Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices 22021

(DEIS) for Yucca Mountain (tentative)

(Open)—The Committee will receive

an information briefing from the NRC

staff on their plans to review the DOE

DEIS for the proposed HLW

repository at Yucca Mountain,

Nevada.

I. 2:00–3 p.m.: Break and Preparation of

Draft ACNW Reports (Open)—

Cognizant ACNW members will

prepare draft reports, as needed, for

consideration by the full Committee.

J. 3:00–5 p.m.: Discussion of Proposed

ACNW Reports (Open)—The

Committee will continue its

discussion of proposed ACNW

reports.

Thursday, May 17, 2001

K. 8:30–8:35 a.m.: Opening Remarks by

the ACNW Chairman (Open)—The

ACNW Chairman will make opening

remarks regarding the conduct of the

meeting.

L. 8:35–10 a.m.: Meeting Reports

(Open)—The Committee will hear

reports from the members and staff on

meetings attended since the 125th

ACNW Meeting, including the

National Research Council Meeting on

their report on long-term institutional

control, the 9th International HLW

Conference and the Nuclear Waste

Technical Review Board Spring

Meeting.

M. 10:15–12 Noon: Discussion of

Proposed ACNW Reports (Open)—

The Committee will continue its

discussion of proposed ACNW

reports.

N. 1:00–1:30 p.m.: Miscellaneous

(Open)—The Committee will discuss

matters related to the conduct of

Committee activities and matters and

specific issues that were not

completed during previous meetings,

as time and availability of information

permit.

Procedures for the conduct of and

participation in ACNW meetings were

published in the Federal Register on

October 11, 2000 (65 FR 60475). In

accordance with these procedures, oral

or written statements may be presented

by members of the public, electronic

recordings will be permitted only

during those portions of the meeting

that are open to the public, and

questions may be asked only by

members of the Committee, its

consultants, and staff. Persons desiring

to make oral statements should notify

Howard J. Larson, ACNW, as far in

advance as practicable so that

appropriate arrangements can be made

to schedule the necessary time during

the meeting for such statements. Use of

still, motion picture, and television

cameras during this meeting will be

limited to selected portions of the

meeting as determined by the ACNW

Chairman. Information regarding the

time to be set aside for taking pictures

may be obtained by contacting the

ACNW office, prior to the meeting. In

view of the possibility that the schedule

for ACNW meetings may be adjusted by

the Chairman as necessary to facilitate

the conduct of the meeting, persons

planning to attend should notify Mr.

Larson as to their particular needs.

Further information regarding topics

to be discussed, whether the meeting

has been canceled or rescheduled, the

Chairman’s ruling on requests for the

opportunity to present oral statements

and the time allotted therefore can be

obtained by contacting Mr. Howard J.

Larson, ACNW (Telephone 301/415–

6805), between 8 A.M. and 5 P.M. EDT.

ACNW meeting notices, meeting

transcripts, and letter reports are now

available for downloading or viewing on

the internet at http://www.nrc.gov/

ACRSACNW.

Videoteleconferencing service is

available for observing open sessions of

ACNW meetings. Those wishing to use

this service for observing ACNW

meetings should contact Mr. Theron

Brown, ACNW Audiovisual Technician

(301/415–8066), between 7:30 a.m. and

3:45 p.m. EDT at least 10 days before the

meeting to ensure the availability of this

service. Individuals or organizations

requesting this service will be

responsible for telephone line charges

and for providing the equipment and

facilities that they use to establish the

videoteleconferencing link. The

availability of videoteleconferencing

services is not guaranteed.

Dated: April 26, 2001.

Andrew L. Bates,

Advisory Committee Management Officer.

[FR Doc. 01–10964 Filed 5–1–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from April 9,

2001, through April 20, 2001. The last

biweekly notice was published on April

18, 2001 (66 FR 19998).

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

22022 Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review and

Directives Branch, Division of Freedom

of 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, located at One White

Flint North, 11555 Rockville Pike (first

floor), Rockville, Maryland 20852. The

filing of requests for a hearing and

petitions for leave to intervene is

discussed below.

By June 1, 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, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852.

Publicly available records will be

accessible and electronically from the

ADAMS Public Library component on

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

(the Electronic Reading Room). If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) The nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

Those permitted to intervene become

parties to the proceeding, subject to any

limitations in the order granting leave to

intervene, and have the opportunity to

participate fully in the conduct of the

hearing, including the opportunity to

present evidence and cross-examine

witnesses.

If a hearing is requested, the

Commission will make a final

determination on the issue of no

significant hazards consideration. The

final determination will serve to decide

when the hearing is held.

If the final determination is that the

amendment request involves no

significant hazards consideration, the

Commission may issue the amendment

and make it immediately effective,

notwithstanding the request for a

hearing. Any hearing held would take

place after issuance of the amendment.

If the final determination is that the

amendment request involves a

significant hazards consideration, any

hearing held would take place before

the issuance of any amendment.

A request for a hearing or a petition

for leave to intervene must be filed with

the Secretary of the Commission, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, Attention:

Rulemaking and Adjudications Branch,

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 20555–0001, and to the

attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR

2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, located at One

White Flint North, 11555 Rockville Pike

(first floor), Rockville, Maryland 20852.

Publicly available records will be

accessible and electronically from the

ADAMS Public Library component on

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

(the Electronic Reading Room).

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of amendments request: April 1,

2001 (102–04552).

Description of amendments request:

The amendments would revise the

requirements on the following programs

in the administrative controls section of

the technical specifications (TSs): (1)

Section 5.5.13, ‘‘Diesel Fuel Oil Testing

Program,’’ (2) Section 5.5.14, ‘‘TS Bases

Control Program,’’ (3) Section 5.5.15,

‘‘Safety Functions Determination

Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices 22023

Program (SFDP),’’ and (4) Section 5.6.5,

‘‘Core Operating Limits Report (COLR).’’

The proposed changes clarify the

program requirements in Section 5.5.13

without changing testing methods or

limits, revise the program in Section

5.5.14 based on changes to 10 CFR 50.59

in the regulations, clarify the program

requirements in Section 5.5.15

including changing the program name to

the plant-specific name for the program,

and add the CENTS code to the list of

analytical methods used, including the

use of CENTS for control element

assembly ejection analyses, to determine

core operating limits and revise the list

of referenced topical reports in the

COLR in Section 5.6.5.

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, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Technical Specification (TS) 5.5.13, Diesel

Generator Fuel Oil Program. TS 5.5.13.a.3

currently states, ‘‘Water and sediment are

within the limits of ASTM D1796,’’ for the

acceptability of new diesel fuel oil. This is

an incorrect reference for the limits of water

and sediment of new fuel oil. The water and

sediment limits for new fuel oil are contained

within the Technical Specification Bases.

ASTM D1796 contains testing methods used

for analysis of new fuel oil for water and

sediment. This proposed amendment

changes the wording of TS 5.5.13.a.3 to state,

‘‘Water and sediment within limits when

tested in accordance with ASTM D1796.’’

This proposed change is an administrative

change and will have no affect on plant

design, operation, or maintenance.

Additionally, this proposed change does not

result in any hardware changes or affect plant

operating practices. The water and sediment

testing methods and limits are not affected by

this change. Thus, this proposed change does

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

TS 5.5.14, TS Bases Control Program,

requires a program for processing changes to

the Bases of the TS [...]

In the initial sentence to TS 5.5.14.b, the

word ‘‘involve’’ will be replaced with

‘‘require.’’ Additionally, the second

allowance for changing TS Bases as described

in TS 5.5.14.b will be revised to state, ‘‘A

change to the updated FSAR or Bases that

requires NRC approval pursuant to 10 CFR

50.59.’’ This change is based on the changes

to 10 CFR 50.59 published in the Federal

Register (Volume 64, Number 191) dated

October 4, 1999. This change is consistent

with NRC approved Technical Specifications

Task Force (TSTF) traveler number 364-

revision 0.

This change will also numerically format

the two options listed in TS 5.5.14.b. This is

consistent with other listings contained in

Section 5.0 of the TS.

This proposed change deletes the reference

to ‘‘unreviewed safety question’’ as

previously used in 10 CFR 50.59[, before the

rule change published October 4, 1999, in the

Federal Register.] Deletion of this definition

was approved by the NRC with the revision

to 10 CFR 50.59.

[These] proposed change[s to TS 5.5.14 are]

administrative change[s] and will have no

affect on plant design, operation, or

maintenance. Additionally, [these] change[s

do] not result in any hardware changes or

affect plant operating practices. Therefore,

[the] proposed change[s to TS 5.5.14 do] not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

TS 5.5.15, Safety Functions Determination

Program (SFDP). Clarification is being added

to TS 5.5.15. The second paragraph of TS

5.5.15 will be changed to read: ‘‘A loss of

safety function exists when, assuming no

concurrent single failure, no concurrent loss

of offsite power, or no concurrent loss of

onsite diesel generator(s), a safety function

assumed in the accident analysis cannot be

performed. For the purpose of this program,

a loss of safety function may exist when a

support system is inoperable, and * * *’’

An additional paragraph will be added to

the end of TS 5.5.15 stating, ‘‘When a loss of

safety function is caused by the inoperability

of a single Technical Specification support

system, the appropriate Conditions and

Required Actions to enter are those of the

support system.’’

Additionally, clarification will be added to

limiting conditions for operation (LCO) 3.0.6

Bases of the ‘‘appropriate LCO for loss of

safety function.’’ The Bases will also clarify

the requirement for the SFDP that

consideration does not have to be made for

a loss of power in determining loss of

function. This change is consistent with NRC

approved TSTF traveler number 273-revision

2, as amended by editorial change WOG–ED–

23.

In addition, an editorial change to remove

the ‘‘s’’ from the word ‘‘Functions’’ in the

title for TS 5.5.15 will occur. The change

reflects the plant specific name for this

program.

[These] proposed change[s to TS Section

5.5.15 are] administrative change[s] and will

have no affect on plant design, operation, or

maintenance. The change[s] clarif[y] the

requirements for determining loss of safety

function and the correct LCO to enter for loss

of safety function. The proposed change[s do]

not result in any hardware changes or affect

plant operating practices. The program will

still determine when a safety function has

been lost and will direct the appropriate

action. Therefore, [the] proposed change[s to

TS 5.5.15 do] not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

TS 5.6.5, Core Operating Limits Report

(COLR) is being revised to add the option to

use the CENTS computer code in licensing

analysis by adding CENTS to the list of

approved core operating limit analytical

methods contained in TS 5.6.5.b. The CENTS

computer code has been generally approved

for the calculation of transient behavior in

Pressurized Water reactors (PWRs) designed

by Combustion Engineering (CE). PVNGS

intends to qualify CENTS for use in future

Palo Verde licensing analyses by following

the guidelines prescribed in Generic Letter

(GL) 83–11, Supplement 1.

CENTS is a best-estimate code designed to

provide realistic simulation of Nuclear Steam

Supply System (NSSS) behavior during

normal and transient conditions. The CENTS

Safety Evaluation (SE) documents the generic

NRC approval of the CENTS code for use in

the licensing analyses for PWRs designed by

CE. The CENTS SE is described in letter,

‘‘Acceptance for Referencing of Licensing

Topical Report CE–NPD 282–P, ‘‘Technical

Manual for the CENTS Code’’ dated March

17, 1994, from USNRC to S. A. Toelle, ABB

Combustion Engineering.

The proposed change does not

immediately alter any methodology used in

[an] reload analysis. It only provides the

option to replace the CESEC transient

simulation code with an alternate NRC

approved code. Providing the option to

substitute the NRC approved CESEC code

with another NRC approved code (CENTS)

will not alter the physical characteristics of

any component involved in the initiation or

mitigation of an accident. The actual

implementation of the CENTS code will be

performed by following the guidance

provided in Generic Letter (GL) 83–11,

Supplement 1. This proposed change does

not result in any hardware changes or affect

plant operating practices. Thus, this

proposed change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

TS 5.6.5, core operating limits report

(COLR) which identifies the methodology

report(s) by number, title, date, and NRC staff

approval document, will be revised to allow

the reports to be identified by number and

title only. A note will be added to TS 5.6.5.b

to specify that a complete citation be

included in the COLR for each report,

including the report number, title, revision,

date, and any supplements.

This change has previously been reviewed

and accepted by the NRC in letter,

‘‘Acceptance for Siemens References to

Approved Topical Reports in Technical

Specifications’’ from S.A. Richards, NRC to

J.F. Mallay, Siemens Power Corporation

dated December 15, 1999. This change is also

consistent with NRC accepted TSTF 363-

revision 0.

