11050 Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices

U.S.C. 552b(e) and § 9.107(a) of the

Commission’s rules that ‘‘Affirmation of

Private Fuel Storage (Independent Spent

Fuel Storage Installation) Docket No.

72–22; Certified Review of LBP–01–03’’

be held on February 14, and on less than

one week’s notice to the public.

By a vote of 5–0 on February 14, the

Commission determined pursuant to

U.S.C. 552b(e) and § 9.107(a) of the

Commission’s rules that ‘‘Affirmation of

Carolina Power & Light Company

(Shearon Harris Nuclear Power Plant);

Orange County’s Petition for Review

and Request for Immediate Suspension

and Stay of the NRC Staff’s ‘No

Significant Hazards Consideration’

Determination and Issuance of License

Amendment for Shearon Harris Spent

Fuel Pool Expansion’’ be held on

February 14, and on less than one

week’s notice to the public.

* * * * *

The NRC Commission Meeting

Schedule can be found on the Internet

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

schedule.htm

* * * * *

This notice is distributed by mail to

several hundred subscribers; if you no

longer wish to receive it, or would like

to be added to the distribution, please

contact the Office of the Secretary,

Washington, D.C. 20555 (301–415–

1969). In addition, distribution of this

meeting notice over the Intenet system

is available. if you are interested in

receiving this Commission meeting

schedule electronically, please send an

electronic message to dkw@nrc.gov.

Dated: February 15, 2001.

Sandra M. Joosten,

Executive Assistant, Office of the Secretary.

[FR Doc. 01–4369 Filed 2–16–01; 10:29 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 January 29,

2001, through February 9, 2001. The last

biweekly notice was published on

February 7, 2001.

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

Administrative Services, Office of

Administration, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001, and should cite the publication

date and page number of this Federal

Register notice. Written comments may

also be delivered to Room 6D22, Two

White Flint North, 11545 Rockville

Pike, Rockville, Maryland from 7:30

a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received

may be examined at the NRC Public

Document Room, 11555 Rockville Pike,

Room O–1F15, Rockville, Maryland.

The filing of requests for a hearing and

petitions for leave to intervene is

discussed below.

By March 23, 2001, 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, 11555

Rockville Pike, Room O–1F15,

Rockville, Maryland, 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

Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices 11051

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:

Rules and Adjudications Branch, or may

be delivered to the Commission’s Public

Document Room, 11555 Rockville Pike,

Room O–1F15, Rockville, Maryland, by

the above date. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and electronically from

the ADAMS Public Library component

on the NRC Web site, http://

www.nrc.gov (the Electronic Reading

Room).

AmerGen Energy Company, LLC, Docket

No. 50–289, Three Mile Island Nuclear

Station, Unit 1, Dauphin County,

Pennsylvania

Date of amendment request: January

15, 2001.

Description of amendment request:

The proposed amendment revises the

Technical Specification (TS) Design

Features Section 5.4.2(f), ‘‘Spent Fuel

Storage,’’ to remove the existing TS fuel

assembly U235 loading criterion for fuel

assemblies stored in the spent fuel

storage pool.

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 has no effect

on the normal operating, design basis

accident, or transient analyses applicable to

the TMI [Three Mile Island] Unit 1 fuel

storage requirements. Other existing TMI

Unit 1 Technical Specification provisions

ensure sub-criticality for normal and

postulated accident conditions.

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. Fuel assembly U235 loading is not

an initial condition of a design basis accident

or transient that either assumes the failure of

or presents a challenge to the integrity of a

fission product barrier. Discussion of fuel

assembly U235 loading in the TMI Unit 1

UFSAR [Updated Final Safety Analysis

Report] ensures that changes to fuel designs

that increase fuel reactivity relative to design

assumptions for fuel storage are evaluated in

accordance with the requirements of 10 CFR

50.59.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change does not affect existing

TMI Unit 1 Technical Specification

requirements controlling maximum fuel

enrichment, allowable enrichment vs.

burnup, soluble boron requirements, storage

rack spacing, allowable rack locations for fuel

assembly storage or sub-criticality

requirements for normal and accident

conditions. These existing Technical

Specification requirements ensure that the

current margin of safety is not reduced. The

fuel assembly U235 loading criterion does not

represent an input parameter or limiting

design condition for any supporting design

basis analyses applicable to the TMI Unit 1

spent fuel storage requirements.

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: Edward J.

Cullen, Jr., Esq., PECO Energy Company,

2301 Market Street, S23–1,

Philadelphia, PA 19103.

NRC Section Chief: Marsha

Gamberoni.

11052 Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant (BSEP),

Units 1 and 2, Brunswick County, North

Carolina

Date of amendment request: January

17, 2001.

Description of amendment request:

The proposed amendments would relax

Surveillance Requirement 3.6.1.3.7 by

allowing a ‘‘representative sample’’ of

excess flow check valves to be tested

every 24 months, such that each excess

flow check valve will be tested at least

once every 10 years.

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 license amendments do

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The current surveillance requirement

frequency requires each reactor

instrumentation line excess flow check valve

to be tested every 24 months. The excess flow

check valves at BSEP are designed to close

automatically in the event of a line break

downstream of the valve. The proposed

change allows a reduction in the number of

excess flow check valves to be tested every

24 months to approximately 20 percent of the

valves each operating cycle. Industry

operating experience demonstrates a high

level of reliability for these excess flow check

valves. A failure of an excess flow check

valve to isolate cannot initiate previously

evaluated accidents. Therefore, there is no

increase in the probability of occurrence of

an accident as a result of this proposed

change. The postulated failure of an excess

flow check valve to isolate is bounded by the

limiting analysis in the Updated Final Safety

Analysis Report (UFSAR). For a postulated

break of an instrument line upstream of an

excess flow check valve, leakage from the

line rupture would be minimized by the line

size or the flow-restricting orifice in the line.

The rate and quantity of process fluid loss

from an instrument line rupture is well

within the capability of the reactor coolant

make-up systems. The proposed change does

not alter the design of the plants’ instrument

lines in any manner, and the integrity and

functional performance of the secondary

containment and Standby Gas Treatment

system are not affected by this proposed

change. The potential offsite radiological

exposure associated with a postulated

instrument line rupture upstream of an

excess flow check valve is bounded by the

main steam line break analysis and is

substantially below the guidelines of 10 CFR

100. Therefore, the proposed license

amendments do not involve a significant

increase in the consequences of an accident

previously evaluated.

2. The proposed license amendments will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change allows a reduced

number of excess flow check valves to be

tested each operating cycle. No other change

in requirements are being proposed. Industry

operating experience demonstrates the high

reliability of the excess flow check valves.

The potential failure of an excess flow check

valve to isolate is bounded by the main steam

line break analysis. The proposed license

amendments do not physically alter the

plants and will not alter the operation of the

structures, systems, and components

described in the UFSAR. Therefore, a new or

different kind of accident will not be created.

3. The proposed license amendments do

not involve a significant reduction in a

margin of safety.

Industry experience with excess flow

check valves indicates that they have very

low failure rates. The postulated failure of an

excess flow check valve to isolate as a result

of reduced testing is bounded by the limiting

analysis in the UFSAR, which is the main

steam line break analysis. For a postulated

break of an instrument line upstream of an

excess flow check valve, leakage from the

line rupture would be minimized by the line

size or the flow-restricting orifice in the line.

The rate and quantity of process fluid loss

from an instrument line rupture is well

within the capability of the reactor coolant

make-up systems. The proposed change does

not alter the design of the plants’ instrument

line design in any manner, and the integrity

and functional performance of the secondary

containment and standby gas treatment

system are not affected by this proposed

change. Therefore, the proposed license

amendments 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: William D.

Johnson, Vice President and Corporate

Secretary, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Section Chief: Richard P.

Correia.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of amendment request: October

24, 2000.

Description of amendment request:

The proposed amendment would revise

the technical specifications to change

the Westinghouse references for Best

Estimate Large Break Loss of Coolant

Accident (LOCA) analysis methodology.

Reanalysis of large break LOCA

transients, utilizing the NRC approved

Westinghouse Best Estimate LOCA

model WCOBRA/TRAC, was performed

to demonstrate that 10 CFR 50.46

acceptance criteria are satisfied at

uprated power conditions.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

No physical plant changes are being made

as a result of using the Westinghouse Best

Estimate Large Break LOCA analysis

methodology. The proposed TS changes

simply involve updating the references in TS

5.6.5.b, ‘‘Core Operating Limits Report

(COLR),’’ to reference the Westinghouse Best

Estimate Large Break LOCA analysis

methodology (i.e., Westinghouse topical

report, WCAP–12945–P–A, Volume 1,

Revision 2, and Volumes 2 through 5,

Revision 1, ‘‘Code Qualification Document

for Best Estimate LOCA Analysis,’’ March

1998). The plant conditions assumed in the

analysis are bounded by the design

conditions for all equipment in the plant;

therefore, there will be no increase in the

probability of a LOCA. The consequences of

a LOCA are not being increased, since the

analysis has shown that the Emergency Core

Cooling System (ECCS) is designed such that

its calculated cooling performance conforms

to the criteria contained in 10 CFR 50.46,

‘‘Acceptance criteria for emergency core

cooling systems for light-water nuclear power

reactors.’’ Furthermore, the re-performance of

the Large Break LOCA analysis has no effect

on the performance of the ECCS equipment.

No other accident consequence is potentially

affected by this change.

All systems will continue to be operated in

accordance with current design requirements

under the new analysis, therefore no new

components or system interactions have been

identified that could lead to an increase in

the probability of any accident previously

evaluated in the Updated Final Safety

Analysis Report (UFSAR). No changes were

required to the Reactor Protection System

(RPS) or Engineered Safety Features (ESF)

setpoints because of the new analysis

methodology.

Based on the analysis, it is concluded that

the proposed TS changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. Does the proposed change create the

possibility of a new or different kind accident

from any accident previously evaluated?

There are no physical changes being made

to the plant as a result of using the

Westinghouse Best Estimate Large Break

LOCA analysis methodology. No new modes

of plant operation are being introduced. The

Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices 11053

configuration, operation and accident

response of the Byron Station and the

Braidwood Station systems, structures or

components are unchanged by utilization of

the new analysis methodology. Analyses of

transient events have confirmed that no

transient event results in a new sequence of

events that could lead to a new accident

scenario. The parameters assumed in the

analysis are within the design limits of

existing plant equipment.

In addition, employing the Westinghouse

Best Estimate Large Break LOCA analysis

methodology does not create any new failure

modes that could lead to a different kind of

accident. The design of all systems remains

unchanged and no new equipment or

systems have been installed which could

potentially introduce new failure modes or

accident sequences. No changes have been

made to any RPS or ESF actuation setpoints.

Based on this review, it is concluded that

no new accident scenarios, failure

mechanisms or limiting single failures are

introduced as a result of the proposed

changes. Therefore, the proposed TS changes

do not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. Does the proposed change involve a

significant reduction in a margin of Safety?

It has been shown that the analytic

technique used in the Westinghouse Best

Estimate Large Break LOCA analysis

methodology realistically describes the

expected behavior of the Byron Station and

Braidwood Station reactor system during a

postulated LOCA. Uncertainties have been

accounted for as required by 10 CFR 50.46.

A sufficient number of LOCAs with different

break sizes, different locations, and other

variations in properties have been considered

to provide assurance that the most severe

postulated LOCAs have been evaluated. The

analysis has demonstrated that there is a high

probability that all acceptance criteria

contained in 10 CFR 50.46, paragraph b,

continues to be satisfied. Based on this

review, the proposed TS changes do not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Ms. Pamela B.

Stroebel, Senior Vice President and

General Counsel, Commonwealth

Edison Company, P.O. Box 767,

Chicago, Illinois 60690–0767

NRC Section Chief: Anthony J.

Mendiola.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of amendment request:

November 7, 2000.

