood that credited equipment will fail

to perform. As a result, the proposed change

to the Surry Operating Licenses and

Technical Specifications does not involve

any increase in the probability or the

consequences of any accident or malfunction

of equipment important to safety previously

evaluated.

Criterion 2—The proposed license

amendment does not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed change is administrative

(and in part editorial) in nature. The license

conditions and Technical Specifications that

are being removed or relocated by this

proposed change do not impact station

operations or station equipment in any

manner. The proposed change does not

involve a physical alteration of the plant, nor

a change in the methods used to respond to

plant transients that has not been previously

analyzed. No new or different equipment is

being installed and no installed equipment is

being removed or operated in a different

manner. There is no alteration to the

parameters within which the plant is

normally operated or in the setpoints, which

initiate protective or mitigative actions.

Consequently, no new failure modes are

introduced and the proposed change to the

Surry Operating Licenses and Technical

Specifications does not create the possibility

of a new or different kind of accident or

malfunction of equipment important to safety

from any previously evaluated.

Criterion 3—The proposed license

amendment does not involve a significant

reduction in a margin of safety.

The proposed change is administrative

(and in part editorial) in nature and neither

station operations nor design are affected by

the change. Margin of safety is established

through the design of the plant structures,

systems and components, the parameters

within which the plant is operated, and the

establishment of the setpoints for the

actuation of equipment relied upon to

respond to an event. Since station operations

are not affected by the proposed change, the

change does not impact the condition or

performance of structures, systems or

components relied upon for accident

mitigation or any safety analysis

assumptions. Therefore, the proposed change

to the Surry Operating Licenses and

Technical Specifications does not involve a

reduction in any margin of safety described

in the bases of the Technical Specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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., Hinton and Williams, Riverfront

Plaza, East Tower, 951 E. Byrd Street,

Richmond, Virginia 23219.

NRC Section Chief: Richard L. Emch.

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.

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:

September 22, 2000 (PCN–520).

Brief Description of amendment

requests: The proposed amendments

would revise the San Onofre Units 2

and 3 technical specifications (TSs)

applicable in shutdown MODES relating

to positive reactivity additions. For a

summary of specific proposed TS

changes, see Tables 1 and 2 of the

licensee’s application dated September

22, 2000. The licensee’s proposal

generally conforms to industry

Technical Specification Task Force

(TSTF) TSTF–286, Revision 2.

Date of publication of individual

notice in Federal Register: October 13,

2000 (65 FR 60984).

Expiration date of individual notice:

November 13, 2000.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, located at One White Flint North,

11555 Rockville Pike (first floor),

Rockville, Maryland 20852. Publicly

available records will be accessible

electronically from the ADAMS Public

Library component on the NRC Web

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

Reading Room).

AmerGen Energy Company, LLC, Docket

No. 50–461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of application for amendment:

July 14, 2000.

Brief description of amendment: The

amendment slightly reduces the

required minimum reactor cavity water

level.

Date of issuance: October 12, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 133.

Facility Operating License No. NPF–

62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51347).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 12,

2000.

No significant hazards consideration

comments received: No.

Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices 65353

AmerGen Energy Company, LLC, Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit 1, Dauphin County,

Pennsylvania

Date of application for amendment:

November 30, 1999, as supplemented

August 11, and September 14, 2000.

Brief description of amendment: The

amendment revised the test standard for

laboratory testing of activated charcoal

to test in accordance with the ASTM

(American Society for Testing and

Materials) D3803–1989 standard in

response to Generic Letter 99–02,

‘‘Laboratory Testing of Nuclear-Grade

Activated Charcoal.’’

Date of issuance: October 18, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 226.

Facility Operating License No. DPR–

50. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 8, 2000 (65 FR 12287).

The August 11, and September 14,

2000, letters provided clarifying

information that did not change the

initial no significant hazards

consideration determination and did not

expand the scope of the original notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 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 31, 2000.

Brief description of amendments: The

amendments deleted the requirements

related to the shorting links from

Technical Specification (TS) Sections 3/

4.3.1, ‘‘Reactor Protection System

Instrumentation;’’ 3/4.9.2, ‘‘Refueling

Operations Instrumentation;’’ and 3/

4.10.3, ‘‘Shutdown Margin

Demonstrations;’’ and increased the

required signal-to-noise ratio for the

source range monitor in TS Sections 3/

4.9.2 and 3/4.3.7.6, ‘‘Source Range

Monitors.’’