Additionally, TS 5.6.5.b.6 and 5.6.5.b.7

both list the same topical report (Calculative

Methods for the CE Small Break LOCA

Evaluation Model, CENPD–137). TS 5.6.5.b.7

is the supplement to the topical report listed

in [TS] 5.6.5.b.6. TS 5.6.5.b.7 will be deleted

and the ‘‘Calculative Methods for the CE

Small Break LOCA Evaluation Model,

CENPD–137’’ topical report (along with its

supplement) will be listed in full text within

the COLR.

[The] proposed change[s related to the

listing of topical reports in TS 5.6.5.b are]

administrative change[s] and will have no

22024 Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices

affect on plant design, operation, or

maintenance. Thus, [these] proposed

change[s do] not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

TS 5.5.13, Diesel Generator Fuel Oil

Program. The proposed change is an

administrative change. This change would

have no affect on the physical plant.

Consequently, plant configuration and the

operational characteristics remain unchanged

and the proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

TS 5.5.14, TS Bases Control Program. The

proposed changes associated with TS

5.5.14.b do not involve any physical changes.

These changes allow PVNGS to be in

compliance with NRC approved changes to

10 CFR 50.59. This change is an

administrative change. Plant configuration

and the operational characteristics remain

unchanged and thus, the proposed changes

do not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

TS 5.5.15, SFDP. The proposed change to

TS 5.5.15 does not involve any physical

changes to the plant[s]. This change is an

administrative change. The loss of function

of the specific component is addressed in its

specific TS LCO and plant configuration will

be governed by the required actions of those

LCOs. Since this proposed change is a

clarification that does not degrade the

availability or capability of safety related

equipment, this change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

TS 5.6.5, COLR is being revised to add the

option to use the CENTS computer code in

licensing analysis by adding CENTS to the

list of approved core operating limit

analytical methods contained in TS 5.6.5.b.

The proposed change will not affect reload

analysis other than providing an option to

replace the CESEC transient simulation code

with an equivalent code. Providing this

option in and of itself will not alter the

physical characteristics of any component in

the plant. Since providing the option to use

the CENTS code will not alter the physical

characteristics of any component in the

plant, this proposed change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

TS 5.6.5, Core Operating Limits Report

(COLR) which identifies the methodology

report(s) by number, title, date, and NRC staff

approval document, will be revised to allow

the reports to be identified by number and

title only. This is an administrative change.

This change has no affect on the physical

plant. Plant configuration and the operational

characteristics remain unchanged and thus,

this 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.

TS 5.5.13, Diesel Generator Fuel Oil

Program. The proposed change to TS

5.5.13.a.3 is an administrative change. This

change would have no affect on the physical

plant and has no effect on any safety analyses

assumptions. Therefore, this proposed

change does not involve a significant

reduction in a margin of safety.

TS 5.5.14, TS Bases Control Program. The

proposed changes associated with TS

5.5.14.b will not reduce a margin of safety

because it has no direct effect on any safety

analyses assumptions. Changes to the TS

Bases that result in meeting the criteria in

paragraph (c)(2) of 10 CFR 50.59 will still

require NRC approval pursuant to 10 CFR

50.59. This change is administrative in

nature and is based on NRC reviewed and

approved changes to 10 CFR 50.59.

Therefore, this proposed change does not

involve a significant reduction in a margin of

safety.

TS 5.5.15, SFDP. The proposed change to

TS 5.5.15 are clarifications only. No changes

are made in the LCO, the time required for

the TS required actions to be completed, or

the out of service time for the components

involved. The NRC has approved the

proposed administrative changes (TSTF 273-

revision 2, as amended by editorial change

WOG–ED–23). Safety-related equipment

controlled by the TS will still perform as

credited in the safety analysis. Therefore, this

proposed change does not involve a

significant reduction in a margin of safety.

TS 5.6.5, COLR is being revised to add the

option to use the CENTS computer code in

licensing analysis by adding CENTS to the

list of approved core operating limit

analytical methods. The proposed change

will allow running existing analyses with a

different method that has been reviewed and

approved by NRC. The actual

implementation of the CENTS code will be

performed by following the guidance

provided in Generic Letter (GL) 83–11,

Supplement 1. Thus, this proposed change

does not involve a significant reduction in a

margin of safety.

TS 5.6.5, Core Operating Limits Report

(COLR) which identifies the methodology

report(s) by number, title, date, and NRC staff

approval document, will be revised to allow

the reports to be identified by number and

title only. This is an administrative change.

This change has no affect on the physical

plant. Plant configuration and the operational

characteristics remain unchanged. Therefore,

change does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on that

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the request

for amendments involves no significant

hazards consideration.

Attorney for licensee: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072–3999.

NRC Section Chief: Stephen Dembek.

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of amendments request: April 4,

2001 (102–04554).

Description of amendments request:

The amendments would revise

Specification 3.3.12, ‘‘Boron Dilution

Alarm System (BDAS),’’ and

Specification 3.9.2, ‘‘Refueling

Operations—Nuclear Instrumentation’’

of the technical specification (TSs).

Specification 3.9.2 applies to the

required operability of startup range

monitors (SRMs). The applicability

modes for limiting condition for

operation (LCO) 3.3.12 would be

extended to Mode 6, refueling. A note

to ‘‘Enter applicable Conditions and

Required Actions of LCO 3.3.12, ‘‘Boron

Dilution Alarm System (BDAS),’’ for

BDAS made inoperable by SRMs’’

would be added to the Actions for LCO

3.9.2 and the Required Action B.2 on

performing surveillance requirement

(SR) 3.9.1.1, and associated completion

time, for LCO 3.9.2 would be deleted.

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. The proposed changes to Technical

Specifications 3.3.12 and 3.9.2 do not

significantly increase the probability or

consequences of an accident previously

evaluated in the Updated Final Safety

Analysis Report (UFSAR) [for Palo Verde

Nuclear Generating Station]. The proposed

amendment[s] would add MODE 6

Applicability to TS 3.3.12 for the BDAS. In

addition, the proposed amendment[s] would

add a note to the Actions of TS 3.9.2 which

directs the operator to enter the applicable

Conditions and Required Actions of TS

3.3.12 in the event that the BDAS is made

inoperable by inoperable startup range

monitors (SRMs). Finally, the proposed

amendment[s] would delete the TS 3.9.2

Required Action B.2.

The boron dilution alarm system (BDAS)

and chemical monitoring of the reactor

coolant system (RCS) boron concentration are

established in the MODE 6 inadvertent

deboration analysis in UFSAR Section 15.4.6

to alert the operator of a boron dilution event

at least 30 minutes prior to a loss of

subcriticality. The BDAS and RCS boron

monitoring are not accident initiators. The

proposed changes will ensure that the

assumptions of UFSAR Section 15.4.6, for

mitigating an inadvertent deboration event,

are met. In addition, the proposed changes do

Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices 22025

not alter the design or configuration of the

plant but establish requirements for operating

the plant as analyzed and designed. The

amendment[s do] not physically affect the

operability or availability of the boron

dilution alarm system (BDAS), but ensures it

is available as required or that sufficient

actions are taken if it becomes inoperable.

Furthermore, the inadvertent deboration

event analysis does not involve dose

consequences since the acceptance criteria is

to provide operator notification at least 30

minutes prior to the loss of subcriticality

such that the operator may terminate the

event before subcriticality is achieved [and

exceeded,] and the RCS and fuel clad

boundaries are challenged. Therefore, the

proposed amendment[s] to TS 3.3.12 and TS

3.9.2 [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 of

accident from any accident previously

evaluated?

No. The proposed amendment[s] to

Technical Specifications 3.3.12 and 3.9.2 [do]

not create the possibility of an accident of a

new or different kind from any accident

previously evaluated. The proposed

amendment[s] would add MODE 6

Applicability to TS 3.3.12 for the BDAS. In

addition, the proposed amendment[s] would

add a note to the Actions of TS 3.9.2 which

directs the operator to enter the applicable

Conditions and Required Actions of TS

3.3.12 in the event that the BDAS is made

inoperable by inoperable startup range

monitors (SRMs). Finally, the proposed

amendment[s] would delete the TS 3.9.2

Required Action B.2. The proposed changes

do not alter the design or configuration of the

plant but establish requirements for operating

the plant as analyzed and designed.

In MODE 6, the BDAS and the startup

range monitors (SRM) are the primary means

to monitor reactivity changes during core

alterations and to alert the operator of a

boron dilution event in time to prevent a loss

of subcriticality. Chemical sampling to

monitor RCS boron concentration is used

when the BDAS is unavailable. Accidents

involving reactivity anomalies are evaluated

in UFSAR Section 15.4, Reactivity and Power

Distribution Anomalies. Inadvertent

deboration is described in UFSAR Section

15.4.6 as requiring the BDAS or chemical

monitoring of the RCS boron concentration to

alert the operator at least 30 minutes prior to

the loss of subcriticality in MODE 6. The

proposed changes to TS 3.3.12 and 3.9.2 will

require the BDAS to be OPERABLE in MODE

6 or perform RCS boron concentration

monitoring if the BDAS is inoperable.

The BDAS and RCS boron concentration

monitoring are means to detect a boron

dilution event. The proposed changes ensure

this detection occurs as required. The

proposed amendment[s do] not physically

affect the response or operation of the plant.

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?

No. The proposed changes to Technical

Specifications 3.3.12 and 3.9.2 do not involve

a significant reduction in a margin of safety.

The proposed amendment[s] would add

MODE 6 Applicability to TS 3.3.12 for the

BDAS. In addition, the proposed

amendment[s] would add a note to the

Actions of TS 3.9.2 which directs the

operator to enter the applicable Conditions

and Required Actions of TS 3.3.12 in the

event that the BDAS is made inoperable by

inoperable startup range monitors (SRMs).

Finally, the proposed amendment[s] would

delete the TS 3.9.2 Required Action B.2.

These changes ensure the adequate detection

of a boron dilution event.

The current Technical Specifications

3.3.12 satisfies the inadvertent deboration

safety analysis requirements to have the

BDAS OPERABLE in MODES 3, 4,and 5. In

accordance with UFSAR Section 15.4.6,

Inadvertent Deboration, the same

requirements and actions apply for MODE 6.

Therefore, it is proposed that MODE 6

Applicability for the BDAS be added to TS

3.3.12. In addition, the Action section of TS

3.9.2 would be modified with a note to

ensure the safety analysis assumptions are

satisfied in MODE 6, since the SRM must be

OPERABLE for the corresponding BDAS

channel to be OPERABLE. Technical

Specification Bases 3.3.12 and UFSAR

Section 15.4.6 indicate that the BDAS is

necessary to alert the operator of an

inadvertent deboration event at least 15

minutes before the reactor loses subcriticality

in MODES 3, 4, and 5. UFSAR Section 15.4.6

also indicates that 30 minutes is required in

MODE 6. These criteria are in agreement with

the guidance of NUREG 0800, [NRC’s]

Standard Review Plan. Therefore, the margin

of safety being considered for [these]

proposed amendment[s] is the 30 minutes

before the loss of subcriticality that the

operator must be notified [...] in the event of

a boron dilution event. The proposed

changes to TS 3.3.12 and TS 3.9.2 will

require the BDAS to be OPERABLE in MODE

6 and, if the BDAS is inoperable, will require

that the RSC boron concentration be

monitored at pre-analyzed frequencies via

chemical sampling in order to satisfy the 30

minute acceptance criteria. Finally, the

proposed change[s] also serve to clarify that

an inoperable SRM will cause the

corresponding BDAS channel to be

inoperable, thus requiring action in

accordance with TS 3.3.12, in addition to TS

3.9.2. [The proposed changes add a

requirement to TS 3.3.12 and account for the

BDAS being inoperable because of inoperable

SRMS] Therefore, the proposed change[s do]

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on that

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the request

for amendments involves no significant

hazards consideration.

Attorney for licensee: Nancy C. Loftin,

Esq., Corporate Secretary and Counsel,

Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix,

Arizona 85072–3999.

NRC Section Chief: Stephen Dembek.

Consumers Energy Company, Docket

No. 50–155, Big Rock Point Plant,

Charlevoix, County, Michigan

Date of amendment request: October

26, 2000, as supplemented by letters

dated February 9, February 28, March

14, March 15, and March 23, 2001.