Description of amendment request:

The proposed amendment would revise

the technical specifications to extend

the TS Surveillance Test Interval (STI)

from a 92-day STI to an 18-month STI,

for the Solid State Protection System

(SSPS) slave relay types that meet the

acceptance criteria for the reliability

assessments performed in accordance

with the methodology described in the

NRC approved Westinghouse Electric

Corporation Topical Reports.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed changes are consistent with

the NRC approved Westinghouse Electric

Corporation Topical Reports, WCAP–13877,

‘‘Reliability Assessment of Westinghouse

Type AR Relays Used as SSPS Slave Relays,’’

WCAP–13878, ‘‘Reliability Assessment of

Potter & Brumfield MDR Series Relays,’’ and

WCAP–13900, ‘‘Extension of Slave Relay

Surveillance Testing Intervals,’’ that analyze

extending the Solid State Protection System

(SSPS) slave relay surveillance test interval

(STI) for the Westinghouse Type AR slave

relays and for the Potter & Brumfield MDR

Series slave relays. The reliability assessment

of the slave relays was comprised of a failure

modes and effects analysis (FMEA) and an

aging assessment of the slave relays. WCAP–

13877 and WCAP–13878 verified that the

Westinghouse Type AR and the Potter &

Brumfield MDR Series slave relays are highly

reliable and that degradation of the slave

relays is sufficiently slow (i.e., the time to

failure due to degradation is sufficiently

long) that an 18-month STI will adequately

identify slave relay failures. A 92-day STI is

no more likely to detect significant changes

in the SSPS slave relays than an 18-month

STI. The results demonstrate that extending

the SSPS slave relay STI from 92 days to 18

months does not adversely affect the

reliability of the SSPS slave relays utilized in

Engineered Safety Features Actuation System

(ESFAS) functions.

The high reliability of these slave relays

precludes the need for more frequent

periodic surveillance testing to verify

operability.

As stated in WCAP–13877 and WCAP–

13878, the overly conservative 92-day STI

can be extended to an 18-month STI without

impact or consequence to slave relay

reliability. In addition, the proposed changes

will not adversely affect the ability of the

SSPS to perform its safety function. The same

ESFAS instrumentation is being used and the

ESFAS reliability is being maintained with

the proposed changes. Because the reliability

of the slave relays used in the ESFAS

applications is so high, elimination of the

routine surveillance testing of the slave

relays when the reactor is at power will have

a positive impact on ESFAS availability and

plant safety. The proposed changes will not

modify any system interface and will not

increase the likelihood of any accident

initiator because such events are

independent of the proposed changes.

Therefore, the probability of an accident

previously evaluated is not increased.

The proposed changes will not modify,

degrade, or prevent actions or alter any

assumptions previously made in evaluating

the radiological consequences of any

accident described in the Updated Final

Safety Analysis Report (UFSAR). The ESFAS

instrumentation remains capable of

performing its intended safety function of

mitigation of consequences of accidents or

transients. Therefore, the consequences of an

accident previously evaluated are not

increased.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the proposed change create the

possibility of a new or different kind accident

from any accident previously evaluated?

The proposed changes do not alter the

performance of the ESFAS. The proposed

changes only extend the STI, and no changes

to the testing methodology or the way in

which the slave relays are tested are being

proposed. No new equipment is being

installed, and no installed equipment is

being operated in a new or different manner

with the proposed changes. Extending the

STI will maintain the reliability of the slave

relays as demonstrated by the NRC approved

FMEA and aging assessment, and may

improve the reliability of the system by

reducing potential test-induced degradation.

As documented in WCAP–13877 and WCAP–

13878, an STI of 92 days is no more likely

to detect significant changes in the SSPS

Type AR and MDR Series slave relays than

a STI of 18 months.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the proposed change involve a

significant reduction in a margin of safety?

The proposed changes do not affect the

total ESFAS response assumed in the safety

analysis. The periodic slave relay functional

verification is relaxed because of the

demonstrated high reliability of the slave

relays and their insensitivity to any shortterm wear or aging effects. The Westinghouse

Owners Group (WOG) program to extend the

STI for the slave relays, as documented in the

NRC approved WCAP–13877 and WCAP–

13878, has concluded that the slave relays

used in the SSPS are highly reliable and that

the surveillance testing at a frequency of 18

11054 Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices

months, instead of the 92-day STI currently

required, does not significantly decrease any

margin of safety assumed in the safety

analysis. Plant safety will be improved by

limiting the amount of on-line testing that

will be performed, because on-line testing of

the slave relays results in the removal of a

train of equipment from service or

manipulation of specific safety-related

equipment which is then no longer able to

perform its safety function if called upon

until the surveillance test is completed. The

proposed changes also act to improve plant

safety by reducing equipment degradation

and reducing unnecessary burden on the

operating personnel. There are no changes in

testing methodology or performance criteria.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.929c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Ms. Pamela B.

Stroebel, Senior Vice President and

General Counsel, Commonwealth

Edison Company, P.O. Box 767,

Chicago, Illinois 60609–0767.

NRC Section Chief: Anthony J.

Mendiola.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of amendment request:

November 13, 2000.

Description of amendment request:

The proposed amendment would revise

the technical specifications to delete the

‘‘Power Range Neutron Flux High

Negative Rate,’’ Trip Function from

Reactor Trip System Instrumentation.

The proposed change allows

elimination of this unnecessary function

and thereby reduces the potential for a

transient. The proposed changes are

consistent with the Westinghouse

Topical report previously accepted by

the NRC.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of he

issue of no significant hazards

consideration, which is presented

below:

1. Does the proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The removal of the Power Range Neutron

Flux High Negative Rate Trip (i.e., Negative

Flux Rate Trip (NFRT)) Function does not

increase the probability or consequences of

reactor core damage accidents resulting from

dropper Rod Cluster Control Assembly

(RCCA) events previously analyzed. The

safety functions of other safety related

systems and components, which are related

to mitigation of these events, have not been

altered. All other primary Reactor Trip

System (RTS) and Engineered Safety Features

Actuation Systems (ESFAS) protection

functions are not impacted by the

elimination of the NFRT Function. The NFRT

circuitry detects and responds to negative

reactivity insertion due to RCCA

misoperation events should they occur.

Therefore, the NFRT Function is not assumed

in the initiation of such events. Because the

NFRT Function is being eliminated from the

plant, it can no longer actuate and cause a

transient. The consequences of accidents

previously evaluated in the Updated Final

Safety Analysis Report (UFSAR) are

unaffected by the proposed changes because

no change to any equipment response or

accident mitigation scenario has resulted.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the proposed change create the

possibility of a new or different kind accident

from any accident previously evaluated?

The deletion of the NFRT Function does

not create the possibility of a new or different

kind of accident than any accident

previously evaluated in the UFSAR. No new

accident scenarios, failure mechanisms, or

limiting single failures are introduced as a

result of the proposed changes. The proposed

changes do not challenge the performance or

integrity of any safety related systems. It has

been demonstrated that the NFRT Function

can be eliminated by the NRC approved

methodology described in Westinghouse

Topical Report WCAP–11394–P–A,

‘‘Methodology for the Analysis of the

Dropped Rod Event,’’ dated January 1990.

The Braidwood Station and the Byron Station

cycle-specific analyses have confirmed that

for a dropped RCCA(s) event, no direct

reactor trip or automatic power reduction is

required to meet the Departure From

Nucleate Boiling (DNB) limits for this

Condition II, ‘‘Faults of Moderate

Frequency,’’ event. The NFRT Function is

not credited either as a primary or backup

mitigation feature for any other UFSAR

event. Therefore, the proposed changes do

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the proposed change involve a

significant reduction in a margin of safety?

The margin of safety associated with the

licensing basis acceptance criteria for any

postulated accident is unchanged. It has been

demonstrated that the NFRT Function can be

eliminated by the NRC approved

methodology described in WCAP 11394–P–

A. The Braidwood Station and the Byron

Station cycle-specific analyses have

confirmed that for a dropped RCCA(s) event,

DNB limits are not exceeded with the

proposed changes. Conformance to our

licensing basis acceptance criteria for Design

Basis Accidents (DBAs) and transients with

the deletion of the NFRT Function is

demonstrated, and DNB limits are not

exceeded. The proposed changes will have

no adverse effect on the availability,

operability, or performance of the safety

related systems and components assumed to

actuate in the event of a DBA or transient.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Ms. Pamela B.

Stroebel, Senior Vice President and

General Counsel, Commonwealth

Edison Company, P.O. Box 767,

Chicago, Illinois 60690–0767.

NRC Section Chief: Anthony J.

Mendiola.

Commonwealth Edison Company,

Docket Nos. 50–237 and 50–249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois

Date of amendment request: February

29, 2000.

Description of amendment request:

The proposed amendment would reduce

the number of safety valves required for

overpressure protection at Dresden, Unit

2, by removing from Technical

Specifications (TS) Section 3.6.E, the

safety valve function of the Target Rock

safety/relief valve (SRV). The proposed

amendment would move the safety

valve lift pressure setpoints from TS

Section 3.6.E to TS Section 4.6.E,

remove the Target Rock SRV setpoints

from TS, and change the number of

safety valves from nine to eight. The

proposed amendment would also

remove footnote ‘‘c’’ of Unit 3, TS

Section 4.6.E, since this footnote was

only applicable to Unit 3, Cycle 15

which has been completed.

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:

Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The probability of an evaluated accident is

derived from the probabilities of the

individual precursors to that accident. The

consequences of an evaluated accident are

determined by the operability of plant

systems designed to mitigate those

Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices 11055

consequences. Limits have been established,

consistent with Nuclear Regulatory

Commission (NRC) approved methods to

ensure that fuel performance during normal,

transient, and accident conditions is

acceptable. The proposed change to reduce

the number of required safety valves from

nine (9) to eight (8) does not affect the ability

of plant systems to adequately mitigate the

consequences of an accident previously

evaluated.

This conclusion was derived by evaluating

all applicable analyses including thermal

limit, American Society of Mechanical

Engineers (ASME) Boiler and Pressure Vessel

(B&PV) pressurization events, margin to

unpiped safety valve, anticipated transient

analysis without scram events, Loss Of

Coolant Accident (LOCA), station blackout,

and 10 CFR 50, Appendix R analyses.

Therefore, there is no increase in the

probability or consequences of an accident

previously evaluated because the analyses

supports operation without crediting the

Target Rock Safety Relief Valve safety mode

function.

Does the change create the possibility of a

new or different kind of accident from any

accident previously evaluated?

The requested change has been previously

evaluated by evaluating all applicable

analyses including thermal limit, ASME

B&PV pressurization events, margin to

unpiped safety valve, anticipated transient

analysis without scram events, station

blackout, LOCA, and 10 CFR 50, Appendix

R analyses. The proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated because the analyses

support operation without crediting the

Target Rock safety relief valve safety

function. No new failure modes will be

introduced upon implementation of the

proposed changes, therefore, the possibility

of a new and different accident has not been

created.

Does the change involve a significant

reduction in a margin of safety?

Changing the required number of safety

valves from nine (9) to eight (8) will not

involve any reduction in margin of safety.

This conclusion was derived by evaluating

all existing analyses including thermal limit,

ASME B&PV pressurization events, margin to

unpiped safety valve, anticipated transient

analysis without scram events, station

blackout, LOCA, and 10 CFR 50, Appendix

R analyses. The analyses previously

evaluated remain valid, therefore, a

significant reduction in the margin of safety

does not exist.

Therefore, based upon the above

evaluation, ComEd has concluded that these

changes do not constitute a significant

hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Ms. Pamela B.

Stroebel, Senior Vice President and

General Counsel, Commonwealth

Edison Company, P.O. Box 767,

Chicago, Illinois 60690–0767.

NRC Section Chief: Anthony J.

Mendiola.

Consolidated Edison Company of New

York, Docket No. 50–247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of amendment request: February

14, 2000.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TSs) to

correct various editorial errors and make

other administrative changes.