Date of issuance: October 10, 2000.

Effective date: Immediately, to be

implemented within 60 days.

Amendment Nos.: 142 and 128.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 9, 2000 (65 FR 48745).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 10,

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 1, 2000, as supplemented by letter

dated August 11, 2000.

Brief description of amendments: The

amendments revise Technical

Specification 3/4.8.1, ‘‘A. C. Sources—

Operating,’’ to permit functional testing

of the emergency diesel generators to be

performed during power operation.

Date of issuance: October 16, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment Nos.: 143 and 129.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 14, 2000 (65 FR 37423).

The supplemental information

contained clarifying information and

did not change the initial no significant

hazards consideration determination

and did not expand the scope of the

original Federal Register notice. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated October 16, 2000.

No significant hazards consideration

comments received: No.

Commonwealth Edison Company,

Docket Nos. 50–254 and 50–265, Quad

Cities Nuclear Power Station, Units 1

and 2, Rock Island County, Illinois

Date of application for amendments:

December 30, 1999; as supplemented on

August 3, 2000.

Brief description of amendments: The

amendments revised the technical

specifications to: (1) Remove the Main

Steamline Radiation Monitor (MSLRM)

scram and main steamline isolation

functions, and (2) add a new

requirement for the MSLRM mechanical

vacuum pump trip function.

Date of issuance: October 13, 2000.

Effective date: Immediately, to be

implemented within 120 days.

Amendment Nos.: 196 and 192.

Facility Operating License Nos. DPR–

29 and DPR–30: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 9, 2000 (65 FR 48747).

The August 3, 2000, supplement

provided additional information that

did not change the scope of the original

proposed no significant hazards

findings.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 13,

2000.

No significant hazards consideration

comments received: No.

Consolidated Edison Company of New

York, Inc., Docket No. 50–003, Indian

Point Nuclear Generating Station, Unit

1

Date of amendment request: February

14, 2000.

Brief description of amendment: The

amendment change to Indian Point

Nuclear Generating Station, Unit 1,

revised Technical Specification Sections

2.10.2, 3.1.2, 3.2.1, 4.1.8.1, and 4.1.8.1.b.

Specifically, Sections 3.1.2, 3.2.1, and

4.1.8.1.b, make organizational title

changes that are administrative in

nature and reflect a streamlining of the

Consolidated Edison Company of New

York, Inc.’’s management structure,

Section 4.1.8.1 is changed to reflect the

renumbering of 10 CFR Part 20, and a

footnote was moved from Section 2.11

to Section 2.10.2 to improve the clarity

of the Technical Specification since it

pertains to text in subsection 2.10.2.4.

Date of issuance: October 12, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No: 48.

Facility Operating License No. DPR–5:

The amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 5, 2000 (65 FR 17912).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 12,

2000.

No significant hazards consideration

comments received: No.

Entergy Nuclear Generation Company,

Docket No. 50–293, Pilgrim Nuclear

Power Station, Plymouth County,

Massachusetts

Date of application for amendment:

November 22, 1999.

Brief description of amendment: The

amendment authorized a change to the

Pilgrim Nuclear Power Station Updated

Final Safety Analysis Report (UFSAR).

The change involves the use of

containment overpressure to ensure

sufficient net positive suction head for

the emergency core cooling system

pumps following a loss-of-coolant

accident.

Date of issuance: October 6, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 day from the date of issuance.

Amendment No.: 185.

65354 Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices

Facility Operating License No. DPR–

35: Amendment revised the UFSAR.

Date of initial notice in Federal

Register: March 8, 2000 (65 FR 12290).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 6, 2000.

No significant hazards consideration

comments received: No.

Entergy Nuclear Generation Company,

Docket No. 50–293, Pilgrim Nuclear

Power Station, Plymouth County,

Massachusetts

Date of application for amendment:

June 16, 1999, as supplemented on May

4 and July 10, 2000.