Description of amendment request:

The proposed amendment reflects the

replacement of the original 75-ton

reactor building gantry crane by an

upgraded single-failure proof 125-ton

crane designed to meet Crane

Manufacturers Association of America

(CMAA) Specification 70 and American

Society of Mechanical Engineers

(ASME) B30.2. The proposed

amendment to the Technical

Specifications (TSs) would revise (1)

Definition 1.8 on fuel handling, (2) the

applicability of TS 3/4.2.1 on fuel

handling support system requirements,

and (3) Section 3.2.2.d of the limiting

conditions for operation for TS 3/4.2.2

on fuel handling general requirements,

and would delete TS 3/4.3.1 on control

of heavy loads. The licensee also

submitted revisions to the bases for TSs

3/4.2.2 and 3/4.3.1. The crane has a

Design Rated Load (DRL) of 125 tons;

however, it has been analyzed to safely

retain a load of 105 tons under the sitespecific earthquake and the Maximum

Critical Load (MCL) for the crane is 105

tons.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee provided its analysis in its

letters dated October 26, 2000, and

March 14, 2001, which address the issue

of no significant hazards consideration,

and is presented below:

The proposed [amendment] does not:

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

A significant increase in the probability of

an accident is not created because:

• The replacement crane will not be

utilized for a greater number of fuel handling

evolutions than was the case for the existing

75-ton crane. The existing crane was utilized

for each transfer of fuel assemblies between

the reactor and the Spent Fuel Pool; in the

case of full-core offloads, which was the

normal practice during refueling outages at

Big Rock Point [Plant], the existing crane

would make 84 transfers of irradiated fuel

from the reactor to the Spent Fuel Pool, and

a nominal 62 transfers of irradiated fuel from

the Spent Fuel Pool back to the reactor. The

replacement crane will handle fuel only after

it has been placed into the W100 Transfer

Cask. It is anticipated that the W100 Fuel

Transfer Cask will be handled 14 times while

it contains fuel (one movement from the

22026 Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices

Spent Fuel Pool to a staging area in Room

444, and one movement from Room 444 to

a W150 Storage Cask), during loading of

seven W150 Storage Casks. Additional moves

of the W100 Transfer Cask when it is loaded

with fuel would be required only if an offnormal condition required a loaded cask to

be returned to the Spent Fuel Pool.

• The replacement crane has been

analyzed to safely handle the 105-ton W100

Fuel Transfer Cask under seismic conditions

that include the Big Rock Point [Plant] sitespecific safe shutdown earthquake of 0.104g.

The UFHSR [Updated Final Hazards

Summary Report] is being revised to limit the

weight of loads being moved over the Spent

Fuel Pool to 105 tons.

• The existing crane has been used to lift

the properly rigged 24-ton fuel transfer cask

over fuel; the probability of dropping the 24-

ton fuel transfer cask was minimized by the

proper rigging that consisted of attaching a

safety catch device to the transfer cask. In the

case of the replacement crane, loads will be

prevented from dropping by the design of the

single-failure proof Ederer X–SAM hoist,

which prevents loads from dropping more

than 18 inches in the event of any single

failure. Administrative controls will be

instituted on the use of the replacement

crane to require lifts of any heavy loads over

fuel or over structures, the failure of which

would jeopardize safe storage of fuel, to be

done at a height of greater than 18 inches.

Administrative controls will be instituted to

prohibit use of the replacement crane for

movement of any cask over fuel; these

controls will be specified in the Big Rock

Point [Plant] UFHSR. Administrative controls

that apply to our [the licensee’s] current 75-

ton crane will be maintained, and

strengthened, as appropriate, to provide

greater assurance that heavy loads

transported over fuel will be safely

transported. Strengthened administrative

controls include limiting the number of crane

operators to approximately 12 individuals,

and requiring that they receive Operator

Engineer training in the use of the upgraded

crane.

• The existing crane met single-failure

proof criteria only when it was used to

handle the properly-rigged 24-ton fuel

transfer cask; the 105-ton single-failure proof

crane will be single-failure proof for all lifts

of loads which are 105 tons or less.

• The replacement of the existing crane

with a 105-ton single-failure proof crane is

being performed as a safety-related

modification, and 10 CFR [Part] 50 Appendix

B [Quality Assurance] criteria are being

applied to all critical elements of design,

purchasing, installation and testing.

Therefore, the replacement crane and trolley

can be expected to perform in accordance

with their design specifications. As a result,

the probability of a trolley failure on the

replacement crane is considered to be no

greater than the probability of a failure of the

safety catch device which was employed

with the existing crane when it was used to

handle the 24-ton fuel transfer cask.

A significant increase in the consequences

of an accident is not created because:

• This change affects fuel handling, and

fuel handling accidents have already been

analyzed and bound all other categories of

accidents at the Big Rock Point Plant.

Analysis indicates that the dose from the

bounding fuel accident (a 24-ton fuel transfer

cask drop), assuming a free release path

without isolation of ventilation from

containment, falls below the Protective

Action Guidelines (PAGs) of Environmental

Protection Agency–400 (EPA–400) 68 days

following plant shutdown (the reactor was

shutdown [as] of 8/29/1997). The analysis

assumed a total of 500 damaged assemblies

in the Spent Fuel Pool, with 84 of them being

freshly discharged from the reactor. The

Spent Fuel Pool contains 441 fuel assemblies.

With more than three years of radiological

and heat decay since the plant was

shutdown, the potential source terms for

gaseous and volatile radionuclides associated

with the remaining design basis accidents

has continued to decrease; therefore, the

doses at the site boundary associated with a

postulated accident involving any number of

the available fuel assemblies have also

decreased. The design of the Ederer X–SAM

trolley and hoist is such that upon a single

failure of the trolley that would allow the

suspended load to free-fall, the load could

fall for a maximum of 18 inches before the

drum brake mechanism would engage to stop

the downward travel. An 18-inch drop of the

105-ton dry fuel storage system fuel transfer

cask has been analyzed and has been

determined not to result in failure of the

floors of the Spent Fuel Pool, Room 444 or

the laydown area at the 599-foot 5-inch

elevation of containment. These are the only

floors in containment over which the cask

will [be] moved with the 105-ton singlefailure proof crane at a height of less than 18

inches. The 105-ton W100 Transfer Cask is

the largest load that will be handled over the

Spent Fuel Pool, when fuel is being stored in

the Pool. For other floors/structures, (i.e., the

Reactor Deck at elevation 632′ 6′′ [632 feet 6

inches]), administrative controls will be

imposed to require the 105-ton cask to be

suspended at least 18 inches above the floor/

structure.

[The proposed amendment reflects the

replacement of the original 75-ton non singlefailure proof crane, with a single-failure proof

crane. The replacement crane addresses

malfunctions (e.g., dropping loads under

single-failure conditions) that were possible

with the original crane.]

Based on this discussion, it is concluded

that this proposed change to the Defueled

Technical Specifications does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change is requested to reflect

the removal of the original 75-ton reactor

building non single-failure proof semi-gantry

crane and its replacement with a singlefailure proof 105-ton crane, which will be

designed to meet the applicable criteria and

guidelines of NUREG–0554 and NUREG–

0612. The change results from installation of

a crane that replaces another crane. The

general functions performed by the

replacement crane (cask handling and

movement of heavy loads) do not differ from

those performed by the original crane.

Therefore, new or different accidents will not

be created by elimination of restrictions

associated with the original 75-ton crane,

since the design of the replacement crane

addresses malfunctions (for example,

dropping loads under single-failure

conditions) that were possible with the

original crane.

Based on this discussion, it is concluded

that this proposed change to the Defueled

Technical Specifications does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Involve a significant reduction in the

margin of safety.

To prevent failure of the Spent Fuel Pool

structure when handling loads over the Pool

with the existing 75-ton crane, loads were

limited to 24 tons, and cask handling

evolutions were limited to the southwest

corner of the Spent Fuel Pool. These

measures ensured that the Spent Fuel pool

would not fail as a result of a load being

dropped into it. The replacement crane has

been designed such that a load will not drop

more than 18 inches if a single failure should

occur in its trolley. A drop of the 105-ton

W100 Fuel Transfer Cask from a height of 18

inches to the Spent Fuel Pool floor has been

determined not to result in failure of the

Spent Fuel Pool; loads handled by the

replacement crane will be restricted to 105

tons to ensure that the structural integrity of

the Spent Fuel Pool will not be compromised

by a postulated drop of the 105-ton W100

Fuel Transfer Cask.

The existing crane is designed to handle

loads up to 75 tons Because the existing

crane was not designed as a single-failure

proof crane, restrictions were placed on load

paths, load weights, and the configuration of

the 24-ton fuel transfer cask (the cask was

required to have a safety catch device

attached between the cask and the crane

structure to prevent dropping the transfer

cask in the event of a trolley failure) to

ensure that a margin of safety existed with

respect to dropping heavy loads on spent fuel

and to prevent a dropped load from causing

structural failure of the Spent Fuel Pool. The

replacement crane is designed to withstand

the Big Rock Point [Plant] site-specific safe

shutdown earthquake of 0.104g while safely

retaining a load equal to 105 tons. Therefore,

handling the 105-ton W100 transfer cask with

this crane provides equivalent margins with

respect to crane failure as the current

restriction that limits loads being handled

over the Spent Fuel Pool to 24 tons. The

UFHSR will restrict handling of loads over

the Spent Fuel Pool to 105 tons whenever

fuel is stored in the Pool. The trolley and

hoist for the replacement crane are designed

to be single-failure proof, and provide a

margin of safety for dropping a suspended

load equivalent to the safety catch employed

with 24-ton transfer cask safety catch.

[The proposed amendment reflects the

replacement of the original 75-ton non singlefailure proof crane, with a single-failure proof

125-ton crane having an MCL of 105 tons..

The replacement crane is designed to meet

the applicable criteria of NUREG–0554 and

Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices 22027

NUREG–0612, CMAA Specification 70, and

ASME B30.2.]

Based on this discussion, it is concluded

that this proposed change to the Defueled

Technical Specifications does not involve a

significant decrease in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis in both letters of

October 26, 2000, and March 14, 2001,

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: David A.

Mikelonis, Esquire, Consumers Energy

Company, 212 West Michigan Avenue,

Jackson, Michigan 49201.

NRC Section Chief: Robert A. Gramm.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: March 5,

2001, as revised March 30, 2001.

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 5.5.12,

‘‘Technical Specifications (TS) Bases

Control Program,’’ to be consistent with

the changes to 10 CFR 50.59 published

in the Federal Register on October 4,

1999 (64 FR 53582), as reflected in the

Nuclear Energy Institute’s Technical

Specification Task Force (TSTF)

Standard TS Change Traveler, TSTF–

364, ‘‘Revision to TS Bases Control

Program to Incorporate Changes to 10

CFR 50.59.’’ Specifically, Palisades TS

5.5.12b currently states, in part, that

licensees may make changes to Bases

without prior NRC approval provided

the changes do not ‘‘involve * * * [a]

change to the updated FSAR [Final

Safety Analysis Report] or Bases that

involves an unreviewed safety question

as defined in 10 CFR 50.59.’’ The

proposed amendment would change

this quoted portion of TS 5.5.12b to

state ‘‘require * * * [a] change to the

updated FSAR or Bases that requires

NRC approval pursuant to 10 CFR

50.59.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The following evaluation supports the

finding that operation of the facility in

accordance with the proposed changes would

not:

a. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change deletes the reference

to unreviewed safety question as defined in

10 CFR 50.59. Deletion of the definition of

unreviewed safety question was approved by

the NRC with the revision of 10 CFR 50.59.

Consequently, the probability of an accident

previously evaluated is not significantly

increased. Changes to the TS Bases are still

evaluated in accordance with 10 CFR 50.59.

As a result, the consequences of any accident

previously evaluated are not significantly

affected. Therefore, this change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

b. Create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed change does not involve a

physical alteration of the plant (no new or

different type of equipment will be installed)

or a change in the methods governing normal

plant operation. Therefore, this change does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

c. Involve a significant reduction in the

margin of safety.

The proposed change will not reduce a

margin of safety because it has no direct

effect on any safety analyses assumptions.

Changes to the TS Bases that result in

meeting the criteria in paragraph 10 CFR

50.59(c)(2) will still require NRC approval

pursuant to 10 CFR 50.59. This change is

administrative in nature based on the

revision to 10 CFR 50.59. 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: Arunas T.

Udrys, Esquire, Consumers Energy

Company, 212 West Michigan Avenue,

Jackson, Michigan 49201.