Specifically, the proposed amendment

would make administrative changes that

revise: (a) Tables 3.6–1 and 4.4–1 to

correct listing and editorial errors, (b)

TS 3.8.B.10 to reflect the wording in 10

CFR 50.54(m)(2)(iv), (c) Figures 3.10–2

through 3.10–6 to remove these figures,

(d) Table 4.1–1 to reflect change in level

indication components, (e) TS 4.19.B

and 6.14.1.1 to correct editorial errors,

(f) TS 6.12.1 to reference the current

sections of 10 CFR Part 20, (g) TS 6.12.1

to reflect an organizational title change,

and (h) TS 6.13.2 to correct a

typographical error.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(a) Changes To Tables 3.6–1 And 4.4–1 To

Correct Listing And Editorial Errors

(1) Does the proposed license amendment

involve a significant increase in the

probability or in the consequences of an

accident previously evaluated?

No. The proposed changes are

administrative in nature. The changes

involve correcting errors in Table 3.6–1 and

additions to Tables 3.6–1 and 4.4–1 to reflect

UFSAR [Updated Final Safety Analysis

Report] Table 5.2–1 and the IST [inservice

testing] Program. These changes do not affect

possible initiating events for accidents

previously evaluated or alter the

configuration or operation of the facility. The

Limiting Safety System Settings and Safety

Limits specified in the current Technical

Specifications remain unchanged. Therefore,

the proposed changes would not involve a

significant increase in the probability or in

the consequences of an accident previously

evaluated.

(2) Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

No. The proposed changes are

administrative in nature. The safety analysis

of the facility remains complete and accurate.

There are no physical changes to the facility

and the plant conditions for which the design

basis accidents have been evaluated are still

valid. The operating procedures and

emergency procedures are unaffected.

Consequently no new failure modes are

introduced as a result of the proposed

changes. Therefore, the proposed changes

would not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

No. The proposed changes are

administrative in nature. Since there are no

changes to the operation of the facility or the

physical design, the Updated Final Safety

Analysis Report (UFSAR) design basis,

accident assumptions, or Technical

Specification Bases are not affected.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

(b) Change To Section 3.8.B.10 To Reflect

The Wording In 10 CFR 50.54(m)(2)(iv)

(1) Does the proposed license amendment

involve a significant increase in the

probability or in the consequences of an

accident previously evaluated?

No. The proposed change is administrative

in nature. The change involves updating

Section 3.8.B.10 to reflect 10 CFR

50.54(m)(2)(iv). This change does not affect

possible initiating events for accidents

previously evaluated or alter the

configuration or operation of the facility. The

Limiting Safety System Settings and Safety

Limits specified in the current Technical

Specifications remain unchanged. Therefore,

the proposed change would not involve a

significant increase in the probability or in

the consequences of an accident previously

evaluated.

(2) Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

No. The proposed change is administrative

in nature. The safety analysis of the facility

remains complete and accurate. There are no

physical changes to the facility and the plant

conditions for which the design basis

accidents have been evaluated are still valid.

The operating procedures and emergency

procedures are unaffected. Consequently no

new failure modes are introduced as a result

of the proposed change. Therefore, the

proposed change would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

No. The proposed change is administrative

in nature. Since there are no changes to the

operation of the facility or the physical

design, the Updated Final Safety Analysis

Report (UFSAR) design basis, accident

assumptions, or Technical Specification

Bases are not affected. Therefore, the

proposed change does not involve a

significant reduction in a margin of safety.

(c) Deletion Of Figures 3.10–2 Through

3.10–6

(1) Does the proposed license amendment

involve a significant increase in the

11056 Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices

probability or in the consequences of an

accident previously evaluated?

No. The proposed change is administrative

in nature. The change involves the deletion

of Figures 3.10–2, 3.10–3, 3.10–4, 3.10–5 and

3.10–6. This change does not affect possible

initiating events for accidents previously

evaluated or alter the configuration or

operation of the facility. The Limiting Safety

System Settings and Safety Limits specified

in the current Technical Specifications

remain unchanged. Therefore, the proposed

change would not involve a significant

increase in the probability or in the

consequences of an accident previously

evaluated.

(2) Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

No. The proposed change is administrative

in nature. The safety analysis of the facility

remains complete and accurate. There are no

physical changes to the facility and the plant

conditions for which the design basis

accidents have been evaluated are still valid.

The operating procedures and emergency

procedures are unaffected. Consequently no

new failure modes are introduced as a result

of the proposed change. Therefore, the

proposed change would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

No. The proposed change is administrative

in nature. Since there are no changes to the

operation of the facility or the physical

design, the Updated Final Safety Analysis

Report (UFSAR) design basis, accident

assumptions, or Technical Specification

Bases are not affected. Therefore, the

proposed change does not involve a

significant reduction in a margin of safety.

(d) Change To Table 4.1–1 To Reflect

Change In Level Indication Components

(1) Does the proposed license amendment

involve a significant increase in the

probability or in the consequences of an

accident previously evaluated?

No. This change does not affect possible

initiating events for accidents previously

evaluated or operation of the facility. While

the configuration of the facility has changed

with installation of the new level sensors, the

safety-related function of theses sensors

remains unchanged (i.e., at a predetermined

level of approximately 35% of instrument

span, a low level alarm will annunciate in

the CCR [control room]). The Limiting Safety

System Settings and Safety Limits specified

in the current Technical Specifications

remain unchanged. Therefore, the proposed

change would not involve a significant

increase in the probability or in the

consequences of an accident previously

evaluated.

(2) Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

No. The safety analysis of the facility

remains complete and accurate. The plant

conditions for which the design basis

accidents have been evaluated are still valid.

While the configuration of the facility has

changed with installation of the new level

sensors, the safety-related function of theses

[sic] sensors remains unchanged (i.e., at a

predetermined level of approximately 35% of

instrument span, a low level alarm will

annunciate in the CCR). Consequently no

new failure modes are introduced as a result

of the proposed change. Therefore, the

proposed change would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

No. While the configuration of the facility

has changed with installation of the new

level sensors, the safety-related function of

theses sensors remains unchanged (i.e., at a

predetermined level of approximately 35% of

instrument span, a low level alarm will

annunciate in the CCR). Also, there are no

changes to the operation of the facility. Thus

the Updated Final Safety Analysis Report

(UFSAR) design basis, accident assumptions,

or Technical Specification Bases are not

affected. Therefore, the proposed change

does not involve a significant reduction in a

margin of safety.

(e) Change To Sections 4.19.B And 6.14.1.1

To Correct Editorial Errors

(1) Does the proposed license amendment

involve a significant increase in the

probability or in the consequences of an

accident previously evaluated?

No. The proposed changes are

administrative in nature. The change in

Sections 4.19.B and 6.14.1.1 involve

amending ‘‘the Semiannual Radioactive

Effluent Release Report’’ to ‘‘the Annual

Radioactive Effluent Release Report.’’ These

changes do not affect possible initiating

events for accidents previously evaluated or

alter the configuration or operation of the

facility. The Limiting Safety System Settings

and Safety Limits specified in the current

Technical Specifications remain unchanged.

Therefore, the proposed changes would not

involve a significant increase in the

probability or in the consequences of an

accident previously evaluated.

(2) Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

No. The proposed changes are

administrative in nature. The safety analysis

of the facility remains complete and accurate.

There are no physical changes to the facility

and the plant conditions for which the design

basis accidents have been evaluated are still

valid. The operating procedures and

emergency procedures are unaffected.

Consequently no new failure modes are

introduced as a result of the proposed

change. Therefore, the proposed changes

would not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

No. The proposed changes are

administrative in nature. Since there are no

changes to the operation of the facility or the

physical design, the Updated Final Safety

Analysis Report (UFSAR) design basis,

accident assumptions, or Technical

Specification Bases are not affected.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

(f) Change To Section 6.12.1 To Reference

The Current Sections Of 10 CFR [Part] 20

(1) Does the proposed license amendment

involve a significant increase in the

probability or in the consequences of an

accident previously evaluated?

No. The proposed change is administrative

in nature. The change involves updating

Section 6.12.1 to reference 10 CFR 20.1601(a)

and 10 CFR 20.1601(b). This change does not

affect possible initiating events for accidents

previously evaluated or alter the

configuration or operation of the facility. The

Limiting Safety System Settings and Safety

Limits specified in the current Technical

Specifications remain unchanged. Therefore,

the proposed change would not involve a

significant increase in the probability or in

the consequences of an accident previously

evaluated.

(2) Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

No. The proposed change is administrative

in nature. The safety analysis of the facility

remains complete and accurate. There are no

physical changes to the facility and the plant

conditions for which the design basis

accidents have been evaluated are still valid.

The operating procedures and emergency

procedures are unaffected. Consequently no

new failure modes are introduced as a result

of the proposed change. Therefore, the

proposed change would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

No. The proposed change is administrative

in nature. Since there are no changes to the

operation of the facility or the physical

design, the Updated Final Safety Analysis

Report (UFSAR) design basis, accident

assumptions, or Technical Specification

Bases are not affected. Therefore, the

proposed change does not involve a

significant reduction in a margin of safety.

(g) Change To Section 6.12.1 To Reflect An

Organizational Title Change

(1) Does the proposed license amendment

involve a significant increase in the

probability or in the consequences of an

accident previously evaluated?

No. The proposed change is administrative

in nature. The change involves updating

Section 6.12.1 to use the title ‘‘Shift

Manager’’ instead of ‘‘Senior Watch

Supervisor.’’ This change does not affect

possible initiating events for accidents

previously evaluated or alter the

configuration or operation of the facility. The

Limiting Safety System Settings and Safety

Limits specified in the current Technical

Specifications remain unchanged. Therefore,

the proposed change would not involve a

significant increase in the probability or in

the consequences of an accident previously

evaluated.

(2) Does the proposed amendment create

the possibility of a new or different kind of

Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices 11057

accident from any accident previously

evaluated?

No. The proposed change is administrative

in nature. The safety analysis of the facility

remains complete and accurate. There are no

physical changes to the facility and the plant

conditions for which the design basis

accidents have been evaluated are still valid.

The operating procedures and emergency

procedures are unaffected. Consequently no

new failure modes are introduced as a result

of the proposed change. Therefore, the

proposed change would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

No. The proposed change is administrative

in nature. Since there are no changes to the

operation of the facility or the physical

design, the Updated Final Safety Analysis

Report (UFSAR) design basis, accident

assumptions, or Technical Specification

Bases are not affected. Therefore, the

proposed change does not involve a

significant reduction in a margin of safety.

(h) Change To Section 6.13.2 To Correct A

Typographical Error

(1) Does the proposed license amendment

involve a significant increase in the

probability or in the consequences of an

accident previously evaluated?

No. The proposed change is administrative

in nature. The change involves updating

Section 6.13.2 from ‘‘DOR [Division of

Operating Reactors] Guidelines of NUREG–

0588’’ to ‘‘DOR Guidelines or NUREG–0588.’’

This change does not affect possible

initiating events for accidents previously

evaluated or alter the configuration or

operation of the facility. The Limiting Safety

System Settings and Safety Limits specified

in the current Technical Specifications

remain unchanged. Therefore, the proposed

change would not involve a significant

increase in the probability or in the

consequences of an accident previously

evaluated.

(2) Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

No. The proposed change is administrative

in nature. The safety analysis of the facility

remains complete and accurate. There are no

physical changes to the facility and the plant

conditions for which the design basis

accidents have been evaluated are still valid.

The operating procedures and emergency

procedures are unaffected. Consequently no

new failure modes are introduced as a result

of the proposed change. Therefore, the

proposed change would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Does the proposed amendment involve

a significant reduction in a margin of safety?

No. The proposed change is administrative

in nature. Since there are no changes to the

operation of the facility or the physical

design, the Updated Final Safety Analysis

Report (UFSAR) design basis, accident

assumptions, or Technical Specification

Bases are not affected. 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: Brent L.

Brandenburg, Esq., 4 Irving Place, New

York, New York 10003.

NRC Section Chief: Marsha

Gamberoni.

Consolidated Edison Company of New

York, Docket No. 50–247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of amendment request:

December 11, 2000.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TSs) to

provide editorial revisions,

clarifications, and corrections.