Brief description of amendment: The

requested changes incorporate

Technical Specification (TS) changes to

comply with the operating requirements

derived from GE Report, NEDO–21231,

‘‘Banked Position Withdrawal Sequence

(BPWS)’’, dated January 1977, as

referenced in NEDE–24011–P–A. The

TS changes incorporate Specifications

and Actions based upon the plantspecific CRDA and BPWS for 20% rated

thermal power (RTP) and 280 cal/gram

peak fuel enthalpy. The TS changes also

include changes to the control rod

worth limits to resolve Licensee Event

Report (LER) 98–006–00, dated April 30,

1998, and its supplemental LER 98–

006–01, dated August 27, 1988.

Date of issuance: October 16, 2000.

Effective date: As of the date of

issuance, and shall be implemented

within 30 days.

Amendment No.: 186.

Facility Operating License No. DPR–

35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51350).

The May 4 and July 10, 2000, letters

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination or expand the scope of

the application as published in the

Federal Register.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 16,

2000.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2,

Pope County, Arkansas

Date of application for amendment:

August 18, 1999, as supplemented by

letters dated June 29, July 19, and

August 9, 2000.

Brief description of amendment: The

amendment revised Technical

Specification 4.4.5, ‘‘Steam Generators,’’

to note that the requirements for

inservice inspection do not apply

during the steam generator replacement

outage (2R14), to revise the requirement

for tube inspection to mean an

inspection from tube end (cold leg side)

to tube end (hot leg side), to delete

inspection requirements associated with

steam generator tube sleeving and repair

limits, to revise the preservice

inspection requirements on when the

hydrostatic test and the eddy current

inspection of the tubes would be

performed, and to revise the reporting

frequency of the results of steam

generator tube inspections to within 12

months following completion of the

inservice inspection. Related changes to

the Bases were also made.

Date of issuance: October 4, 2000.

Effective date: As of the date of

issuance to be implemented prior to

startup from the 2R14 refueling outage.

Amendment No.: 223.

Facility Operating License No. NPF–6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 23, 2000 (65 FR 9005).

The application was renoticed on

August 23, 2000 (65 FR 51353).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 4, 2000.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

Company, et al., Docket No. 50–412,

Beaver Valley Power Station, Unit 2,

Shippingport, Pennsylvania

Date of application for amendment:

September 1, 2000.

Brief description of amendment: The

amendment revised certain 18-month

surveillance requirements in the

technical specifications by eliminating

the condition that testing be conducted

during shutdown, or during cold

shutdown or refueling mode. The

affected systems are emergency core

cooling system, containment

depressurization and cooling system,

chemical addition system, and

containment isolation valve system.

Date of issuance: October 13, 2000.

Effective date: As of date of issuance

and shall be implemented within 60

days.

Amendment No: 118.

Facility Operating License No. NPF–

73. Amendment revised the technical

specifications.

Date of initial notice in Federal

Register: September 12, 2000 (65 FR

55056).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 13,

2000.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

Company, Docket No. 50–440, Perry

Nuclear Power Plant, Unit 1, Lake

County, Ohio

Date of application for amendment:

November 1, 1999, and as supplemented

by letter dated May 22, 2000.

Brief description of amendment:

Consistent with the guidance of Generic

Letter 99–02, ‘‘Laboratory Testing of

Nuclear Grade Activated Charcoal,’’ this

amendment modifies existing Technical

Specification 5.5.7, ‘‘Ventilation Filter

Testing Program (VFTP),’’ to reference

ASTM D3803—1989, ‘‘Standard Test

Method for Nuclear-Grade Activated

Carbon.’’ The amendment also

incorporates the suggested safety factor

for charcoal filter efficiency regarding

methyl iodide penetration.

Date of issuance: October 12, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 90 days.

Amendment No.: 117.

Facility Operating License No. NPF–

58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 15, 1999 (64 FR

70088).

The supplemental information

contained clarifying information and

did not change the initial no significant

hazards consideration determination

and did not expand the scope of the

original Federal Register notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 12,

2000.

No significant hazards consideration

comments received: No.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Units 1 and 2,

Berrien County, Michigan

Date of application for amendments:

May 30, 2000.

Brief description of amendments: The

amendments would make changes to

several Technical Specifications to

reflect implementation of the revised 10

CFR Part 20, ‘‘Standards for Protection

Against Radiation.’’

Date of issuance: October 10, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 45 days.

Amendment Nos.: 245 and 226.

Facility Operating License Nos. DPR–

58 and DPR–74: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: August 9, 2000 (65 FR 48752).

Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices 65355

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 10,

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: April 28,

2000.

Description of amendment request:

The amendment implements the

Relaxed Axial Offset Control (RAOC)

operating strategy in support of the use

of upgraded Westinghouse fuel with

Intermediate Flow Mixers.

Date of issuance: October 6, 2000.

Effective date: As of its date of

issuance, and shall be implemented at

commencement of Cycle 8 operation

(scheduled for November 2000).

Amendment No.: 76.

Facility Operating License No. NPF–

86: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 31, 2000 (65 FR 34747).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 6, 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:

April 19, 2000.

Brief description of amendment: This

amendment modifies Technical

Specification (TS) 3.7.1.5, ‘‘Plant

System—Main Steam Line Isolation

Valves.’’ Specifically, the change

removes the requirement to perform

partial stroke testing of the main steam

line isolation valves during power

operation, modifies the TS wording for

clarity, combines two surveillance

requirements into one, and modifies the

associated TS Bases for consistency.

Date of issuance: October 19, 2000.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 185.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 23, 2000 (65 FR 51360).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 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:

December 15, 1999, as supplemented

August 24, 2000.

Brief description of amendments: The

amendments removed TS Table 3.6.3–1

‘‘Primary Containment Isolation

Valves,’’ and references to the TS Table

from the TSs and relocated the

information from the TS Table to the

Technical Requirements Manual. In

addition, an administrative change was

made which deleted references to TS

Tables 3.6.5.2.1–1 and 3.6.5.2.2–1.

Date of issuance: October 18, 2000.

Effective date: As of date of issuance

and shall be implemented within 30

days.

Amendment Nos.: 146 and 107.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 8, 2000 (65 FR 12294).

The August 24, 2000, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination or expand the scope of

the original Federal Register Notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 18,

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 3, 2000.

Brief description of amendment: The

amendment revises Technical

Specification Table 3.2–7 by changing

the reactor water level setpoint for the

anticipated transient without scram, the

recirculation pump trip function, and

the alternate rod insertion function.

Date of issuance: October 10, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 264.

Facility Operating License No. DPR–

59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 22, 2000 (65 FR 15383).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 10,

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:

December 20, 1999, as supplemented

February 4, 2000.

Brief description of amendment: The

amendment changes the Main Steam

Isolation Valve closure scram trip level

setting from ≤10 percent to ≤15 percent

valve closure.

Date of issuance: October 10, 2000.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 265.

Facility Operating License No. DPR–

59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 9, 2000 (65 FR 6410).

The February 4, 2000, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination and did not expand the

amendment beyond the scope of the

original notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 10,

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:

July 21, 2000

Brief description of amendment: The

amendment revises the Technical

Specifications to remove the Control

Room Emergency Air Treatment System

Actuation Instrumentation operability

during Modes 5 and 6 except during

core alterations and fuel movement

based on the control room dose

calculations.

Date of issuance: October 10, 2000.

Effective date: October 10, 2000.

Amendment No.: 78.

Facility Operating License No. DPR–

18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 9, 2000 (65 FR 48757).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated October 10,

2000.

No significant hazards consideration

comments received: No.

65356 Federal Register / Vol. 65, No. 212 / Wednesday, November 1, 2000 / Notices

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 application for amendments:

February 23, 2000, as supplemented

July 7, 2000.

Brief description of amendments: The

amendments revise the licensing basis

for San Onofre Nuclear Generating

Station (SONGS) Units 2 and 3

regarding the methodology for

measuring the reactivity worth of

control element assembly (CEA) groups

during low power physics testing

following a refueling. The amendments

allow measuring the worth of

approximately three-fourths of the fulllength CEA groups each refueling cycle

rather than the present methodology,

which measures the worth of all fulllength CEA groups each refueling cycle.

Date of issuance: October 10, 2000.

Effective date: October 10, 2000.

Implementation includes incorporation

of the changes into the Updated Final

Safety Analysis Report (UFSAR) at the

next update of the UFSAR in

accordance with the schedule in 10 CFR

50.71(e).

Amendment Nos.: Unit 2–173; Unit

3–164.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

authorized revision of the UFSAR

Section 4.2.1.5.2, CEA Performance

Testing.

Date of initial notice in Federal

Register: March 22, 2000 (65 FR 15385).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated October 10,

2000.