NRC Section Chief: Claudia M. Craig.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: April 2,

2001.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) by

removing all requirements for, and

references to, the ‘‘Assembly Radial

Peaking Factor,’’ (FRA). Consequently, in

TS Section 1.0, the definition of

Assembly Radial Peaking Factor would

be deleted and the definition of the

Total Radial Peaking Factor (FRT) would

be corrected to read: ‘‘FRT shall be the

maximum ratio of the individual fuel

pin power to the core average pin power

integrated over the total core height,

including tilt.’’ In Limiting Condition

for Operation (LCO) 3.2.2, the title

would be changed to ‘‘TOTAL RADIAL

PEAKING FACTOR (FRT);’’ the wording

would state ‘‘FRT shall be within the

limits specified in the [Core Operating

Limits Report] COLR;’’ Condition A

would state ‘‘FRT not within limits

specified in the COLR;’’ Required

Action A.1 would state ‘‘Restore FRT to

within limits;’’ and Surveillance

Requirement (SR) 3.2.2.1 would state

‘‘Verify FRT is within limits specified in

the COLR.’’ In LCO 3.2.3, Required

Action A.1 would state: Verify FRT is

within the limits of LCO 3.2.2, ‘‘Total

Radial Peaking Factor (FRT)’.’’

Associated changes would be made to

the TS Bases and table of contents.

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. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

There are no changes in plant systems,

plant control operating procedures or

instrument alarm or trip settings associated

with this [TS Change Request] TSCR.

Because neither physical equipment, nor

operating methods for that equipment

change, the probability of accident initiation

would not change. Therefore, the proposed

technical specification change would not

involve a significant increase in the

probability of an accident previously

evaluated.

The assembly radial peaking (FRA) has been

used in the past safety analyses and

radiological consequence analyses. These

analyses utilized the assumption that FRA

would remain within the Technical

Specifications limit during plant operations.

These analyses verify, for Anticipated

Operational Occurrences (AOOs) and

Postulated Accidents (PAs), that:

(1) The Departure from Nucleate Boiling

Ratio (DNBR) remains above the appropriate

Technical Specifications Safety Limit, and

(2) The calculated offsite doses and control

room dose for the affected events remained

within the guidelines of 10 CFR 100, Section

11, ‘‘Determination of exclusion area, low

population zone and population center

distance,’’ and 10 CFR 50, Appendix A,

General Design Criteria (GDC) 19, ‘‘Control

room.’’

Improved DNB correlations and better

spacer grid design have allowed the safety

analysis calculations to be performed using

only the total radial peaking factor (FRT) limit

(which remains unchanged), without

exceeding the specified Safety Limits. The

radiological consequence events that

previously used the FRA limit have been reanalyzed using the slightly higher FRT limit

to determine the source strength. The revised

calculated offsite dose and control room dose

for the affected events remained within the

guidelines of 10 CFR 100 and GDC 19.

Because the results of the transient

analyses, which were performed without FRA

22028 Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices

assumptions, continue to meet the Safety

Limits, and because the dose consequences of

all analyzed events, which were also

performed without FRA assumptions,

continue to be within the guidelines of 10

CFR 100 and GDC 19, the proposed technical

specification change would not involve a

significant increase in the consequences of an

accident previously evaluated.

Therefore, operation of the plant in

accordance with the proposed Technical

Specifications would not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

B. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

Operation of the plant in accordance with

the proposed Technical Specifications would

not add any new equipment, settings, or alter

any plant operating practices. The only

change is the deletion of all Technical

Specifications references to the Assembly

Radial Peaking Factor, FRA, (a peaking factor

no longer used in core design or safety

analyses). Since there will be no change in

operating plant equipment, settings, or

normal operating practices, operation in

accordance with the proposed Technical

Specifications would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

C. Does this change involve a significant

reduction in a margin of safety?

The disposition of the [Standard Review

Plan] SRP Chapter 15 events, the setpoint

verification, the [fuel centerline melt] FCM

and the [minimum departure from nucleate

boiling ratio] MDNBR analyses documented

in Siemens report EMF–2259 Revision 1,

‘‘Palisades Cycle 15 Safety Analysis Report’’

dated August 1999 considered the impact of

several changes in fuel design and plant

operations for Cycle 15. A detailed and

simplified XCOBRA–IIIC model that

incorporated limiting radial and axial power

distributions, as well as the removal of the

FRA peaking limit, were developed for Cycle

15. This model was applied to all DNB event

analyses for Cycle 15 and the MDNBR values

for limiting AOOs and PAs were evaluated

with the [High Thermal Performance] HTP

DNB correlation. The limiting MDNBR is

calculated for SRP event 15.3.3 Reactor

Coolant Pump Rotor Seizure and the limiting

FCM is calculated for SRP event 15.4.3 Single

Rod Withdrawal. The calculated results for

the limiting events meet the Safety Limits

specified in TS LCO 2.1.

The SRP events were dispositioned in

accordance with Siemens approved

methodologies listed in Palisades TS Section

5.6.5, Amendment 189. The completed safety

analysis supports Palisades plant operation at

2530 Mwt.

The results of the transient analyses, which

were performed without FRA assumptions,

continue to meet the Safety Limits, and the

dose consequence of all analyzed events,

which were also performed without FRA

assumptions, continue to be within the

guidelines of 10 CFR 100 and GDC 19 * * *

[Therefore] operation of the Facility in

accordance with the proposed technical

specification change would not involve a

significant reduction in the margin of safety.

Therefore, operation of the plant in

accordance with the proposed Technical

Specifications would 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: Arunas T.

Udrys, Esquire, Consumers Energy

Company, 212 West Michigan Avenue,

Jackson, Michigan 49201.

NRC Section Chief: Claudia M. Craig.

Duke Energy Corporation, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of amendment request: March 1,

2001.

Description of amendment request:

The amendments would allow

implementation of 10 CFR Part 50,

Appendix J, Option B, which governs

performance-based containment leakage

testing requirements for Types B and C

testing. Catawba has previously

implemented 10 CFR Part 50, Appendix

J, Option B requirements for Type A

testing. In addition to the changes

associated with the adoption of 10 CFR

Part 50, Appendix J, Option B, the

licensee is also proposing the following

two changes: (1) Technical Specification

(TS) 3.6.3 will be modified to delete the

requirement for conducting soap bubble

tests of welded penetrations during

Type A tests which are not individually

Type B or Type C testable, and (2) the

Bases for TS 3.6.2 will be modified to

clarify that for the purpose of certain TS

3.6.2 Required Actions, the air lock door

bulkhead is considered to be part of the

door.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

The following discussion is a summary of

the evaluation of the changes contained in

this proposed amendment against the 10 CFR

50.92(c) requirements to demonstrate that all

three standards are satisfied. A no significant

hazards consideration is indicated if

operation of the facility in accordance with

the proposed amendment would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated, 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.

First Standard

The proposed amendment will not involve

a significant increase in the probability or

consequences of an accident previously

evaluated. Implementation of these changes

will provide continued assurance that

specified parameters associated with

containment integrity will remain within

acceptance limits as delineated in 10 CFR

[Part] 50, Appendix J, Option B. The changes

are consistent with current safety analyses.

Although some of the proposed changes

represent minor relaxation to existing TS

requirements, they are consistent with the

requirements specified by Option B of 10

CFR [Part] 50, Appendix J. The systems

affecting containment integrity related to this

proposed amendment request are not

assumed in any safety analyses to initiate any

accident sequence. Therefore, the probability

of any accident previously evaluated is not

increased by this proposed amendment. The

proposed changes maintain an equivalent

level of reliability and availability for all

affected systems. In addition, maintaining

leakage within analyzed limits assumed in

accident analyses does not adversely affect

either onsite or offsite dose consequences.

Therefore, the proposed amendment does not

increase the consequences of any accident

previously evaluated.

Second Standard

The proposed amendment will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. No changes are being proposed

which will introduce any physical changes to

the existing plant design. The proposed

changes are consistent with the current safety

analyses. Some of the changes may involve

revision in the testing of components;

however, these are in accordance with the

Catawba current safety analyses and provide

for appropriate testing or surveillance that is

consistent with 10 CFR [Part] 50, Appendix

J, Option B. The proposed changes will not

introduce new failure mechanisms beyond

those already considered in the current safety

analyses. No new modes of operation are

introduced by the proposed changes. The

proposed changes maintain, at minimum, the

present level of operability of any system that

affects containment integrity.

Third Standard

The proposed amendment will not involve

a significant reduction in a margin of safety.

The provisions specified in Option B of 10

CFR [Part] 50, Appendix J allow changes to

Type B and Type C test intervals based upon

the performance of past leak rate tests. 10

CFR [Part] 50, Appendix J, Option B allows

longer intervals between leakage tests based

on performance trends, but does not relax the

leakage acceptance criteria. Changing test

intervals from those currently provided in

the TS to those provided in 10 CFR [Part] 50,

Appendix J, Option B does not increase any

risks above and beyond those that the NRC

has deemed acceptable for the performance

Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices 22029

based option. In addition, there are risk

reduction benefits associated with reduction

in component cycling, stress, and wear

associated with increased test intervals. The

proposed changes provide continued

assurance of leakage integrity of containment

without adversely affecting the public health

and safety and will not significantly reduce

existing safety margins. Similar proposed

changes have been previously reviewed and

approved by the NRC, and they are

applicable to Catawba.

Based upon the preceding discussion,

Duke Energy has concluded that the

proposed amendment does not involve 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

amendments request involves no

significant hazards consideration.

Attorney for licensee: Ms. Lisa F.

Vaughn, Legal Department (PB05E),

Duke Energy Corporation, 422 South

Church Street, Charlotte, North Carolina

28201–1006.

NRC Section Chief: Richard L. Emch,

Jr..

Duke Energy Corporation, Docket Nos.

50–269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of amendment request: March

29, 2001.

Description of amendment request:

The proposed amendments would

revise the Keowee Hydro Unit (KHU)

Technical Specifications Surveillance

Requirements (SRs) to address concerns

related to voltage and frequency

overshoot during surveillance testing.

This would be accomplished by

removing the note that had been

implemented by Amendment Nos. 316,

316, and 316 (October 4, 2000) for

Oconee Nuclear Station, Units 1, 2, and

3, respectively, to temporarily waive the

upper limits specified in Surveillance

Requirement 3.8.1.9, thereby

reinstalling the original SR. In addition,

as a result of an upgrade of the KHU

governors, the proposed amendments

would reduce the time delay specified

in Technical Specification 3.8.1 and SR

3.8.1.17 from 12 seconds ±1 second to

5 seconds ±1 second. In addition,

related Bases changes have been

proposed.

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.

No. The License Amendment Request

(LAR) removes a Note to Surveillance

Requirement (SR) 3.8.1.9 that temporarily

waived the surveillance requirements

associated with the upper limits for Keowee

Hydro Unit (KHU) voltage and frequency.

The waiver of these requirements allowed

Duke to avoid an unplanned forced

shutdown of all three Oconee units, and the

potential safety consequences and

operational risks associated with that action.

This LAR also changes the arming time

delay associated with the out-of-tolerance

logic that had been approved for installation

in Amendment Nos. 312, 312, and 312. This

change lowers the allowed time delay,

thereby resulting in the activation of the outof-tolerance logic more quickly after KHU

startup.

Since this LAR assures that each KHU

reaches its required operating band within

the required time, and that if maloperation of

a unit occurs, the KHU will be taken off line,

the probability or consequences of an

accident previously evaluated is not

significantly increased.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

No. The LAR involves removing a Note

that temporarily waived SR 3.8.1.9.a

associated with the KHUs. This LAR also

changes the time delay associated with the

activation of out-of-tolerance logic that had

been approved for installation in

Amendment Nos. 312, 312, and 312. This

change lowers the allowed time delay,

thereby resulting in the activation of the outof-tolerance logic more quickly after KHU

startup.

Since this LAR restores Technical

Specification SR 3.8.1.9 to the condition

prior to Amendment Nos. 316, 316, and 316

and provides a shortened arming delay for

the out-of-tolerance logic that was approved

in Amendment Nos. 312, 312, and 312, no

new failure mechanism or accident sequence

is introduced. Therefore, the possibility of a

new or different kind of accident from any

kind of accident previously evaluated is not

created.

3. Involve a significant reduction in a

margin of safety.

No. The LAR involves removing a Note

that allowed temporary waiver of the

requirements to meet SR 3.8.1.9.a and

shortens the arming time delay associated

with the activation of out-of-tolerance logic

that had been approved for installation in

Amendment Nos. 312, 312, and 312.