Specifically, the proposed amendment

would: (1) Provide updated information

and corrections to the TS cover page,

table of contents, and list of figures, (2)

revise TS 4.5.E, ‘‘Control Room Air

Filtration System,’’ to remove an

incorrect system test description and

provide consistent test values for system

flow rate and filter efficiency, (3) revise

TS 6.2.1.a, ‘‘Facility Management and

Technical Support,’’ to reference the

Quality Assurance Program Description

as the location of the documentation

rather than the Updated Final Safety

Analysis Report, (4) revise TS 6.9.1.7,

‘‘Monthly Operating Report,’’ to change

the recipient of the Monthly Operating

Report, and (5) correct the periodicity of

the Radioactive Effluent Release Report

from annual to semiannual in TS 6.15,

‘‘Offsite Dose Calculation Manual’’ and

TS 6.16, ‘‘Major Changes to Radioactive

Liquid, Gaseous and Solid Waste

Systems.’’ In addition, the proposed

change revises TS Figure 5.1–1B

concerning the indicated vent location

associated with Indian Point Unit 3

(IP3). The labels for the plant vent and

the machine shop are reversed.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the proposed license amendment

involve a significant increase in the

probability or [ * * * ] consequences of an

accident previously evaluated?

The proposed changes consist of editorial

changes, administrative changes,

clarifications, and corrections to existing TSs.

These changes do not involve a change to the

design or operation of any plant system nor

are any of the safety analyses affected as a

result of these changes. Accordingly, the

initiators of any accident as well as any

system relied upon for the mitigation of an

accident are not affected by the proposed

changes. Therefore, there is no increase in

the probability or in the consequences of an

accident previously evaluated.

2. Does the proposed amendment create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed changes do not involve a

change to the design or operation of any

plant system. These changes include editorial

changes, administrative changes,

clarifications, and corrections of existing TSs

and, therefore, do not create the possibility

of a new or different kind of accident from

any accident previously evaluated.

3. Does the proposed amendment involve

a significant reduction in a margin of safety?

The proposed changes consist of editorial

changes, administrative changes, and

clarifications to existing TSs and do not

involve changes to any margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Brent L.

Brandenburg, Esq., 4 Irving Place, New

York, New York 10003.

NRC Section Chief: Marsha

Gamberoni.

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: January

8, 2001.

Description of amendment request:

The proposed change revises the lower

limit of the allowable containment

internal pressure in Technical

Specification (TS) 3.6.1.4, ‘‘Containment

Systems—Internal Pressure,’’ from

14.375 pounds per square inch, absolute

(psia) to 14.275 psia.

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 operation of the facility in

accordance with this proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response: The proposed change revises the

lower limit of the allowable containment

internal pressure in TS 3.6.1.4 from 14.375 to

14.275 psia. This change will allow

11058 Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices

additional operating margin for the

containment atmosphere purge (CAP) system

during conditions of low atmospheric

pressure. The containment minimum

pressure parameter is not an accident

initiator and does not affect the probability

of any initiating event scenario. Although the

TSs will allow a lower initial containment

internal pressure, the current analyses for the

associated design events are not affected

since the lower pressure has already been

conservatively included. The proposed

change in initial containment internal

pressure is bounded in the current design.

Therefore, this proposed change does not

involve an increase in the probability or

consequences of an accident previously

evaluated.

2. Will the operation of the facility in

accordance with this proposed change create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: The proposed change affects the

TS allowed lower limit on containment

internal pressure and consequently the

atmospheric range in which the CAP system

can be operated. The change in the lower

limit on containment internal pressure is

encompassed by current design analyses and

does not result in a change of analyzed

conditions or analyzed operating ranges.

Based on the proposed TS change, CAP

system operation will be allowed at a lower

atmospheric pressure. This change does not

change the function of the system or its

method of operation. Although the initial

atmospheric pressure at which the CAP

system can be initiated is being lowered, this

is within the current design of the CAP

system and does not change the differential

pressures at which it will be operated.

Therefore, the proposed change[d] does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Will the operation of the facility in

accordance with this proposed change

involve a significant reduction in a margin of

safety?

Response: The proposed change makes use

of the initial containment pressure

assumption values used in the current

analyses to provide additional operating

margin for the CAP system. The margin of

safety that was inherent in the results of

these safety analyses has been preserved. The

associated analyses ensure the negative

pressure differential associated with an

inadvertent actuation of the containment

spray system is acceptable, and ensure that

the emergency core cooling system can

satisfy its design safety function under worst

case conditions. The calculated maximum

differential pressure is 0.49 psid [pounds per

square inch differential] which is within the

design limit of 0.65 psid. The peak clad

temperature for the worst case large break

loss of coolant accident is 2177°F which is

within the acceptance criteria given in

10CFR50.46. Since the proposed change does

not affect the initial containment pressure

utilized in these analyses, the results of the

analyses are unchanged. Therefore, there is

no affect on any margin of safety associated

with this parameter.

Based on the above No Significant Hazards

Consideration Determination, it is concluded

that: (1) The proposed change does not

constitute a significant hazards consideration

as defined by 10CFR50.92; (2) there is a

reasonable assurance that the health and

safety of the public will not be endangered

by the proposed change; and (3) this action

will not result in a condition which

significantly alters the impact of the station

on the environment as described in the NRC

final environmental statement.

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: N. S. Reynolds,

Esquire, Winston & Strawn 1400 L

Street NW., Washington, DC 20005–

3502.

NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, Docket

Nos. 50–237 and 50–249, Dresden

Nuclear Power Station, Units 2 and 3,

Grundy County, Illinois

Date of amendment request:

September 1, 2000.

Description of amendment request:

The proposed amendments would

revise the technical specifications to

add a new requirement for the Main

Steam Line Radiation Monitor

mechanical vacuum pump trip function.

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:

Does the change involve a significant

increase in the probability of occurrence or

consequences of an accident previously

evaluated?

The addition of the MSLRM [Main Steam

Line Radiation Monitor] automatic trip signal

to the MVP [mechanical vacuum pump] has

no adverse effect on safety. The addition of

Surveillance Requirements (SRs) and the

Limiting Condition for Operation (LCO) to

our TS enhances current safety features of the

plant by establishing controls for a required,

and currently functional, safety feature. The

automatic trip function of the MVP does not

serve as an initiator for any accidents

evaluated in Chapter 15, ‘‘Accident and

Transient Analysis,’’ of the Updated Final

Safety Analysis Report. Therefore, this

change will not result in an increase of either

the probability or consequences of an

accident.

Does the change create the possibility of a

new or different kind of accident from any

accident previously evaluated?

These proposed changes involve the

addition of the MVP trip input from the Main

Steam Line Tunnel High Radiation signal.

The addition of this function does not

represent a change in operating parameters or

equipment configuration for DNPS [Dresden

Nuclear Power Station], Units 2 and 3.

Operation of DNPS, Units 2 and 3, under the

proposed changes does not create the

possibility of a new or different type of

accident previously evaluated.

Does the change involve a significant

reduction in a margin of safety?

These proposed changes create a TS LCO

and identify SRs for the MVP trip input from

the MSLRM signal. Operation under the

proposed change will not change any plant

operation parameters, nor any protective

system setpoints. The calculations of off site

dose demonstrate that with the MVP trip

instrumentation operating properly, the

doses that result from a CRDA [control rod

drop accident] with the MVP operating are

well within 10 CFR Part 100, ‘‘Reactor Site

Criteria,’’ limits. [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

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Ms. Pamela B.

Stroebel, Senior Vice President and

General Counsel, Commonwealth

Edison Company, P.O. Box 767,

Chicago, Illinois 60690–0767.

NRC Section Chief: Anthony J.

Mendiola.

Exelon Generating Company, LLC

(Exelon), Docket No. 50–353, Limerick

Generating Station, Unit 2, Montgomery

County, Pennsylvania

Date of amendment request:

November 20, 2000.

Description of amendment request:

PECO Energy Company (PECO)

proposed changes to the Technical

Specifications (TSs) that would revise

the heatup, cooldown, and inservice test

Pressure-Temperature (P–T) limitations

(TS Figure 3.4.6.1–1) of the Limerick

Generating Station (LGS), Unit 2,

Reactor Pressure Vessel (RPV) to a

maximum of 32 Effective Full Power

Years (EFPY). In addition, the licensee

proposed text changes to both Limiting

Condition for Operation 3.4.6.1 and

Surveillance Requirement 4.4.6.1.1 to

delete the reference to the A’ curve on

TS Figure 3.4.6.1–1 since this curve will

not be included in the proposed Figure

3.4.6.1–1. The licensee also proposed

adding an intermediate hydrotest curve

(Curve A22) to TS Figure 3.4.6.1–1,

which is valid to 22 EFPY. By letter

dated January 30, 2001, Exelon stated

that it has assumed responsibility, as of

the date of the transfer, for the active

items on the Limerick Units 1 and 2

Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices 11059

dockets previously submitted by PECO,

including the subject amendment

request.

Moreover, Exelon is revising its TS

Bases Section B 3/4.4.6 to update

several RPV material chemistry

parameters. The licensee identified the

need for these revisions during a

Certified Material Test Report data

search performed by General Electric

Company during development of the

proposed P–T curves.

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:

1. The proposed TS changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

There are no physical changes to the plant

being introduced by the proposed changes to

the P–T curves. The proposed changes do not

modify the reactor coolant pressure

boundary, i.e., there are no changes in

operating pressure, materials or seismic

loading. The proposed changes do not

adversely affect the integrity of the reactor

coolant pressure boundary such that its

function in the control of radiological

consequences is affected. The proposed P–T

curves were generated in accordance with the

fracture toughness requirements of 10 CFR

50, Appendix G, and American Society of

Mechanical Engineers (ASME) Boiler and

Pressure Vessel (B&PV) Code, Section XI,

Appendix G, in conjunction with ASME

Code Case N–640. The proposed P–T curves

were established in compliance with the

methodology used to calculate the predicted

irradiation effects on vessel beltline

materials. Usage of these procedures provides

compliance with the intent of 10 CFR 50,

Appendix G, and provides margins of safety

that ensure that failure of the reactor vessel

will not occur. The proposed P–T curves

prohibit operational conditions in which

brittle fracture of reactor vessel materials is

possible. Consequently, the primary coolant

pressure boundary integrity will be

maintained. Therefore, the proposed changes

do not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes to the P–T curves

were generated in accordance with the

fracture toughness requirements of 10 CFR

50, Appendix G, and ASME B&PV Code,

Section XI, Appendix G, in conjunction with

ASME Code Case N–640. Compliance with

the proposed P–T curves will ensure that

conditions in which brittle fracture of

primary coolant pressure boundary materials

are possible will be avoided. No new modes

of operation are introduced by the proposed

changes. The proposed changes will not

create any new failure mode from previously

evaluated accidents. Further, the proposed

changes to the P–T curves do not affect any

activities or equipment, and are not assumed

in any safety analysis to initiate nor mitigate

any accident sequence. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The proposed changes reflect an update of

the P–T curves to extend the reactor pressure

vessel operating limit to 32 Effective Full

Power Years (EFPY). The revised curves are

based on the latest ASME guidance. The

revised P–T limits, which provide more

operational flexibility than the current limits,

were established in accordance with current

regulations and the latest ASME Code

information. No plant safety limits, set

points, or design parameters are adversely

affected by the proposed TS changes. These

proposed changes maintain the relative

margin of safety commensurate with that

which existed at the time that the ASME

B&PV Code, Section XI, Appendix G, was

approved in 1974.

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: J.W. Durham,

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

Counsel, Exelon Generating Company,

2301 Market Street, Philadelphia, PA

19101

NRC Section Chief: James W. Clifford.

Exelon Generation Company, LLC

(Exelon), Docket Nos. 50–352 and 50–

353, Limerick Generating Station, Units

1 and 2, Montgomery County,

Pennsylvania

Date of amendment request: January

18, 2001.