No significant hazards consideration

comments received: No.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

September 8, 2000.

Brief description of amendment: The

amendment revises the Callaway

Technical Specifications (TS) to

annotate the frequency for Surveillance

Requirement (SR) 3.5.2.5 that

verification of the automatic closure

function of the residual heat removal

pump suction Valve BNHV8812A shall

be performed prior to startup from the

first shutdown to Mode 5 (cold

shutdown) occurring after September 8,

2000, but no later than June 1, 2001. The

next refueling outage is scheduled for

April 2001. This amendment defers the

test of the automatic closure function

until the next plant shut down to cold

shutdown.

Date of issuance: October 6, 2000.

Effective date: October 6, 2000, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 140.

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes (65 FR 56943 dated

September 20, 2000). The notice

provided an opportunity to submit

comments on the Commission’s

proposed no significant hazards

consideration determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by October 20,

2000, but indicated that if the

Commission makes a final no significant

hazards consideration determination

any such hearing would take place after

issuance of the amendment.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated October 6, 2000.

Attorney for licensee: John O’Neill,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Stephen Dembek.

Dated at Rockville, Maryland, this 25th day

of October 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 00–27938 Filed 10–31–00; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

[Rel. No. IC–24697; 812–11786]

PMC Capital, Inc., et al.; Notice of

Application

October 25, 2000.

AGENCY: Securities and Exchange

Commission (‘‘Commission’’ or ‘‘SEC’’).

ACTION: Notice of application for an

order under sections 6(c) and 57(c) of

the Investment Company Act of 1940

(the ‘‘Act’’) for an exemption from

sections 57(a)(1) and 57(a)(2) of the Act,

and under section 57(i) of the Act and

rule 17d–1 under the Act authorizing

certain joint transactions otherwise

prohibited by section 57(a)(4) of the Act.

SUMMARY OF APPLICATION: Applicants

request an order that would permit (1)

a business development company

(‘‘BDC’’) to engage in a loan origination

agreement with an affiliated real estate

investment trust, (2) investment

management agreements between

subsidiaries of the BDC and the real

estate investment trust, and (3) the

establishment of special purpose

entities owned by the BDC and the real

estate investment trust to engage in joint

loan securitizations. The requested

order would supersede an existing

order.

APPLICANTS: PMC Capital, Inc. (‘‘PMC’’),

PMC Commercial Trust (the ‘‘REIT’’),

PMC Advisers, Ltd. (‘‘Advisers’’), and

PMC Asset Management, Inc.

(‘‘Managers’’).

FILING DATES: The application was filed

on September 9, 1999 and amended on

March 29, 2000 and October 24, 2000.

HEARING OR NOTIFICATION OF HEARING: An

order granting the application will be

issued unless the SEC orders a hearing.

Interested persons may request a

hearing by writing to the SEC’s

Secretary and serving applicants with a

copy of the request, personally or by

mail. Hearing requests should be

received by the SEC by 5:30 p.m. on

November 20, 2000, and should be

accompanied by proof of service on

applicants, in the form of an affidavit,

or, for lawyers, a certificate of service.

Hearing requests should state the nature

of the writer’s interest, the reason for the

request, and the issues contested.

Persons may request notification of a

hearing by writing to the SEC’s

Secretary.

ADDRESSES: Secretary, SEC, 450 Fifth

Street, N.W., Washington, D.C. 20549.

Applicants, 18111 Preston Road, Suite

600, Dallas, Texas 75252.

FOR FURTHER INFORMATION CONTACT:

Marilyn Mann, Senior Counsel, at (202)

942–0582, or Mary Kay Frech, Branch

Chief, at (202) 942–0564 (Division of

Investment Management, Office of

Investment Company Regulation).

SUPPLEMENTARY INFORMATION: The

following is a summary of the

application. The complete application

may be obtained for a fee from the SEC’s

Public Reference Branch, 450 5th Street,

N.W., Washington D.C. 20549–0102

(telephone (202) 942–8090).

Applicants’ Representations

1. PMC, a Florida corporation, is a

closed-end diversified management

investment company. On June 7, 1994,

PMC filed notification of its election to

operate as a BDC. PMC provides early

stage financing and makes available

significant managerial assistance to

small businesses and receives interest

income, loan servicing and other fees