This LAR, therefore, improves the margin

of safety by assuring that SR 3.8.1.9.a can be

implemented. The change to a shorter arming

time delay for the out-of-tolerance circuit

activation also improves the margin of safety

by limiting the time that a KHU would be

carrying safety loads in an out-of-tolerance

condition.

Therefore, this request 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: Anne W.

Cottington, Winston and Strawn, 1200

17th Street, NW., Washington, DC

20005.

NRC Section Chief: Richard L. Emch,

Jr.

Entergy Operations Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: April 13,

2001.

Description of amendment request:

The proposed change relaxes the

allowable cooldown rate in the Reactor

Coolant System (RCS) Technical

Specifications (TS) 3.4.8.1, ‘‘Pressure /

Temperature Limits.’’ Specifically, the

change eliminates the limitation of a 10

°F per hour cooldown rate when the

RCS temperature is below 135 °F. The

proposed limitations permit a 100 °F per

hour cooldown rate to continue down to

an RCS temperature of 110 °F, at which

point the rate is reduced to 30 °F per

hour.

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:

Limitations have been imposed on

cooldown of the Reactor Coolant System

(RCS) to assure compliance with the

minimum temperature requirements of 10

CFR [Part] 50, Appendix G. The proposed

changes revise the allowable cooldown limits

in a way such that operation remains

consistent with the design assumptions and

satisfies the stress limits for cyclic operation.

By ensuring operation remains within the

bounds of the existing design basis and

assumptions, the probability of a brittle

fracture of the reactor vessel has not been

increased.

Therefore, the proposed change will not

involve a significant 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 changes will not create the

possibility of a new or different kind of

accident from any previously analyzed since

they do not introduce new systems, failure

22030 Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices

modes, or other plant perturbations. The

proposed changes revise the cooldown

limitations based on the fact the

conservatively estimated peak pressure that

can occur when the RCS cold leg temperature

is below 200 °F is less than the proposed

pressure limit. The limits assure that

operation remains consistent with the design

assumptions and satisfies the stress limits for

cyclic operation.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

3. Will the operation of the facility in

accordance with this proposed change

involve a significant reduction in a margin of

safety?

Response:

The margin of safety provided by

Technical Specification 3.4.8.1 is based on

assuring that the maximum cooldown rates

are consistent with the design assumptions

and satisfy the stress limits for cyclic

operation. The proposed changes will not

involve a significant reduction in the margin

of safety since equivalent pressure and

temperature limit requirements for reactor

operation will be applied. The proposed

changes were derived in accordance with

approved NRC methodology which was

developed to assure the reactor coolant

system pressure boundary is designed with

sufficient margin to withstand any condition

during normal operation including

anticipated operational occurrences and

system in-service leak and hydrostatic tests.

These requirements were revised in

accordance with 10 CFR [Part] 50, Appendix

G utilizing the latest NRC guidance in

Regulatory Guide 1.99, Revision 2 relative to

estimating neutron irradiation damage to the

reactor vessel. In addition, the 16 EFPY

[effective full power year] basis for these

pressure/temperature limits has been found

to include sufficient margin to account for

the limits of uncertainty described in Draft

Regulatory Guide DG–1053.

Therefore, the proposed change will not

involve in a significant reduction in a margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: N.S. Reynolds,

Esquire, Winston & Strawn, 1400 L

Street NW., Washington, DC 20005–

3502.

NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC,

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of amendment request: February

9, 2001.

Description of amendment request:

The proposed amendment (one-time

change) revises the Steam Generator

(SG) inspection frequency requirements

in TS 5.5.9.d.2, ‘‘Steam Generator (SG)

Tube Surveillance Program, Inspection

Frequencies,’’ for the Braidwood

Station, Unit 1, Fall 2001 refueling

outage, to allow a 40 month inspection

interval after one SG inspection, rather

than after two consecutive inspections

resulting in C–1 classification. This onetime change is proposed to eliminate

unnecessary SG inspections during the

upcoming Unit 1, Fall 2001 refueling

outage, thus, resulting in significant

dose, schedule, and cost savings.

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 one-time change revises the

Steam Generator (SG) inspection interval

requirements in Technical Specifications

(TS) 5.5.9.d.2, ‘‘Steam Generator (SG) Tube

Surveillance Program, Inspection

Frequencies,’’ for the Braidwood Station,

Unit 1, Fall 2001 refueling outage, to allow

a 40 month inspection frequency after one

inspection, rather than after two consecutive

inspections results that are within the C–1

category. C–1 category is defined as ‘‘<5% of

the total tubes inspected are degraded tubes

and none of the inspected tubes are

defective.’’

The proposed one-time extension of the

Unit 1, SG tube inservice inspection interval

does not involve changing any structure,

system, or component, or affect reactor

operations. It is not an initiator of an accident

and does not change any existing safety

analysis previously analyzed in the Byron/

Braidwood Stations’ Updated Final Safety

Analysis Report (UFSAR). As such, the

proposed changes do not involve a

significant increase in the probability of an

accident previously evaluated.

Since the proposed change does not alter

the plant design, there is no direct increase

in SG leakage. Industry experience indicates

that the probability of increased SG tube

degradation would not go undetected.

Additionally, steps described below will

further minimize the risk associated with this

extension. For example, the scope of

inspections performed during the last

Braidwood Station, Unit 1, refueling outage

(i.e., the first refueling outage following SG

replacement) exceeded the TS requirements

for the first two refueling outages after SG

replacement. That is, more tubes were

inspected than were required by TS.

Currently, Braidwood Station, Unit 1, does

not have an active SG damage mechanism,

and will meet the current industry

examination guidelines without performing

SG inspections during the next refueling

outage. Additionally, as part of our SG Tube

Surveillance Program, both a Condition

Monitoring Assessment and an Operational

Assessment are performed after each

inspection and compared to the Nuclear

Energy Institute (NEI) 97–06, ‘‘Steam

Generator Program Guidelines,’’ performance

criteria. The results of the Condition

Monitoring Assessment demonstrated that all

performance criteria were met during the

Braidwood Station, Unit 1, Spring 2000

refueling outage, and the results of the

Operational Assessment show that all

performance criteria will be met over the

proposed operating period. Considering these

actions, along with the improved SG design

and reliability of Babcock and Wilcox

International (BWI) replacement SGs,

extending the SG tube inspection frequency

does not involve a significant increase in the

consequences of an accident previously

evaluated.

Therefore, the proposed change does 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 of

accident from any accident previously

evaluated?

The proposed change revises the SG

inspection frequency requirements in TS

5.5.9.d.2, ‘‘Steam Generator (SG) Tube

Surveillance Program, Inspection

Frequencies,’’ for the Braidwood Station,

Unit 1, Fall 2001 refueling outage, to allow

a 40 month inspection interval after one

inspection, rather than after two consecutive

inspections with inspection results within

the C–1 category.

The proposed change will not alter any

plant design basis or postulated accident

resulting from potential SG tube degradation.

The scope of inspections performed during

the last Braidwood Station, Unit 1, refueling

outage (i.e., the first refueling outage

following SG replacement) significantly

exceeded the TS requirements for the scope

of the first two refueling outages after SG

replacement.

Primary to secondary leakage that may be

experienced during all plant conditions is

expected to remain within current accident

analysis assumptions. The proposed change

does not affect the design of the SGs, the

method of SG operation, or reactor coolant

chemistry controls. No new equipment is

being introduced, and installed equipment is

not being operated in a new or different

manner. The proposed change involves a

one-time extension to the SG tube inservice

inspection frequency, and therefore will not

give rise to new failure modes. In addition,

the proposed change does not impact any

other plant system or components.

Therefore, the proposed change does 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 SG tubes are an integral part of the

Reactor Coolant System (RCS) pressure

boundary that are relied upon to maintain the

RCS pressure and inventory. The SG tubes

isolate the radioactive fission products in the

reactor coolant from the secondary system.

Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices 22031

The safety function of the SGs is maintained

by ensuring the integrity of the SG tubes. In

addition, the SG tubes comprise the heat

transfer surface between the primary and

secondary systems such that residual heat

can be removed from the primary system.

SG tube integrity is a function of the

design, environment, and current physical

condition. Extending the SG tube inservice

inspection frequency by one operating cycle

will not alter the function or design of the

SGs. SG inspections conducted during the

first refueling outage following SG

replacement demonstrated that the SGs do

not have an active damage mechanism, and

the scope of those inspections significantly

exceeded those required by the TS. These

inspection results were comparable to similar

inspection results for the same model of

replacement SGs installed at other plants,

and subsequent inspections at those plants

yielded results that support this extension

request. The improved design of the

replacement SGs also provides reasonable

assurance that significant tube degradation is

not likely to occur over the proposed

operating period.

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: Mr. Robert

Helfrich, Senior Counsel, Nuclear, MidWest Regional Operating Group, Exelon

Generation Company, LLC, 1400 Opus

Place, Suite 900, Downers Grove,

Illinois 60515.

NRC Section Chief: Anthony J.

Mendiola.

Nuclear Management Company, LLC,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: April 6,

2001.

Description of amendment request:

The proposed amendment would

change the Kewaunee Nuclear Power

Plant Technical Specification Section

6.2, ‘‘Organization,’’ and Section 6.13,

‘‘High Radiation Area’’ to reflect the title

change from Shift Supervisor to Shift

Manager.

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 proposed change will not alter the

intent of the TS. Changing the title from Shift

Supervisor to Shift Manager is administrative

in nature. It has no impact on accident

initiators or plant equipment, and thus, does

not affect the probability or consequences of

an accident.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed change does not involve a

change to the physical plant or operations.

Since this is an administrative change it does

not contribute to accident initiation.

Therefore, it does not produce a new

accident scenario or produce a new type of

equipment malfunction.

(3) Involve a significant reduction in the

margin of safety.

Since this is an administrative change, it

does not involve a significant reduction in

the margin of safety. The proposed change

does not affect plant equipment or operation.

Safety limits and limiting safety system

settings are not affected by this change.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment 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.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: February

7, 2001.

Description of amendment request:

The proposed amendment would revise

the technical specifications to replace

the accident source term used in all

design basis site boundary and control

room dose analysis with the alternate

source term. Additionally, the proposed

amendment would implement

regulatory guidance provided in

Regulatory Guide (RG) 1.183,

‘‘Alternative Radiological Source Terms

for Evaluating Design Basis Accidents at

Nuclear Power Reactors,’’ regarding the

licensing basis source term for design

basis events.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes to FCS [Fort

Calhoun Station] TS [Technical

Specifications] modify requirements to: place

the control room ventilation system in

operation and in filtered air mode during

refueling operations in the containment or

spent fuel pool, place a spent fuel pool area

radiation monitor in operation during

refueling operations at the spent fuel pool,

delete a specification that requires a

ventilation isolation actuation signal (VIAS)

and two radiation monitors to be operable,

increase the volume of trisodium phosphate

(TSP) in the reactor containment building,

include both internal and external leakage for

the residual heat removal (RHR) system

leakage test, perform an internal leakage test

on the RHR system, and credit the alternative

source term (AST) for the design basis site

boundary and control room dose analyses.

These TS changes do not impact operation of

other equipment or systems important to

safety. The proposed TS changes reflect the

parameters used in the radiological

consequences calculations described in

Attachment E [to the licensee’s February 7,

2001, letter].

The current TS 3.16 limits RHR system

leakage to 1243 cc/hour from external

sources and does not provide a limit for

leakage from internal sources due to valve

seat back leakage to the safety injection

refueling water tank (SIRWT) or require an

internal leakage test to be performed. The reanalysis for LOCA [loss-of-coolant accident]

assumed a total leakage from all RHR sources

of 3800 cc/hour. The internal leakage would

leak back into the water remaining in the

SIRWT. While it appears the allowable

leakage is being increased, the limit is more

inclusive, and therefore, more conservative

than the current leakage limit. The internal

leakage test performed on the RHR system

will measure and quantify the back leakage

into the SIRWT.

The proposed changes to TSs 2.3 and 3.6

are necessary to ensure the post-LOCA pH of

the recirculation water is equal to or greater

than 7.0. Radiation levels in containment

following a LOCA may cause the generation

of hydrochloric and nitric acids from

radiolysis of cable insulation and sump

water. TSP will neutralize these acids. The

radiolysis analysis performed demonstrates

that the sump pH will be greater than or

equal to 7.0 post design basis accident (DBA),

which meets the intent of RG 1.183 regarding

iodine revolatization. Therefore, there is no

increase in the probability or consequences

of an accident previously evaluated due to

radiolysis concerns.