Description of amendment request:

Exelon requested a Technical

Specification (TS) change which will

revise Surveillance Requirement (SR)

4.9.2.d.1 to clarify that ‘‘shorting links’’

do not need to be removed, if adequate

shutdown margin has been

demonstrated, when moving a control

rod during Operational Condition 5.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below:

1. The proposed TS change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

This TS Change Request revises SR

4.9.2.d.1 to clarify that ‘‘shorting links’’ do

not need to be removed if adequate shutdown

margin has been demonstrated when a

control rod is withdrawn during Operational

Condition 5. This revision ensures that the

words and intent of the SR 4.9.2.d.1 match

the words and intent of Limiting Condition

for Operation (LCO) 3.9.2.d, and will

improve the readability of the SR for plant

operators. This change to SR 4.9.2.d.1 will

clarify that ‘‘shorting links’’ can remain

installed if adequate shutdown margin has

been demonstrated any time a control rod is

withdrawn in Operational Condition 5. This

revision does not impact any accident or

transient events. There are no new initiators

created by this revision. Additionally, this

revision will not impact any existing

analyses or requirements contained in the

Updated Final Safety Analysis Report. No

changes in the operation of the plant during

power operation or refueling will occur as a

result of this revision. Therefore, the

proposed TS revision does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS revision does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed TS revision will not impact

any physical changes to plant structures,

systems, or components. The design,

function, and reliability of the Reactor

Protection System will not be impacted by

this revision. This revision does not

adversely impact any equipment which is

required for the prevention or mitigation of

accidents or transients. This revision ensures

that the words and intent of the SR 4.9.2.d.1

match the words and intent of LCO 3.9.2.d,

and will improve the readability of the SR for

plant operators. Therefore, the proposed

change does not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety.

This proposed revision to SR 4.9.2.d.1 does

not affect any safety limits or analytical

limits. There are also no changes to accident

or transient core thermal hydraulic

conditions, minimum combustible

concentration limits, or fuel or reactor

coolant boundary design limits, as a result of

this proposed change. This revision ensures

that the words and intent of the SR 4.9.2.d.1

match the words and intent of LCO 3.9.2.d.

Therefore, the proposed change does 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: J. W. Durham,

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

11060 Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices

Counsel, Exelon Generating Company,

2301 Market Street, Philadelphia, PA

19101.

NRC Section Chief: James W. Clifford.

Exelon Generation Company, LLC

(Exelon), Docket Nos. 50–352 and

50–353, Limerick Generating

Station, Units 1 and 2, Montgomery

County, Pennsylvania

Date of amendment request: January

18, 2001.

Description of amendment request:

Exelon requested a Technical

Specification (TS) change which will

revise the Units 1 and 2 TS Table 1.2,

‘‘Operational Conditions,’’ to allow

placing the reactor mode switch to the

REFUEL position during Operational

Conditions 3 and 4 to accommodate

post maintenance and surveillance

testing on control rod drives.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below:

1. The proposed TS change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed revision allows a single

control rod to be withdrawn under control of

the reactor mode switch REFUEL position

and the one-rod-out interlock in Operational

Conditions 3 and 4. This change does not

affect any existing accident initiators. There

is no change to the coupling integrity of the

control rod during this accident. Although

this change would allow an increase in the

frequency of single control rod withdrawals

in Operational Conditions 3 and 4, the

probability of the previously analyzed

accidents is not affected.

The onsite and offsite radiological

consequences of previously analyzed

accidents in Operational Conditions 3 and 4

are not affected by this proposed change.

This change does not affect any existing

accident mitigators. The shutdown margin

combined with the refueling interlocks

prevent a rod withdrawal error while in

refueling thereby preventing inadvertent

criticality. There is no impact on the ability

of the Reactor Protection System (RPS)

circuitry to mitigate a Control Rod Drop

Accident as described in the Safety Analysis

Report, nor is there an increase in the

number of fuel failures from this accident. As

a result, the probability and consequences of

previously analyzed accidents are not

significantly increased.

2. The proposed TS change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

There are no new accident initiators

created by the proposed revision to Table 1.2.

Single control rods can be withdrawn in

Operational Conditions 3 and 4 under the

existing Technical Specifications to permit

control rod recoupling. The proposed

revision would expand this provision to

other control rod maintenance and testing

activities performed in Operational

Conditions 3 and 4. The withdrawal of

individual control rods in Operational

Conditions 3 and 4 is a mode of operation

permitted under limited circumstances by

the existing TSs. The additional control rod

maintenance and testing activities which

could be performed in Operational

Conditions 3 and 4, are already permitted by

the existing TSs in Operational Conditions 1,

2, 4, and 5.

Based on the above, this change does not

create the possibility of a new or different

kind of accident from any previously

evaluated.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety.

The TSs currently permit single control rod

withdrawal for the purpose of control rod

recoupling when in Operational Conditions 3

or 4 if the one-rod-out interlock is Operable.

This change allows additional activities for

which a single control rod may be withdrawn

when in Operational Conditions 3 or 4, with

the same restriction that the one-rod-out

interlock be Operable.

The operability requirements for the onerod-out interlock and the shutdown margin

requirements of TS 3.1.1 ensure the reactor

will be maintained subcritical during single

control rod withdrawals. Therefore, this

change will 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: J. W. Durham,

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

Counsel, Exelon Generation Company,

2301 Market Street, Philadelphia, PA

19101.

NRC Section Chief: James W. Clifford.

Exelon Generating Company, LLC

(Exelon), Docket No. 50–353, Limerick

Generating Station, Unit 2, Montgomery

County, Pennsylvania.

Date of amendment request: February

1, 2001.

Description of amendment request:

Exelon proposed changes that would

revise Technical Specification (TS) 2.1

to incorporate revised Safety Limit

Minimum Critical Power Ratios due to

the cycle-specific analysis performed by

Global Nuclear Fuel for Limerick

Generating Station, Unit 2, Cycle 7,

which will include the use of the GE–

14 fuel product line.

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 TS change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The derivation of the cycle specific Safety

Limit Minimum Critical Power Ratios

(SLMCPRs) for incorporation into the

Technical Specifications (TS), and its use to

determine cycle specific thermal limits, has

been performed using the methodology

discussed in ‘‘General Electric Standard

Application for Reactor Fuel,’’ NEDE–24011–

P–A–14 (GESTAR–II), and U.S. Supplement,

NEDE–24011–P–A–14–US, June, 2000, which

incorporates Amendment 25. Amendment 25

was approved by the NRC [Nuclear

Regulatory Commission] in a March 11, 1999

safety evaluation report.

The basis of the SLMCPR calculation is to

ensure that greater than 99.9% of all fuel rods

in the core avoid transition boiling if the

limit is not violated. The new SLMCPRs

preserve the existing margin to transition

boiling. The GE–14 fuel is in compliance

with Amendment 22 to ‘‘General Electric

Standard Application for Reactor Fuel,’’

NEDE–24011–P–A–14 (GESTAR–II), and U.S.

Supplement, NEDE–24011–P–A–14–US,

June, 2000, which provides the fuel licensing

acceptance criteria. The probability of fuel

damage will not be increased as a result of

this change. Therefore, the proposed TS

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed TS change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The SLMCPR is a TS numerical value,

calculated to ensure that transition boiling

does not occur in 99.9% of all fuel rods in

the core if the limit is not violated. The new

SLMCPRs are calculated using NRC approved

methodology discussed in ‘‘General Electric

Standard Application for Reactor Fuel,’’

NEDE–24011–P–A–14 (GESTAR–II), and U.S.

Supplement, NEDE–24011–P–A–14–US,

June, 2000, which incorporates Amendment

25. Additionally, the GE–14 fuel is in

compliance with Amendment 22 to ‘‘General

Electric Standard Application for Reactor

Fuel,’’ NEDE–24011–P–A–14 (GESTAR–II),

and U.S. Supplement, NEDE–24011–P–A–

14–US, June 2000, which provides the fuel

licensing acceptance criteria. The SLMCPR is

not an accident initiator, and its revision will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety.

There is no significant reduction in the

margin of safety previously approved by the

NRC as a result of the proposed change to the

SLMCPRs, which includes the use of GE–14

fuel. The new SLMCPRs are calculated using

Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices 11061

methodology discussed in ‘‘General Electric

Standard Application for Reactor Fuel,’’

NEDE–24011–P–A–14 (GESTAR–II), and U.S.

Supplement, NEDE–24011–P–A–14–US,

June, 2000, which incorporates Amendment

25. The SLMCPRs ensure that greater than

99.9% of all fuel rods in the core will avoid

transition boiling if the limit is not violated

when all uncertainties are considered,

thereby preserving the fuel cladding

integrity. Therefore, the proposed TS change

will not involve a significant reduction in the

margin of safety previously approved by the

NRC.

Based on the staff’s review of the

licensee’s evaluation, 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: J. W. Durham,

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

Counsel, Exelon Generating Company,

2301 Market Street, Philadelphia, PA

19101.

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: January

13, 2000.

Description of amendment request:

The proposed amendment would

change the Kewaunee Nuclear Power

Plant Technical Specification 3.6,

‘‘Containment.’’ The proposed

amendment would add limiting

condition for operation and allowed

outage times for containment

penetrations and associated isolation

devices to provide clear guidance. Also,

the proposed amendment would

provide additional information,

clarification, and uniformity to the basis

of the associated technical specification.

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.

This Technical Specification [TS] change

provides definition for the [Allowable Outage

Time] AOT for a containment isolation valve

and containment air lock. The original design

and design basis of the plant is still

maintained and the probability and

consequences of previously evaluated

accidents is unchanged. In our current

Technical Specifications the allowed outage

time for a safeguards 480-volt bus is 24

hours. The basis for this outage time states:

‘‘The intent of this TS is to provide

assurance that at least one external source

and one standby source of electrical power is

always available to accomplish safe

shutdown and containment isolation and to

operate required engineered safety features

equipment following an accident.’’

With one 480-volt safeguards bus out of

service an associated motor operated

containment isolation valve is also out of

service. Since the 24-hour AOT is part of

Kewaunee’s original design basis, allowing

the containment isolation valves to be out of

service for 24 hours does not increase the

probability or consequences of an accident

previously evaluated.

A risk assessment of the probability of a

-loss-of-coolant-accident with a train of

containment isolation failing during a 4-hour

verse a 24-hour time span was conducted.

The probability of [loss-of-coolant accident]

LOCA coincident with the failure of

containment isolation occurring during a 4-

hour period versus a 24-hour period is shown

on Figure 1[ in licensee’s submittal]. This

change in probability is considered

insignificant.

The proposed TS changes do not involve

any physical or operational changes to

structures, systems or components. The

current safety analysis and design basis for

the accident mitigation functions of the

containment, the airlocks, and the

containment isolation valves are maintained.

On-site and off-site dose consequences

remain unaffected.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The function of the containment vessel is

to contain the radiologically hazardous

material following a LOCA. By maintaining

at least one containment isolation barrier

intact the vessel can perform its function.

This amendment still ensures that at least

one barrier is intact or the leakage is

evaluated not to exceed that which is already

evaluated and allowed by current technical

specification.

The accidents considered are found in the

Safety Analysis, Section 14 of the [Updated

Safety Analysis Report] USAR. The proposed

change does not involve a change to the plant

design (structures, systems or components) or

operation. No new failure mechanisms

beyond those already considered in the

current plant Safety Analysis are introduced.

No new accident is introduced and no safetyrelated equipment or safety functions are

altered. The proposed change does not affect

any of the parameters or conditions that

contribute to initiation of any accidents.

3. Involve a significant reduction in a

margin of safety.

With one containment barrier intact during

plant operation the isolation of containment

is still ensured. The plant’s original design

basis addressed the inability of one of the

two containment isolation valves to operate

for a 24-hour period. As this AOT has been

previously considered, there therefore is no

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: Bradley D.

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

Box 1497, Madison, WI 53701–1497.

NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: January

18, 2001.

Description of amendment request:

The proposed amendment would

change the Kewaunee Nuclear Power

Plant Technical Specification 3.10.m for

reactor coolant minimum flow from the

current value of 85,500 gallons per

minute (gpm) to 93,000 gpm due to the

replacement of steam generators

scheduled for the fall 2001.