The proposed change to TS 2.8.2(4)

requires the control room ventilation system

to be in operation and in the Filtered Air

mode. This is a conservative action to reduce

control room operator exposure. This action

is credited in the fuel handling accident

analysis. 10 CFR 50.36 requires, in part, that

if an operating restriction is an initial

condition of a DBA, then a Limiting

Condition for Operation (LCO) should be

established. Therefore, this action, which

will reduce operator exposure, will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change to TS 2.8.3(5) will

delete the requirement for the ventilation

22032 Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices

isolation actuation signal (VIAS) to be

operable with two radiation monitors

operable, and require the control room

ventilation system to be in operation and in

the Filtered Air mode and a spent fuel pool

area radiation monitor to be in operation

during refueling operations in the spent fuel

pool. The current basis for TS 2.8.3(5) is to

ensure the control room ventilation system is

operated in Filtered Air mode upon receipt

of a VIAS. The proposed change will require

the control room ventilation system placed in

the Filtered Air mode during refueling

operations, thereby eliminating the need for

the VIAS to be operable. Therefore, this

change will not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The changes proposed do not affect the

precursors for accidents or transients

analyzed in Chapter 14 of the FCS USAR.

Therefore, there is no increase in the

probability of accidents previously evaluated.

The probability remains the same since the

accident analyses performed and discussed

in the basis for the TS changes, involve no

change to a system, component or structure

that affects initiating events for any USAR

Chapter 14 accident evaluated. A re-analysis

of USAR Chapter 14 events was conducted

with respect to radiological consequences.

This re-analysis was performed in

accordance with current accepted

methodology, and consequences were

expressed in terms of TEDE [total effective

dose equivalent] dose. The current

methodology is no longer exactly comparable

to the previous methods used for dose

consequences. The previous dose

calculations analyzed the dose consequences

to thyroid and whole body as a result of

postulated DBA events. The previous dose

calculations were shown to be well below the

regulatory limits of 10 CFR 100.11 (25

percent) with respect to thyroid and whole

body dose. The current accepted NRC

methodology, as described in 10 CFR 50.67,

specifies new dose acceptance criteria in

terms of TEDE dose. The revised analyses for

all evaluated DBA events meet the applicable

TEDE dose acceptance criteria (specified also

in RG 1.183) for alternative source term

implementation. The most current analyses

do not credit several engineered safeguards

features (ESF) filtration systems as the

previous analyses did, and hence, are more

conservative in that aspect. If a comparison

is performed between the previous

calculations (thyroid and whole body dose)

and revised analyses TEDE results (per

method shown in footnote 7 of RG 1.183), a

slight increase in dose consequences is

exhibited but is not significant, and the TEDE

results are below regulatory acceptance

criteria.

The changes proposed do not increase the

probability of an accident previously

evaluated. Because of the new regulatory

requirements related to AST implementation,

the dose consequences, if compared to

previous ones, are only slightly increased

(using guidance in footnote 7 of RG 1.183).

However, the dose consequences of the

revised analyses are below the AST

regulatory acceptance criteria.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The implementation of the proposed

changes does not create the possibility of an

accident of a different type than was

previously evaluated in the USAR. The

proposed changes to FCS TS modify

requirements to: place the control room

ventilation system in operation and in

filtered air mode during refueling operations

in the containment or spent fuel pool, place

a spent fuel pool area radiation monitor in

operation during refueling operations at the

spent fuel pool, delete a specification that

requires a ventilation isolation actuation

signal (VIAS) and two radiation monitors to

be operable, increase the volume of trisodium

phosphate (TSP) in the reactor containment

building, include both internal and external

leakage for the residual heat removal (RHR)

system leakage test, perform an internal

leakage test on the RHR system, and credit

the alternative source term (AST) for the

design basis site boundary and control room

dose analyses[.]

The changes proposed do not change how

DBA events were postulated nor do the

changes themselves initiate a new kind of

accident with a unique set of conditions. The

changes proposed were based on a complete

re-analysis of offsite and control room

operator doses, where the system

requirements being revised were not credited

in the calculations. The revised analyses are

consistent with the regulatory guidance

established in RG 1.183. The revised analyses

utilize the most current understanding of

source term timing and chemical forms as a

more appropriate mitigation technique. Not

crediting filtration systems and only

crediting natural forces is conservative from

the aspect of dose consequences. Through

this re-analysis, no new accident initiator or

failure mode was identified.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The implementation of the proposed

changes does not reduce the margin of safety.

The radiological analyses results, with the

proposed changes, remain within the

regulatory acceptance criteria (10 CFR 50

Appendix A, 10 CFR 50.67) utilizing the

TEDE dose acceptance criteria directed in RG

1.183. These criteria have been developed for

application to analyses performed with

alternative source terms. These acceptance

criteria have been developed for the purpose

of use in design basis accident analyses such

that meeting these limits demonstrates

adequate protection of public health and

safety. An acceptable margin of safety is

inherent in these licensing limits. Therefore,

there is no significant reduction in the

margin of safety as a result of the proposed

changes.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: James R.

Curtiss, Esq., Winston & Strawn, 1400 L

Street, NW., Washington, DC 20005–

3502.

NRC Section Chief: Stephen Dembek.

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: April 6,

2001.

Description of amendment requests:

The amendment application proposes to

revise the Facility Operating License

No. NPF–10, and Facility Operating

License No. NPF–15 for San Onofre

Nuclear Generating Station, Units 2 and

3, respectively. The licensee proposed

to add annotations to technical

specification Surveillance Requirements

3.8.1.2, 3.8.1.3, 3.8.1.9, 3.8.1.10 and

3.8.1.19 that provide guidance to ensure

a diesel generator sub-component, an

automatic voltage regulator (AVR), is

operable and regularly tested. The

proposed annotations clarify that only

one AVR is required for the associated

diesel generator to be operable and only

one AVR can be in service at any one

time.

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?

Response: No.

The diesel generators provide emergency

power to accident mitigation equipment in

the event of a loss of offsite power. They

cannot cause an accident. The San Onofre

Nuclear Generating Station (SONGS)

emergency diesel generators (EDG) each have

two automatic voltage regulators (AVRs) that

are 100% redundant to each other.

Maintaining both AVRs for each diesel in a

high state of readiness, while minimizing

unnecessary testing on the diesels, optimizes

the overall availability of the diesel generator

systems to perform their function if required.

This change allows testing the two AVRs

for each diesel on a staggered monthly basis.

In addition, it clarifies that each AVR only

needs to be subjected to a dynamically

challenging test once every 24 months

provided that its dynamic performance is

measured and determined to be acceptable.

These testing requirements demonstrate a

high level of assurance that each AVR will

be capable of performing its design function

while minimizing unnecessary wear on the

diesels. The reliability of the diesel

generators to provide emergency power will

not be degraded as a result of this change.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices 22033

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated?

Response: No.

The AVRs are a subcomponent of the

EDGs. This change to the surveillance test

frequency does not physically change the

use, function, or design of the EDG or its

subcomponent, the AVR.

This change ensures both 100% capacity

AVRs are adequately tested to ensure

operability without increasing the number of

test starts of the EDGs.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Involve a significant reduction in a

margin of safety?

Response: No.

This change allows testing the two AVRs

for each diesel on a staggered monthly basis.

In addition, it clarifies that each AVR only

needs to be subjected to a dynamically

challenging test once every 24 months

provided that its dynamic performance is

measured and determined to be acceptable.

These testing requirements demonstrate a

high level of assurance that each AVR will

be capable of performing its design function

while minimizing unnecessary wear on the

diesels. This proposed change does not

involve an alteration of the SONGS 2 and 3

design. The reliability of the diesel generators

to provide emergency power will not be

degraded as a result of this change.

Therefore, this 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 requests involve no

significant hazards consideration.

Attorney for licensee: Douglas K.

Porter, Esquire, Southern California

Edison Company, 2244 Walnut Grove

Avenue, Rosemead, California 91770.

NRC Section Chief: Stephen Dembek.

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 amendment request: January

11, 2001.

Description of amendment request:

The proposed amendments would

revise Technical Specifications 5.5.17,

‘‘Containment Leakage Rate Testing

Program,’’ to add an exception to

Regulatory Guide (RG) 1.163,

‘‘Performance-Based Containment LeakTesting Program.’’ Specifically, the

licensee proposes to use America

Society of Mechanical Engineering,

Subsections IWL and IWE to meet the

intent of RG 1.163. The proposed

change will affect the frequency of

containment concrete visual

examinations and allow the

examinations to be preformed during

power operation instead of exclusively

during refueling outages.

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. The proposed change affects the

frequency of visual examinations that will be

performed for the concrete surfaces of the

Vogtle Electric Generating Plant (VEGP) Unit

1 and Unit 2 containments for the purpose

of the Containment Leakage Rate Testing

Program. In addition, the proposed change

allows those examinations to be performed

during power operation as opposed to during

a refueling outage. The frequency of visual

examinations of the concrete surfaces of the

containments and the mode of operation

during which those examinations are

performed has no relationship to or adverse

impact on the probability of any of the

initiating events assumed for the accident

analyses. Therefore, the proposed change

does not involve a significant increase in the

probability of any accident previously

evaluated. The proposed change would allow

visual examinations that are performed

pursuant to NRC-approved [American

Society of Mechanical Engineering] ASME

Section XI Code requirements (except where

relief has been granted by the NRC) to meet

the intent of visual examinations required by

Regulatory Guide 1.163, without requiring

additional visual examinations pursuant to

the Regulatory Guide. The intent of early

detection of deterioration will continue to be

met by the more rigorous requirements of the

Code-required visual examinations.

Therefore, the safety function of the VEGP

containments as a fission product barrier will

be maintained, and there will not be a

significant increase in the consequences of an

accident previously evaluated.

2. Does the proposed change create the

possibility of a new or different kind of

accident from any previously evaluated?

No. The proposed change affects the

frequency of visual examinations that will be

performed for the concrete surfaces of the

Vogtle Electric Generating Plant (VEGP) Unit

1 and Unit 2 containments for the purpose

of the Containment Leakage Rate Testing

Program. In addition, the proposed change

allows those examinations to be performed

during power operation as opposed to during

a refueling outage. The proposed change does

not adversely affect or otherwise alter plant

operation. No new equipment is introduced,

and no new limiting single failures are

created. Therefore, the proposed change will

not create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Does the proposed change involve a

significant reduction in a margin of safety?

No. The proposed change affects the

frequency of visual examinations that will be

performed for the concrete surfaces of the

Vogtle Electric Generating Plant (VEGP) Unit

1 and Unit 2 containments for the purpose

of the Containment Leakage Rate Testing

Program. In addition, the proposed change

allows those examinations to be performed

during power operation as opposed to during

a refueling outage. The proposed change

would allow visual examinations that are

performed pursuant to NRC-approved ASME

Section XI Code requirements (except where

relief has been granted by the NRC) to meet

the intent of visual examinations required by

Regulatory Guide 1.163, without requiring

additional visual examinations pursuant to

the Regulatory Guide. The intent of early

detection of deterioration will continue to be

met by the more rigorous requirements of the

Code-required visual examinations.

Therefore, the safety function of the VEGP

containments as a fission product barrier will

be maintained, and there will not be 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: Mr. Arthur H.

Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600

Peachtree Street, NE., Atlanta, Georgia

30308–2216.

NRC Section Chief: Richard L. Emch,

Jr.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

April 12, 2001 (TS 00–02).

Brief description of amendments: The

proposed amendment would change the

Sequoyah Nuclear Plant Technical

Specification (TS) surveillance

requirements for the ice condenser. The

request would change the method and

frequency for determining boron

concentration and pH of the ice and

proposes an additional test requirement

for ice that is to be added to the ice

condenser.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a),

Tennessee Valley Authority (TVA), the

licensee, has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The only analyzed accidents of possible

consideration in regards to changes

potentially affecting the ice condenser are a

22034 Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices

loss of coolant accident (LOCA) and a main

steam line break (MSLB) inside containment.

However, the ice condenser is not postulated

as being the initiator of any LOCA or MSLB.

This is because it is designed to remain

functional following a design basis

earthquake, and the ice condenser does not

interconnect or interact with any systems

that interconnect or interact with the reactor

coolant or main steam systems. Since the

proposed changes to the TS and TS bases are

solely to revise and provide clarification of

the ice sampling and chemical analysis

requirements, and are not the result of or

require any physical change to the ice

condenser, there can be no change in the

probability of an accident previously

evaluated in the Safety Analysis Report.

In order for the consequences of any

previously evaluated event to be changed,

there would have to be a change in the ice

condenser’s physical operation during a

LOCA or MSLB, or in the chemical

composition of the stored ice. The proposed

changes do not alter either from existing

requirements, except to add an upper limit

on boron concentration, which is the

bounding value for the hot leg switchover

timing calculation. Though the frequency of

the existing surveillance requirement (SR) for

sampling the stored ice is changed from once

every 18 months to once every 54 months,

the sampling requirements are strengthened

overall with: (1) the requirement to obtain

one randomly selected sample from each ice

condenser bay (24 total samples) rather than

9 ‘‘representative’’ samples, and (2) the

addition of a new SR to verify each addition

of ice meets the existing requirements for

boron concentration and pH value. The only

other change is to clarify that each sample of

stored ice is individually analyzed for boron

concentration and pH, but that the

acceptance criteria for each parameter is

based on the average values obtained for the

24 samples. This is consistent with the bases

for the boron concentration of the ice, which

is to ensure the accident analysis

assumptions for containment sump pH and

boron concentration are not altered following

complete melting of the ice condenser.

Historically, chemical analysis of the stored

ice has had a very limited number of

instances where an individual sample did

not meet the boron or pH requirements, with

all subsequent evaluations (follow-up

sampling) showing the ice condenser as a

whole was well within these requirements.

Requiring chemical analysis of each sample

is provided to preclude the practice of

melting all samples together before

performing the analysis, and to ensure the

licensee is alerted to any localized anomalies

for investigation and resolution without the

burden of entering a 24-hour action, provided

the averaged results are acceptable. Thus,

based on the above, the proposed changes do

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Because the TS and TS bases changes do

not involve any physical changes to the ice

condenser, any physical or chemical changes

to the ice contained therein, or make any

changes in the operational or maintenance

aspects of the ice condenser as required by

the TS, there can be no new accidents created

from those already identified and evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The ice condenser TSs ensure that during

a LOCA or MSLB the ice condenser will

initially pass sufficient air and steam mass to

preclude over pressurizing lower

containment, that it will absorb sufficient

heat energy initially and over a prescribed

time period to assist in precluding

containment vessel failure, and that it will

not alter the bulk containment sump pH and

boron concentration assumed in the accident

analysis. Since the proposed changes do not

physically alter the ice condenser, but rather

only serve to strengthen and clarify ice

sampling and analysis requirements, the only

area of potential concern is the effect these

changes could have on bulk containment

sump pH and boron concentration following

ice melt. However, this is not affected

because there is no change in the existing

requirements for pH and boron

concentration, except to add an upper limit

on boron concentration. This upper limit is

the bounding value for the hot leg switchover

timing calculation. Averaging the pH and

boron values obtained from analysis of the

individual samples taken is not a new

practice, just one that was not consistently

used by all ice condenser plants. Using the

averaged values provides an equivalent bulk

value for the ice condenser, which is

consistent with the accident analysis for the

bulk pH and boron concentration of the

containment sump following ice melt.

Changing the performance frequency for

sampling the stored ice does not reduce any

margin of safety because: (1) the newly

proposed surveillance (SR 4.6.5.1.f) ensures

ice additions meet the existing boron

concentration and pH requirements, (2) there

are no normal operating mechanisms,

including sublimation, that reduce the ice

condenser bulk pH and boron concentration,

and (3) the number of required samples has

been increased from 9 to 24 (1 randomly

selected ice basket per bay), which is

approximately the same number of samples

that would have been taken in the same time

period under the existing requirements.

Thus, it can be concluded that the proposed

TS and TS bases changes do not involve a

significant reduction in the margin of safety.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Section Chief: Richard P.

Correia.

TXU Electric, Docket Nos. 50–445 and

50–446, Comanche Peak Steam Electric

Station, Units 1 and 2, Somervell

County, Texas

Date of amendment request: April 3,

2001.

Brief description of amendments: The

proposed change would revise

Technical Specification (TS) 3.3.6,

‘‘Containment Ventilation Isolation

Instrumentation’’ to modify the Note for

Required Action B.1 such that it applies

only to ‘‘Required Action and associated

Completion Time of Condition A not

met.’’ The proposed change is the result

of the discovery of an error which

occurred when the TS was converted to

the ‘‘improved TS format’’ with

issuance of License Amendment Nos. 64

and 64, for Comanche Peak Steam

Electric Station, Units 1 and 2 on

February 26, 1999.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

The proposed change removes an

allowance to open containment pressure

relief valves under administrative controls

when one train of Automatic Actuation Logic

and Actuation Relays is inoperable. The

proposed change corrects a non-conservative

technical specification and thus makes the

technical specifications consistent with the

previously evaluated accident analyses.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

The proposed change makes the technical

specifications consistent with the previously

evaluated accident analyses. Therefore, the

proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

Response: No.

The proposed change makes the technical

specifications consistent with the previously

evaluated accident analyses. Therefore the

proposed change does not involve a

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices 22035

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: March

22, 2001 (ET 01–0012).

Description of amendment request:

The amendment would (1) decrease the

allowable values for Function 8,

pressurizer pressure-low, pressurizer

pressure-high, in Table 3.3.1–1,

‘‘Reactor Trip System Instrumentation,’’

and (2) decrease the allowable value for

pressurizer pressure-low for safety

injection in Table 3.3.2–1, ‘‘Engineered

Safety Feature Actuation System

Instrumentation.’’ The changes are

needed because the licensee will be

replacing the existing Tobar pressurizer

pressure transmitters with Rosemount

transmitters in the next refueling outage.

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 existing safety related pressurizer

pressure transmitters are being replaced with

ones of similar characteristics and functions,

and without changing the design or

functional basis of the system, structure, or

components associated with the pressure

transmitters.

The protection system performance will

remain within the bounds of the previously

performed accident analysis. The Reactor

Trip System (RTS) and Engineered Safety

Feature Actuation System (ESFAS)

instrumentation will continue to function in

a manner consistent with the plant design

basis. The replacement of the pressurizer

pressure transmitters and proposed changes

to the affected Allowable Values will not

affect any of the analysis assumptions for any

of the accidents previously evaluated, since

the changes are consistent with the setpoint

methodology and ensure adequate margin to

the Safety Analysis Limit. The proposed

changes will not affect any event initiators

nor will the proposed changes affect the

ability of any safety related equipment to

perform its intended function. There will be

no degradation in the performance of nor an

increase in the number of challenges

imposed on safety related equipment

assumed to function during an accident

situation. There will be no change to normal

plant operating parameters or accident

mitigation capabilities.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

A review of the failure modes and effects

in Updated Safety Analysis Report Section

7.7.2 found that failure of the replacement

pressure transmitters will be the same as for

the existing pressure transmitters. As such,

the effects of such failures on [the safety]

functions of the other equipment are

concluded to be similar to those previously

evaluated.

There are no changes in the method by

which any safety related plant system

performs its safety function. The normal

manner of plant operation remains

unchanged. The increase [or decrease] in the

pressurizer pressure functions Allowable

Values still provides acceptable margin

between the nominal Trip Setpoint and

Allowable Value. The changes in Allowable

Value does not impact the systems capability

to provide both control and protection

functions. No new accident scenarios,

transient precursors, failure mechanisms, or

limiting single failures are introduced as a

result of the proposed changes.

Therefore, the proposed changes do not

create 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 changes do not affect the

acceptance criteria for any analyzed event

nor is there a change in any Safety Analysis

Limit. There will be no effect on the manner

in which safety limits or RTS and ESFAS

settings are determined nor will there be any

affect on those plant systems necessary to

assure the accomplishment of protection

functions. [The proposed changes to the

pressurizer pressure Allowable Values will

maintain the accident analyses in the

Updated Safety Analysis Report.]

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Section Chief: Stephen Dembek.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: March

23, 2001 (CO 01–0013).

Description of amendment request:

The amendment proposes to (1) delete

certain license conditions from the

operating license, and (2) revise Table

5.5.9–2, ‘‘Steam Generator Tube

Inspection,’’ in Section 5.5.9, ‘‘Steam

Generator (SG) Tube Surveillance

Program,’’ of the technical specifications

(TSs). License Conditions 2.C.(4) and

2.C.(6) through 2.C.(14) of the facility

operating license are considered to have

been completed and obsolete, or to

duplicate other license requirements,

and are proposed to be deleted.

Attachments 2 and 3 to the facility

operating license are also proposed to be

deleted. Section 2.F of the facility

operating license is considered to

duplicate the reporting requirements in

10 CFR 50.72 and 50.73 and is proposed

to be deleted. The reporting

requirements in two ‘‘Action Required’’

columns of TS Table 5.5.9–2 are also

considered to duplicate the reporting

requirements in 10 CFR 50.72 and 50.73

and are proposed to be deleted. The list

of the attachments and appendices to

the facility operating license would also

be revised to reflect the proposed

deletion of Attachments 2 and 3.

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.

This request involves administrative

changes only. No actual plant equipment or

accident analyses will be affected by the

proposed changes. Therefore, the proposed

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

This request involves administrative

changes only. No actual plant equipment or

accident analyses will be affected by the

proposed change and no failure modes not

bounded by previously evaluated accidents

will be created. Therefore, the proposed

changes do not create 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.

Margin of safety is associated with

confidence in the ability of the fission

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

Reactor Coolant System pressure boundary,

and containment structure) to limit the level

of radiation dose to the public. This request

involves administrative changes only [and

does not change these barriers].

No actual plant equipment or accident

analyses will be affected by the proposed

change. Additionally, the proposed changes

22036 Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices

will not relax any criteria used to establish

safety limits, will not relax any safety system

settings, or will not relax the bases for any

limiting conditions of operation [in the TSs].

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Section Chief: Stephen Dembek.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: April 3,

2001 (ET 01–0008).

Description of amendment request:

The amendment would make the

following changes to the technical

specifications (TSs):

(1) Revise Safety Limit 2.1.1 by

replacing Figure 2.1.1–1, ‘‘Reactor Core

Safety Limits,’’ with a reference to limits

being specified in the Core Operating

Limits Report (COLR) and by adding

two reactor core safety limits on

departure from nucleate boiling ratio

(DNBR) and peak fuel centerline

temperature.

(2) Revise Note 1 on the over

temperature ∆T in Table 3.3.1–1 of TS

3.3.1, ‘‘Reactor Trip System

Instrumentation,’’ by replacing values of

parameters with a reference to the

values being specified in the COLR and

correcting the expression for one term in

the inequality for over temperature ∆T.

(3) Revise Note 2 on the overpower ∆T

in Table 3.3.1–1 by replacing values of

parameters with a reference to the

values being specified in the COLR.

(4) Replace the limits for the reactor

coolant system (RCS) pressure and

average temperature with a reference to

the limits being specified in the COLR

for Limiting Condition for Operation

(LCO) 3.4.1 and Surveillance

Requirements (SRs) 3.4.1.1 and 3.4.1.2.

(5) Add the phrase ‘‘and greater than

or equal to the limit specified in the

COLR’’ to the RCS total flow rate in LCO

3.4.1 and SRs 3.4.1.3 and 3.4.1.4.

(6) Move items a. and b. to the left in

the Note to the applicability in LCO

3.4.1.

(7) Revise TS Section 5.6.5 by adding

TS 3.3.1 on over temperature and

overpower ‘‘T trip setpoints and TS

3.4.1 on RCS pressure, temperature, and

flow limits to the existing list of core

operating limits for each reload cycle

that are documented in the COLR and

revising the list of topical reports in the

COLR that represent the analytical

methods approved by the Commission

to determine core operating limits.