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 change in Reactor Coolant Minimum

Flow value for TS 3.10.m proposed in this

amendment request is needed to reflect

operating characteristics of the new

[Replacement Steam Generators] RSGs.

Accident analyses affected by the RSGs have

each been evaluated to establish that there is

no significant change in the documented

results (Attachment 3). These evaluations

have shown that the proposed value for

Reactor Coolant Minimum Flow is bounded

by the Thermal Design Flow value used in

the analyses and provides greater margin to

safety analysis acceptance criteria (e.g.,

[Departure from Nucleate Boiling] DNB). All

safety analysis acceptance criteria are

satisfied. Since Reactor Coolant flow values

for the RSG conform to the design bases and

are bounded by the existing safety analyses,

changing the technical specification within

limits of the bounding accident analyses will

not cause an increase in the probability or

consequences of an accident previously

evaluated.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change is fully consistent

with current plant design bases and does not

adversely affect any fission product barrier,

nor does it alter the safety function of safety

related systems, structures, and components

depended upon for accident prevention or

mitigation. Thus, it does not create the

possibility of a new or different kind of

accident.

(3) Involve a significant reduction in the

margin of safety.

The proposed change does not alter the

manner in which Safety Limits, Limiting

11062 Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices

Safety System Setpoints, or Limiting

Conditions for Operation are determined. It

returns TS 3.10.m for Reactor Coolant

Minimum Flow to a value slightly higher,

thus more conservative, than the value

specified for the [Original Steam Generators]

OSG when new. It conforms to plant design

bases, is consistent with current safety

analyses, and limits actual plant operation.

Analysis of the effect of the proposed Reactor

Coolant Minimum Flow limitation on [Lossof-Coolant-Accident] LOCA and non-LOCA

transients determined that all safety analysis

acceptance criteria are satisfied at a [Thermal

Design Flow] TDF that bounds the revised

Reactor Coolant Minimum Flow and all

[Kewaunee Nuclear Power Plant] KNPP

safety requirements continue to be met.

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: Bradley D.

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

Box 1497, Madison, WI 53701–1497.

NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: February

5, 2001.

Description of amendment request:

The proposed amendment would

change the Kewaunee Nuclear Power

Plant Technical Specification 3.1.d.2 to

reduce the maximum allowable leakage

of primary system reactor coolant to the

secondary system from 500 gallons per

day (gpd) through any one steam

generator to 150 gpd through any one

steam generator. In addition, the

proposed amendment would remove

reference to voltage based repair criteria.

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 change in Leakage of Reactor Coolant

value proposed by this request for [Technical

Specification] TS 3.1.d.2 complies with

[Nuclear Energy Institute] NEI 97–06, ‘‘Steam

Generator Program Guidelines.’’ [Nuclear

Management Company, LLC] NMC evaluated

accident analyses affected by [steam

generator] SG tube leakage and determined

that this change continues to be bounded by

the existing licensing and design basis.

Design basis accidents and transients,

including steam generator tube rupture

(SGTR), were analyzed using Westinghouse

Model 54F steam generator assumptions as

part of steam generator replacement. These

evaluations show that the proposed 150 gpd

[gallons per day] value for Leakage of Reactor

Coolant is bounded by the larger value used

in applicable existing design basis accident

and transient analyses. The 150 gpd leak rate

provides increased margin to acceptance

criteria found in these analyses. All

acceptance criteria are satisfied and SG

primary to secondary leakage values for the

[replacement steam generator] RSG conform

to the existing design bases and are bounded

by the existing safety analyses. Changing the

technical specification within limits of the

bounding accident analyses cannot change

the probability or consequence of an accident

previously evaluated. Removal of an

allowance for voltage-based alternate repair

criteria defaults to a more conservative repair

criteria. Thus, nothing in this proposal will

cause an increase in the probability or

consequence of an accident previously

evaluated.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The 150 gpd value proposed for maximum

allowable Leakage of Reactor Coolant is

consistent with current plant design bases

and does not adversely affect any fission

product barrier, nor does it alter the safety

function of safety significant systems,

structures and components or their roles in

accident prevention or mitigation. The

proposed value for maximum allowable

leakage through any one steam generator is

bounded by currently licensed design basis

accident and transient analyses of record.

Removal of a reference in the TS to voltagebased repair criteria leaves in its place a more

conservative, more restrictive criteria for

repair or plugging of steam generator tubes.

Thus, this proposal does not create the

possibility of a new or different kind of

accident.

(3) Involve a significant reduction in the

margin of safety.

The proposed change does not alter the

manner in which Safety Limits, Limiting

Safety System Setpoints, or Limiting

Conditions for Operation are determined. It

sets TS 3.1.d.2 for Leakage of Reactor Coolant

to a lower, thus more conservative, value

than that previously specified and approved

for use by the NRC [Nuclear Regulatory

Commission]. It conforms to plant design

bases, is consistent with current safety

analyses, and limits actual plant operation

within analyzed and licensed boundaries.

Analyses of applicable transients were

performed using a primary to secondary

leakage rate greater than the rate proposed by

this request. All safety analysis acceptance

criteria are satisfied at this value and all

[Kewaunee Nuclear Power Plant] KNPP

safety requirements continue to be met. The

150 gpd leak rate proposed by this

amendment request is bounded by these

analyses. Removal of reference to use of

voltage-based repair criteria from TS 3.1.d.2

and its basis leaves an existing and more

conservative repair criteria in place. Thus,

changes proposed by this request 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.

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: January

11, 2001.

Description of amendment requests:

The proposed amendment deletes

requirements from the Technical

Specifications (and, as applicable, other

elements of the licensing bases) to

maintain a Post Accident Sampling

System (PASS). Licensees were

generally required to implement PASS

upgrades as described in NUREG–0737,

‘‘Clarification of TMI [Three Mile

Island] Action Plan Requirements,’’ and

Regulatory Guide 1.97,

‘‘Instrumentation for Light-WaterCooled Nuclear Power Plants to Assess

Plant and Environs Conditions During

and Following an Accident.’’

Implementation of these upgrades was

an outcome of the lessons learned from

the accident that occurred at TMI, Unit

2. Requirements related to PASS were

imposed by Order for many facilities

and were added to or included in the

technical specifications (TS) for nuclear

power reactors currently licensed to

operate. Lessons learned and

improvements implemented over the

last 20 years have shown that the

information obtained from PASS can be

readily obtained through other means or

is of little use in the assessment and

mitigation of accident conditions.

The NRC staff issued a notice of

opportunity for comment in the Federal

Register on August 11, 2000 (65 FR

49271) on possible amendments to

eliminate PASS, including a model

safety evaluation and model no

significant hazards consideration

(NSHC) determination, using the

consolidated line item improvement

process. The NRC staff subsequently

issued a notice of availability of the

models for referencing in license

amendment applications in the Federal

Register on October 31, 2000 (65 FR

Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices 11063

65018). The licensee affirmed the

applicability of the following NSHC

determination in its application dated

January 11, 2001.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), an

analysis of the issue of no significant

hazards consideration is presented

below:

Criterion 1—The Proposed Change Does

Not Involve a Significant Increase in the

Probability or Consequences of an Accident

Previously Evaluated.

The PASS was originally designed to

perform many sampling and analysis

functions. These functions were designed

and intended to be used in post accident

situations and were put into place as a result

of the TMI–2 accident. The specific intent of

the PASS was to provide a system that has

the capability to obtain and analyze samples

of plant fluids containing potentially high

levels of radioactivity, without exceeding

plant personnel radiation exposure limits.

Analytical results of these samples would be

used largely for verification purposes in

aiding the plant staff in assessing the extent

of core damage and subsequent offsite

radiological dose projections. The system

was not intended to and does not serve a

function for preventing accidents and its

elimination would not affect the probability

of accidents previously evaluated.

In the 20 years since the TMI–2 accident

and the consequential promulgation of post

accident sampling requirements, operating

experience has demonstrated that a PASS

provides little actual benefit to post accident

mitigation. Past experience has indicated that

there exists in-plant instrumentation and

methodologies available in lieu of a PASS for

collecting and assimilating information

needed to assess core damage following an

accident. Furthermore, the implementation of

Severe Accident Management Guidance

(SAMG) emphasizes accident management

strategies based on in-plant instruments.

These strategies provide guidance to the

plant staff for mitigation and recovery from

a severe accident. Based on current severe

accident management strategies and

guidelines, it is determined that the PASS

provides little benefit to the plant staff in

coping with an accident.

The regulatory requirements for the PASS

can be eliminated without degrading the

plant emergency response. The emergency

response, in this sense, refers to the

methodologies used in ascertaining the

condition of the reactor core, mitigating the

consequences of an accident, assessing and

projecting offsite releases of radioactivity,

and establishing protective action

recommendations to be communicated to

offsite authorities. The elimination of the

PASS will not prevent an accident

management strategy that meets the initial

intent of the post-TMI–2 accident guidance

through the use of the SAMGs, the

emergency plan (EP), the emergency

operating procedures (EOP), and site survey

monitoring that support modification of

emergency plan protective action

recommendations (PARs).

Therefore, the elimination of PASS

requirements from Technical Specifications

(TS) (and other elements of the licensing

bases) does not involve a significant increase

in the consequences of any accident

previously evaluated.

Criterion 2—The Proposed Change Does

Not Create the Possibility of a New or

Different Kind of Accident from any

Previously Evaluated.

The elimination of PASS related

requirements will not result in any failure

mode not previously analyzed. The PASS

was intended to allow for verification of the

extent of reactor core damage and also to

provide an input to offsite dose projection

calculations. The PASS is not considered an

accident precursor, nor does its existence or

elimination have any adverse impact on the

pre-accident state of the reactor core or post

accident confinement of radionuclides

within the containment building.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3—The Proposed Change Does

Not Involve a Significant Reduction in the

Margin of Safety.

The elimination of the PASS, in light of

existing plant equipment, instrumentation,

procedures, and programs that provide

effective mitigation of and recovery from

reactor accidents, results in a neutral impact

to the margin of safety. Methodologies that

are not reliant on PASS are designed to

provide rapid assessment of current reactor

core conditions and the direction of

degradation while effectively responding to

the event in order to mitigate the

consequences of the accident. The use of a

PASS is redundant and does not provide

quick recognition of core events or rapid

response to events in progress. The intent of

the requirements established as a result of the

TMI–2 accident can be adequately met

without reliance on a PASS.

Therefore, this change does not involve a

significant reduction in the margin of safety.

Based upon the reasoning presented above

and the previous discussion of the

amendment request, the requested change

does not involve a significant hazards

consideration.

The NRC staff proposes to determine

that the amendment request involves no

significant hazards consideration.

Attorney for licensee: Douglas K.

Porter, Esquire, Southern California

Edison Company, 2244 Walnut Grove

Avenue, Rosemead, California 91770.

NRC Section Chief: Stephen Dembek.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

January 18, 2001.

Brief description of amendment

request: The amendment request

identifies an unreviewed safety question

related to the planned replacement of

the engineered safety features (ESF)

transformers with new transformers

having active automatic load tap

changers (LTCs). Markups to the Final

Safety Analysis Report (FSAR) were

included in the application.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

1. Does the proposed change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

Based on the review of the modification

details there is an insignificant increase in

the probability of a malfunction of equipment

important to safety, however there is no

increase in the probability of an accident

previously evaluated. The modification has

no effect on the radiological consequences of

accidents previously evaluated. Installation

of the LTCs does not impact accident

initiators though a failure mode has been

identified that can increase the probability of

malfunction, a risk study shows this risk is

insignificant.

2. Does the proposed change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The overall effect of the malfunction of the

LTC controllers would lead to a loss of the

associated ESF bus which is not a new failure

mode that can lead to a new or different kind

of accident than previously evaluated. The

LTC failure effects are limited to the

associated ESF train, therefore this type of

failure meets the definition of a single failure

as defined in 10 CFR 50 Appendix A for

operation under normal (Non T/S [technical

specification] action) conditions.

Additionally, during the 10 CFR 50.59

evaluation for the LTCs criteria (a)(2)(ii) with

respect to accidents of a different type was

not met.

3. Does the proposed change involve a

significant reduction in margin of safety?

The installation of the replacement

transformers with load tap changers will help

assure the required minimum NB bus voltage

established by Reference 7.10 [design

calculations] under a wider variation of grid

voltage.

Current Technical Specification Bases for

the offsite power distribution system are

covered in sections B3.8.1–AC SourcesOperating, B3.8.9–Distribution SystemsOperating, B3.8.2–AC Sources-Shutdown,

and B3.8.10–Distribution Systems-Shutdown.

These bases ensure that sufficient power will

be available to supply the safety-related

equipment required for: (1) The safe

shutdown of the facility; and (2) The

mitigation and control of accident conditions

within the facility. The minimum specified

independent and redundant AC power and

distribution systems satisfy the requirements

of General Design Criterion 17 of Appendix

A to 10 CFR Part 50. The ACTIONS sections

of the applicable Technical Specifications

provide requirements specified for various

levels of degradation of the power sources

and provide restrictions upon continued

facility operation commensurate with the

11064 Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices

level of degradation. The Operability of the

power sources are consistent with the initial

condition assumptions of the safety analyses

and are based upon maintaining at least one

redundant set of onsite AC power sources

and associated distribution systems operable

during accident conditions coincident with

an assumed loss of offsite power and single

failure of the other onsite AC source.

The installation of the transformers with

automatic load tap changers reduces the

possibility of the loss of the offsite power

system due to the increased grid voltage

variations as documented in the description

of the change in section 4.1.4. Therefore, the

installation of the transformers with load tap

changers will not reduce the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: John O’Neill,

Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Stephen Dembek.

Virginia Electric and Power Company,

Docket No. 50–338, North Anna Power

Station, Unit No. 1, Louisa County,

Virginia

Date of amendment request: January

9, 2001.

Description of amendment request:

The proposed administrative changes

will remove obsolete license conditions

from the Facility Operating License

(FOL) and implement associated

changes to the Technical Specifications

(TS). These changes involve editorial

revisions, relocation of license

conditions, removal of redundant

license conditions covered throughout

the license, removal of expired license

conditions, and removal of license

conditions and TS associated with

completed modifications.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1—The proposed license

amendment does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed change to the North Anna

Unit 1 Facility Operating License, NPF–4, is

administrative (and in part editorial) in

nature. The removal of license conditions

regarding completed, no longer needed, and

expired requirements has no impact on plant

operations since these requirements no

longer have meaningful applications. The

renumbering and/or relocation within the

FOL of various license conditions in this

proposed administrative change does not

alter the technical basis, requirements or the

implementation of the affected items. The

proposed change is within the current design

and licensing bases of the facility. Since this

change is administrative only and neither

station operations nor design are affected by

the change, it does not involve any

significant 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 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.

Consequently, no new failure modes are

introduced and the proposed administrative

change to the North Anna Unit 1 Facility

Operating License 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. Since station operations are not

affected by the proposed administrative

change and no physical change is being made

to the station, the change does not impact the

condition, design, or performance of any

station structure, system or component.

Therefore, the proposed administrative

change to the North Anna Unit 1 Facility

Operating License does not involve a

significant 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: Mr. Donald P.

Irwin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Section Chief: Maitri Banerjee,

Acting.

Virginia Electric and Power Company,

Docket No. 50–339, North Anna Power

Station, Unit No. 2, Louisa County,

Virginia

Date of amendment request: January

9, 2001.

Description of amendment request:

The proposed administrative changes

will remove obsolete license conditions

from the Facility Operating License

(FOL). These changes involve editorial

revisions, relocation of license

conditions, removal of expired license

conditions, and removal of license

conditions associated with completed

modifications.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Criterion 1—The proposed license

amendment does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed change to the North Anna

Unit 2 Facility Operating License, NPF–7, is

administrative (and in part editorial) in

nature. The removal of license conditions

regarding completed, no longer needed, and

expired requirements has no impact on plant

operations since these requirements no

longer have meaningful applications. The

renumbering and/or relocation within the

FOL of various license conditions in this

proposed administrative change does not

alter the technical basis, requirements or the

implementation of the affected items. The

proposed change is within the current design

and licensing bases of the facility. Since this

change is administrative only and neither

station operations nor design are affected by

the change, it does not involve any

significant 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 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.

Consequently, no new failure modes are

introduced and the proposed administrative

change to the North Anna Unit 2 Facility

Operating License does not create the

possibility of a new or different kind of

accident or malfunction of equipment

Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices 11065

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. Since station operations are not

affected by the proposed administrative

change and no physical change is being made

to the station, the change does not impact the

condition, design, or performance of any

station structure, system or component.

Therefore, the proposed administrative

change to the North Anna Unit 2 Facility

Operating License does not involve a

significant 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: Mr. Donald P.

Irwin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Section Chief: Maitri Banerjee,

Acting.

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.

PSEG Nuclear LLC, Docket No. 50–354,

Hope Creek Generating Station, Salem

County, New Jersey

Date of amendment request: January

8, 2001.

Brief description of amendment

request: The proposed amendment

would revise the Technical

Specifications (TS) to change the

acceptance values for Core Spray

subsystem flow contained in TS

4.5.1.b.1 from the current value of 6350

gallons per minute (gpm) to 6150 gpm.

Date of publication of individual

notice in Federal Register: January 22,

2001 (66 FR 6701).

Expiration date of individual notice:

February 21, 2001.

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 and

electronically from the ADAMS Public

Library component on the NRC Web

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

Reading Room).

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units Nos. 1, 2, and

3, Maricopa County, Arizona

Date of application for amendments:

June 16, 2000.

Brief description of amendments: The

amendments revise TS Table 3.3.10–1,

‘‘Post Accident Monitoring

Instrumentation,’’ to add the high

pressure safety injection (HPSI) cold leg

flow and HPSI hot leg flow

instrumentation to the table.

Date of issuance: February 8, 2001.

Effective date: February 8, 2001, and

shall be implemented within 30 days of

the date of issuance.

Amendment Nos.: Unit 1—131 , Unit

2—131, Unit 3—131.

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 4, 2000 (65 FR 59220)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 08,

2001.

No significant hazards consideration

comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc.,

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

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of application for amendments:

September 14, 2000.

Brief description of amendments: The

amendments incorporate changes

described below into the Technical

Specifications for Culvert Cliffs Units 1

and 2. On September 9, 1996, a final

rule amending 10 CFR 50.55a was

issued requiring owners to implement,

by September 9, 2001, the requirements

of the 1992 Edition through the 1992

Addenda of the American Society of

Mechanical Engineers Boiler and

Pressure Vessel Code Section XI,

Subsections IWE and IWL, as modified

and supplemented by 10 CFR 50.55a.

Calvert Cliffs Nuclear Power Plant, Inc.

has developed a program to effect the

implementation of Subsections IWE and

IWL.

Date of issuance: January 30, 2001.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 240 and 214.

Facility Operating License Nos. DPR–

53 and DPR–69: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 18, 2000 (65 FR

62384).

11066 Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices

The Commission’s related evaluation

of these amendments is contained in a

Safety Evaluation dated January 30,

2001.

No significant hazards consideration

comments received: No.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of application for amendment:

November 19, 1999, as supplemented

May 31, August 2, October 19, and

November 21, 2000.

Brief description of amendment: The

amendment revises the Technical

Specifications (TSs) by changing (1) the

design features description of the fuel

storage equipment and configuration to

allow an increase in the spent fuel pool

(SFP) storage capacity and (2) the

description of the high-density spent

fuel racks program to clarify that the

surveillance program is applicable only

to racks containing Boraflex as a

neutron absorber. Specifically, the

amendment revises the TSs for Fermi 2

to increase the capacity of the SFP from

2,414 to 4,608 fuel assemblies.

Date of issuance: January 25, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 90 days.

Amendment No.: 141.

Facility Operating License No. NPF–

43: Amendment revises the Technical

Specifications

Date of initial in Federal Register

March 13, 2000 (65 FR 13336)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated a January 25,

2001.

No significant hazards consideration

comments received: No.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of application for amendment:

September 20, 2000.

Brief description of amendment: The

amendment changes Technical

Specification (TS) 5.5.7.d to decrease

the maximum allowed pressure drops

across control room emergency filtration

(CREF) make-up and recirculation train

filters and charcoal absorbers. The

words ‘‘(CREF only)’’ are also removed

from the TS.

Date of issuance: February 8, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 60 days

Amendment No.: 142.

Facility Operating License No. NPF–

43: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: November 1, 2000 (65 FR

65340).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 8,

2001.

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:

November 30, 2000.

Brief description of amendment: The

amendment relocated the boration

systems requirements from the

Technical Specifications to the

Technical Requirements Manual.

Date of issuance: January 31, 2001.

Effective date: As of the date of

issuance to be implemented within 60

days from the date of issuance.

Amendment No.: 229.

Facility Operating License No. NPF–6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 27, 2000 (65 FR

81916).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 31,

2001.

No significant hazards consideration

comments received: No.

Entergy Nuclear Operations, Inc.,

Docket No. 50–333, James A. FitzPatrick

Nuclear Power Plant, Oswego County,

New York

Date of application for amendment:

November 19, 1999, as supplemented

October 12, 2000.

Brief description of amendment: The

amendment changes the Technical

Specification surveillance testing

requirements of the charcoal adsorbers

in the Standby Gas Treatment System

and the Control Room Emergency

Ventilation Air Supply System to meet

the requested actions of Generic Letter

99–02.

Date of issuance: February 5, 2001.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 269.

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 October 12, 2000, supplemental

letter provided clarifying information

that did not change the initial proposed

no significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 5,

2001.

No significant hazards consideration

comments received: No.

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant, Units 3 and 4, Dade County,

Florida

Date of application for amendments:

May 22, 2000, as supplemented October

4, 2000.

Brief description of amendments:

Changed the Technical Specifications to

incorporate that portion of the August 8,

1996, Final Amended Rule (61 FR

41303) related to revised requirement of

inservice inspection of the containment

post-tensioning system.

Date of issuance: January 31, 2001.

Effective date: January 31, 2001.

Amendment Nos. 210 and 204.

Facility Operating License Nos. DPR–

31 and DPR–41: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: August 9, 2000 (65 FR 48750).

The October 4, 2000 letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 31,

2001.

No significant hazards consideration

comments received: No.

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant, Units 3 and 4, Dade County,

Florida

Date of application for amendments:

December 6, 2000.

Brief description of amendments: The

amendments delete Technical

Specifications (TS) Section 6.8.4.d,

‘‘Post Accident Sampling,’’ for Turkey

Point Units 3 and 4 and thereby

eliminate the requirements to have and

maintain the post-accident sampling

system (PASS) for those units.

Date of issuance: January 31, 2001.

Effective date: January 31, 2001.

Amendment Nos. 211 and 205.

Facility Operating License Nos. DPR–

31 and DPR–41: Amendments revised

the TSs.

Date of initial notice in Federal

Register: December 27, 2000 (65 FR

81923).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 31,

2001.

No significant hazards consideration

comments received: No.

Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices 11067

Indiana Michigan Power Company,

Docket No. 50–316, Donald C. Cook

Nuclear Plant, Unit 2, Berrien County,

Michigan

Date of application for amendment:

September 30, 2000, as supplemented

November 22, and December 20, 2000.

Brief description of amendment: The

amendment would allow an extension

of the steam generator tube inspection

surveillance requirements of Technical

Specification Surveillance Requirement

4.4.5.3. Specifically, the licensee

requested to extend the required

inspection interval from 40 to 56

calendar months.

Date of issuance: January 30, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 232.

Facility Operating License No. DPR–

74: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: October 18, 2000 ( 65 FR

62387).

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 January 30,

2001.

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:

June 1, 2000.

Brief description of amendment: The

amendment approves changes to

Technical Specifications (TSs) 3.3.3.2,

‘‘Instrumentation, Movable Incore

Detectors’’; 3.3.3.3, ‘‘Instrumentation,

Seismic Instrumentation’’; 3.3.3.4,

‘‘Instrumentation, Meteorological

Instrumentation’’; 3.3.3.8, ‘‘Loose-Part

Detection System’’; 3.3.4, ‘‘Turbine

Overspeed Protection’’; and Index Pages

vi and vii. The changes relocate the

requirements for the incore detectors,

seismic instrumentation, meteorological

instrumentation, loose-part detection

system, and turbine overspeed

protection system from the TSs to the

Technical Requirements Manual. The

Bases for these TSs have been modified

to reflect the TS changes.

Date of issuance: January 29, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 60 days from the date of

issuance.

Amendment No.: 193.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 29, 2000 (65 FR

71136).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 29,

2001.

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:

July 31, 2000 as supplemented January

5, 2001.

Brief description of amendment: The

amendment changes Technical

Specifications (TSs) 3.8.1.1, ‘‘Electrical

Power Systems—A.C. Sources—

Operating,’’ and 3.8.1.2, ‘‘Electrical

Power Systems—A.C. Sources—

Shutdown.’’ The changes allow certain

EDG surveillance requirements to be

performed when the plant is operating

instead of shut down as currently

required. The index and Bases for these

TSs are modified to reflect the changes.

Date of issuance: February 2, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 194.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 6, 2000 (65 FR

54087).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 2,

2001.

No significant hazards consideration

comments received: No.

Nuclear Management Company, LLC,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

November 18, 1999, as supplemented

August 7, 2000.

Brief description of amendment: The

amendment to the Kewaunee Nuclear

Power Plant Technical Specifications

approves an increase in the allowable

number of spent fuel assemblies in the

spent fuel pools. The addition of the 215

storage locations in the new north canal

pool will extend the full-core reserve

capability until after the 2009 outage,

and increase the total capacity to 1,205

spent fuel assemblies.

Date of issuance: January 23, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 150.

Facility Operating License No. DPR–

43: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 1 and December

21, 2000 (65 FR 65347 and 65 FR 80471

respectively)

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 January 23, 2001.

No significant hazards consideration

comments received: No.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

May 12, 2000, as supplemented by letter

dated January 25, 2001.

Brief description of amendments:

These amendments authorize (1) a

design upgrade of the refueling water

purification (RWP) system to permit

reclassification of this system from

Design Class II/non-Seismic Category 1

to Design Class I/Seismic Category 1 to

allow filtering of the refueling water

storage tank (RWST) water while the

RWST is required to be operable, and (2)

the use of a temporary reverse osmosis

skid mounted system to reduce RWST

silica concentration levels while the

RWST is required to be operable

following upgrade of the RWP system to

satisfy reactor coolant chemistry limits.

Date of issuance: January 29, 2001.

Effective date: January 29, 2001, and

shall be implemented in the next

periodic update to the FSAR Update,

following upgrade of the refueling water

purification system, in accordance with

10 CFR 50.71(e).

Amendment Nos.: Unit 1—144 ; Unit

2–143.

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments

revised the FSAR Update.

Date of initial notice in Federal

Register: July 12, 2000 (65 FR 43050).

The January 25, 2001, supplemental

letter provided additional clarifying

information, did not expand the scope

of the application as originally noticed,

11068 Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices

and did not change the staff’s original

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 29,

2001.

No significant hazards consideration

comments received: No.

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:

September 6, 2000 (PCN–274,

Supplement 1).

Brief description of amendments: The

amendments revised the San Onofre,

Units 2 and 3 Technical Specification

3.3.11, ‘‘Post Accident Monitoring

Instrumentation (PAMI),’’ to extend the

PAMI surveillance frequency from 18 to

24 months to accommodate a 24-month

fuel cycle.

Date of issuance: January 30, 2001.

Effective date: January 30, 2001, to be

implemented within 30 days of

issuance.

Amendment Nos.: Unit 2—176; Unit

3–167.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 18, 2000 (65 FR

62391).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 30,

2001.

No significant hazards consideration

comments received: No.

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:

October 6, 2000 (PCN–518).

Brief description of amendments: The

amendments revise TS 3.7.11, ‘‘Control

Room Emergency Air Cleanup System

(CREACUS),’’ to establish actions to be

taken for inoperable ventilation systems

due to a degraded control room pressure

boundary. The amendments allow up to

24 hours to restore the pressure

boundary to operable status when two

ventilation trains are inoperable due to

an inoperable pressure boundary in

Modes 1, 2, 3, and 4. In addition, a

limiting condition for operation note is

added to allow the pressure boundary to

be opened intermittently under

administrative control without affecting

CREACUS operability.

Date of issuance: January 30, 2001.

Effective date: January 30, 2001, to be

implemented within 30 days of

issuance.

Amendment Nos.: Unit 2—177; Unit

3—168.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 15, 2000 (65 FR

69066).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 30,

2001.

No significant hazards consideration

comments received: No.

South Carolina Electric & Gas Company,

South Carolina Public Service

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

Summer Nuclear Station, Unit No. 1,

Fairfield County, South Carolina

Date of application for amendment:

November 10, 2000.

Brief description of amendment: This

amendment will allow: (a) the minimum

fuel oil stored in the fuel oil storage tank

(FOST) for each emergency diesel

generator (EDG) to be raised from 47,100

gallons to 48,500 gallons for Modes 1–

4, and from 33,200 gallons to 42,500

gallons for Modes 5 and 6; and (b) the

minimum fuel oil maintained in the day

fuel tank for each EDG to be raised from

300 gallons to 360 gallons for Modes 1–

6.

Date of issuance: February 2, 2001.

Effective date: February 2, 2001.

Amendment No.: 150.

Facility Operating License No. NPF–

12: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: November 20, 2000 (65 FR

69795).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 2,

2001.

No significant hazards consideration

comments received: No.

Southern Nuclear Operating Company,

Inc., Docket Nos. 50–348 and 50–364,

Joseph M. Farley Nuclear Plant, Units 1

and 2, Houston County, Alabama

Date of amendments request: October

9, 2000, supplemented December 4,

2000.

Brief Description of amendments: The

amendments revise Technical

Specification 5.5.14, ‘‘Technical

Specification (TS) Bases Control

Program,’’ to provide consistency with

the changes in 10 CFR 50.59 which were

published in the Federal Register on

October 4, 2000.

Date of issuance: January 31, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 148 and 140.

Facility Operating License Nos. NPF–

2 and NPF–8: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: December 20, 2000 (65 FR

79907) The supplement dated December

4, 2000, provided clarifying information

that did not change the scope of the

October 4, 2000, application nor the

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 31,

2001.

No significant hazards consideration

comments received: No.

Southern Nuclear Operating Company,

Inc., et al., Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

June 14, 2000.

Brief description of amendments: The

amendments revised the Technical

Specifications (TSs) Surveillance

Requirements (SR) 3.8.1.9 and 3.8.1.14

to reduce diesel generators loading

requirements from ≥ 6800 kW and ≤

7000 kW to ≥ 6500 kW and ≤ 7000 kW.

These changes will make the above SRs

consistent with SRs 3.8.1.3 and 3.8.1.13,

which are in the current TSs. In

addition, the proposed changes would

correct a typographical error in Section

5.6.7, ‘‘EDG Failure Report,’’ in the

Vogtle TS. This editorial change will

correctly reference Regulatory Position

C.4 of Regulatory Guide 1.9, Revision 3

instead of Regulatory Position C.5.

Date of issuance: January 31, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: Unit 1—117; Unit

2—95.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 6, 2000 (65 FR

54087).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated January 31,

2001.

No significant hazards consideration

comments received: No.

Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices 11069

Tennessee Valley Authority, Docket

Nos. 50–260 and 50–296, Browns Ferry

Nuclear Plant, Units 2 and 3, Limestone

County, Alabama

Date of application for amendments:

August 11, 2000 (TS–400) as

supplemented by letter dated October

20, 2000.

Brief description of amendments: The

amendments revised the surveillance

test requirements for excess flow check

valves.

Date of issuance: January 29, 2001.

Effective date: January 29, 2001.

Amendment Nos.: 268 and 228.

Facility Operating License Nos. DPR–

52 and DPR–68: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 6, 2000 (65 FR

54088). The October 20, 2000, letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated January 29,

2001.

No significant hazards consideration

comments received: No.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request:

December 7, 2000 (ET 00–0041).

Brief description of amendment: The

amendment changes Table 3.3.2–1,

‘‘Engineered Safety Feature Actuation

System Instrumentation,’’ of the

Technical Specifications. The change

adds Surveillance Requirement (SR)

3.3.2.10 for the following two

engineered safety feature actuation

system instrumentation in the table:

item 6.f, loss of offsite power, and item

6.h, auxiliary feedwater pump suction

transfer on suction pressure—low.

Date of issuance: February 06, 2001.

Effective date: February 06, 2001, and

shall be implemented including the

changes to the Bases for the response

times, within 60 days of the date of

issuance.

Amendment No.: 136.

Facility Operating License No. NPF–

42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 27, 2000 (65 FR

81932).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 06,

2001.

No significant hazards consideration

comments received: No.

Notice of Issuance of Amendments to

Facility Operating Licenses and Final

Determination of No Significant

Hazards Consideration and

Opportunity for a Hearing (Exigent

Public Announcement or Emergency

Circumstances)

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act of 1954, as amended

(the Act), and the Commission’s rules

and regulations. The Commission has

made appropriate findings as required

by the Act and the Commission’s rules

and regulations in 10 CFR Chapter I,

which are set forth in the license

amendment.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the application for

amendment, (2) the amendment to

Facility Operating License, and (3) the

Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment, as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, located at One White Flint North,

11555 Rockville Pike (first floor),

Rockville, Maryland 20852, and

electronically from the ADAMS Public

Library component on the NRC Web

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

Reading Room).

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By

March 23, 2001, 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

11070 Federal Register / Vol. 66, No. 35 / Wednesday, February 21, 2001 / Notices

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,

and electronically from the ADAMS

Public Library component on the NRC

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

Electronic Reading Room). If a request

for a hearing or petition for leave to

intervene is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–001, Attention:

Rulemakings and Adjudications Staff, or

may be delivered to the Commission’s

Public Document Room, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852,

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 20555–001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of the

factors specified in 10 CFR

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

Entergy Nuclear Operations, Inc.,

Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request:

December 19, 2000.

Description of amendment request:

The amendment revises the Technical

Specifications to indicate that quadrant

power tilt limits apply only when

reactor power is greater than 50 percent.

Date of issuance: December 20, 2000.

Effective Date: As of its date of

issuance and shall be implemented

within 30 days.

Amendment No.: 204.

Facility Operating License No. DPR–

64: Amendment revises the Technical

Specifications.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendment, finding of emergency

circumstances, and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated December 20, 2000.

Attorney for licensee: Mr. John M.

Fulton, Assistant General Counsel

Entergy Nuclear Generating Co. Pilgrim

Station, 600 Rocky Hill Road Plymouth,

MA 02360.

NRC Section Chief: Marsha

Gamberoni.

Dated at Rockville, Maryland, this 14th day

of February 2001.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–4228 Filed 2–20–01; 8:45 am]

BILLING CODE 7590–01–U

SECURITIES AND EXCHANGE

COMMISSION

Proposed Collection; Comment

Request

Upon Written Request, Copies Available

From: Securities and Exchange

Commission, Office of Filings and

Information Services, Washington, DC

20549.

Extension:

Form SE, OMB Control No. 3235–

0327, SEC File No. 270–289;

Form ID, OMB Control No. 3235–

0328, SEC File No. 270–291;

Form ET, OMB Control No. 3235–

0329, SEC File No. 270–290; and

Form TH, OMB Control No. 3235–

0425, SEC File No. 270–377.

Notice is hereby given that pursuant

to the Paperwork Reduction Act of 1995

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

and Exchange Commission

(‘‘Commission’’) is soliciting comments

on the collections of information

summarized below. The Commission

plans to submit these existing

collections of information to the Office

of Management and Budget for

extension and approval.