The proposed changes remove cyclespecific parameter limits and relocate

them to the COLR, but they (1) do not

change any of the limits, (2) add more

specific requirements regarding DNBR

limit and peak fuel centerline

temperature limit to the TSs, (3) revise

the list of topical reports in the list of

NRC-approved analytical methods, (4)

correct one term of an expression, and

(5) move terms in a Note to the mode

applicability for an LCO.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes are programmatic

and administrative in nature which do not

physically alter safety related systems, nor

affect the way in which safety related

systems perform their functions. More

specific requirements regarding the safety

limits (i.e., DNBR limit and peak fuel

centerline temperature limit) are being

imposed in TS 2.1.1, ‘‘Reactor Core Safety

Limits,’’ which replace the Reactor Core

Safety Limits figure and are consistent with

the values stated in the USAR [Updated

Safety Analysis Report]. The proposed

changes remove the cycle-specific parameter

limits from TS 3.4.1 and relocate them to the

COLR which do not change plant design or

affect system operating parameters. In

addition, the minimum limit for RCS total

flow rate is being retained in TS 3.4.1 to

assure that a lower flow rate than reviewed

by the NRC will not be used. The proposed

changes do not, by themselves, alter any of

the parameter limits. The removal of the

cycle-specific parameter limits from the TS

does not eliminate existing requirements to

comply with the parameter limits. The

existing TS Section 5.6.5b, COLR Reporting

Requirements, continues to ensure that the

analytical methods used to determine the

core operating limits meet NRC reviewed and

approved methodologies. The existing TS

Section 5.6.5c, COLR Reporting

Requirements, continues to ensure that

applicable limits of the safety analyses are

met.

The proposed changes to reference only the

Topical Report number and title do not alter

the use of the analytical methods used to

determine core operating limits that have

been reviewed and approved by the NRC.

This method of referencing Topical Reports

would allow the use of current Topical

Reports to support limits in the COLR

without having to submit an amendment to

[the TS of] the operating license.

Implementation of revisions to Topical

Reports would still be reviewed in

accordance with 10 CFR 50.59 and where

required receive NRC review and approval.

Although the relocation of the cyclespecific parameter limits to the COLR would

allow revision of the affected parameter

limits without prior NRC approval, there is

no significant effect on the probability or

consequences of an accident previously

evaluated. Future changes to the COLR

parameter limits could result in event

consequences which are either slightly less

or slightly more severe than the

consequences for the same event using the

present parameter limits. The differences

would not be significant and would be

bounded by the existing requirement of TS

Section 5.6.5c to meet the applicable limits

of the safety analyses.

The cycle-specific parameter limits being

transferred from the TS to the COLR will

continue to be controlled under existing

programs and procedures. The USAR

accident analyses will continue to be

examined with respect to changes in the

cycle-dependent parameters obtained using

NRC reviewed and approved reload design

methodologies, ensuring that the transient

evaluation of new reload designs are

bounded by previously accepted analyses.

This examination will continue to be

performed pursuant to 10 CFR 50.59

requirements ensuring that future reload

designs will not involve a significant increase

in the probability or consequences of an

accident previously evaluated. Additionally,

the proposed changes do not allow for an

increase in plant power levels, do not

increase the production, nor alter the flow

path or method of disposal of radioactive

waste or byproducts. Therefore, the proposed

changes do not change the types or increase

the amounts of any effluents released offsite.

[The proposed changes to the expression of

the f1(∆I) term, which is in the over

temperature ∆T inequality, clarifies and

corrects the term. Moving the terms in a Note

to the LCO mode applicability is an

administrative action.]

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

[The proposed changes are programmatic

and administrative in nature which do not

physically alter safety related systems, nor

affect the way in which safety related

systems perform their functions.]

The proposed changes that retain the

minimum limit for RCS total flow rate in the

TS, and that relocate certain cycle-specific

parameter limits from the TS to the COLR,

thus removing the requirement for prior NRC

approval of revisions to those parameters, do

not involve a physical change to the plant.

No new equipment is being introduced, and

installed equipment is not being operated in

a new or different manner. There are no

Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices 22037

changes being made to the parameters within

which the plant is operated, other than their

relocation to the COLR. There are no

setpoints affected by the proposed changes at

which protective or mitigative actions are

initiated. The proposed changes will not alter

the manner in which equipment operation is

initiated, nor will the function demands on

credited equipment be changed. No alteration

in the procedures which ensure the plant

remains within analytical limits is being

proposed, and no change is being made to the

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

are being introduced.

The proposed changes to reference only the

Topical Report number and title do not alter

the use of the analytical methods used to

determine core operating limits that have

been reviewed and approved by the NRC.

This method of referencing Topical Reports

would allow the use of current Topical

Reports to support limits in the COLR

without having to submit an amendment to

[the TS of] the operating license.

Implementation of revisions to Topical

Reports would still be reviewed in

accordance with 10 CFR 50.59 and where

required receive NRC review and approval.

Relocation of cycle-specific parameter

limits has no influence or impact on, nor

does it contribute in any way to the

possibility of a new or different kind of

accident. The relocated cycle-specific

parameter limits will continue to be

calculated using the NRC reviewed and

approved methodology. The proposed

changes do not alter assumptions made in the

safety analysis and operation within the core

operating limits will continue.

[The proposed changes to the expression of

the f1(∆I) term, which is in the over

temperature ∆T inequality, clarifies and

corrects the term.]

Therefore, the proposed changes do not

create 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 margin of safety is established through

equipment design, operating parameters, and

the setpoints at which automatic actions are

initiated. The proposed changes [are

programmatic and administrative in nature

and] do not physically alter safety related

systems, nor does it [a]ffect the way in which

safety-related systems perform their

functions. The setpoints at which protective

actions are initiated are not altered by the

proposed changes. Therefore, sufficient

equipment remains available to actuate upon

demand for the purpose of mitigating an

analyzed event. As the proposed changes to

relocate cycle-specific parameter limits to the

COLR will not affect plant design or system

operating parameters, there is no detrimental

impact on any equipment design parameter,

and the plant will continue to operate within

prescribed limits.

The development of cycle-specific

parameter limits for future reload designs

will continue to conform to NRC reviewed

and approved methodologies, and will be

performed pursuant to 10 CFR 50.59 to

assure that plant operation [is] within cyclespecific parameter limits.

The proposed changes to reference only the

Topical Report number and title do not alter

the use of the analytical methods used to

determine core operating limits that have

been reviewed and approved by the NRC.

This method of referencing Topical Reports

would allow the use of [the] current Topical

Reports to support limits in the COLR

without having to submit an amendment to

[the TS of] the operating license.

Implementation of revisions to Topical

Reports would still be reviewed in

accordance with 10 CFR 50.59 and where

required receive NRC review and approval.

[The proposed changes to the expression of

the f1(∆I) term, which is in the over

temperature ∆T inequality, clarifies and

corrects the term. Moving the terms in a Note

to the LCO mode applicability is an

administrative action.]

Therefore, the proposed changes do not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Section Chief: Stephen Dembek.

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:

February 28, 2001.

Brief description of amendments: The

amendments revise the definitions of

engineered safety feature response time

and reactor protection system response

time in Technical Specification (TS) 1.1,

‘‘Definitions,’’ to add the following

statement: ‘‘In lieu of measurement,

response time may be verified for

selected components provided that the

components and methodology for

verification have been previously

reviewed and approved by the [Nuclear

Regulatory Commission ] NRC.’’

Approval of the amendments will allow

either an allocated sensor response time

or a measured sensor response time for

the identified Reactor Protection System

and Engineered Safety Features

Actuation System pressure sensors

when performing response time testing.

Date of issuance: April 19, 2001.

Effective date: April 19, 2001, and

shall be implemented within 45 days of

the date of issuance.

Amendment Nos.: Unit 1–135, Unit

2–135, Unit 3–135.

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 20, 2001 (66 FR

15766).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 19, 2001.

No significant hazards consideration

comments received: No. Calvert Cliffs

Nuclear Power Plant, Inc., Docket Nos.

50–317 and 50–318, Calvert.

22038 Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland

Date of application for amendments:

December 21, 2000, as supplemented on

February 12, 2001, and March 5, 2001.

Brief description of amendments: The

amendments revise Technical

Specification 5.2.2.e by removing the

reference to the Nuclear Regulatory

Commission Policy Statement on

working hours.

Date of issuance: April 5, 2001.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 245 and 219.

Renewed Facility Operating License

Nos. DPR–53 and DPR–69: Amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: February 7, 2001 (66 FR

9380).

The February 12, 2001, and March 5,

2001, letters provided clarifying

information that did not change the

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of these amendments is contained in a

Safety Evaluation dated April 5, 2001.

No significant hazards consideration

comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc.,

Docket No. 50–318, Calvert Cliffs

Nuclear Power Plant, Unit No. 2, Calvert

County, Maryland

Date of application for amendment:

September 14, 2000, as supplemented

on December 21, 2000.

Brief description of amendment: The

amendment permits operation of Calvert

Cliffs Unit 2 with a core containing a

lead fuel (test) assembly that includes

fuel rods with advanced zirconium alloy

cladding.

Date of issuance: April 5, 2001.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 220.

Renewed License No. DPR–69:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 10, 2001 (66 FR

2012).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 5, 2001.

No significant hazards consideration

comments received: No.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois.

Date of application for amendments:

June 19, 2000, as supplemented by

letters dated March 16, 2001, and April

4, 2001.

Brief description of amendments: The

proposed amendments revised the

technical specifications to remove their

applicability related to the Boron

Dilution Protection System (BDPS) after

the next refueling outage for each unit.

During the refueling outages,

modifications are scheduled to be made

which will permit mitigation of a boron

dilution event without the use of the

BDPS.

Date of issuance: April 6, 2001.

Effective date: Immediately, to be

implemented upon completion of the

modifications scheduled to be

completed after cycle 9 for Byron, Unit

2, and Braidwood, Units 1 and 2, and

after cycle 11 for Byron, Unit 1.

Amendment Nos.: 117 and 111.

Facility Operating License Nos. NPF–

37, NPF–66, NPF–72 and NPF–77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 6, 2000 (65 FR

54084).

Since the proposed additional

changes provided in this supplement

are more restrictive than the originally

proposed changes, it does not change

the previous determination of no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 6, 2001.

No significant hazards consideration

comments received: No.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

February 12, 2001.

Brief description of amendment: The

amendment changes Technical

Specification Section 5.6.5b, ‘‘Reporting

Requirements—Core Operating Limits

Report (COLR),’’ to add a report

pertaining to statistical setpoint

methodology to the list of approved

methodology references.

Date of issuance: April 9, 2001.

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 195.

Facility Operating License No. DPR–

20. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 7, 2001 (66 FR 13801).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 9, 2001.

No significant hazards consideration

comments received: No.

Duke Energy Corporation, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina

Date of application for amendments:

August 22, 2000, as supplemented by

letter dated November 7, 2000.

Brief description of amendments: The

amendments revise the Technical

Specifications (TS) of each unit to

restore a time limit for an allowable

condition for the occurrence of an

inoperable refueling water storage tank

level transmitter in TS 3.3.2.

Date of issuance: April 12, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment Nos.: 198 and 179.

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: November 1, 2000 (65 FR

65341).

The supplement dated November 7,

2000, provided clarifying information

that did not change the scope of the

August 22, 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 April 12, 2001.

No significant hazards consideration

comments received: No.

Entergy Nuclear Operations, Inc.,

Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of application for amendment:

September 6, 2000, as supplemented on

January 18, and April 2, 2001.

Brief description of amendment: The

amendment revises Technical

Specification 5.5.15 to allow a one time

change in the 10 CFR part 50, Appendix

J, Type A test interval from the required

10 years to a test interval of 15 years.

Date of issuance: April 17, 2001.

Effective date: April 17, 2001.

Amendment No.: 206.

Facility Operating License No. DPR–

64:

Date of initial notice in Federal

Register: January 24, 2000 (66 FR

7665).

The January 18, and April 2, 2001,

submittals contained clarifying

information only, and did not change

the initial no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 17, 2001.

Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices 22039

No significant hazards consideration

comments received: No.

Entergy Nuclear Generation Company,

Docket No. 50–293, Pilgrim Nuclear

Power Station, Plymouth County,

Massachusetts

Date of application for amendment:

February 16, 2001.

Brief description of amendment: This

amendment substitutes a surveillance

interval of ‘‘Once/Operating Cycle’’ for

the current surveillance interval of

‘‘Each Refueling Outage,’’ for the

following instruments in Technical

Specification Table 4.2.F: Containment

High Radiation Monitor, Reactor

Building Vent Radiation Monitor, Main

Stack Vent Radiation Monitor, and

Turbine Building Vent Radiation

Monitor.

Date of issuance: April 9, 2001.

Effective date: As of the date of

issuance, and shall be implemented

within 30 days.

Amendment No.: 189.

Facility Operating License No. DPR–

35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 7, 2001 (66 FR 13802).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 9, 2001.

No significant hazards consideration

comments received: No.

Entergy Nuclear Generation Company,

Docket No. 50–293, Pilgrim Nuclear

Power Station, Plymouth County,

Massachusetts

Date of application for amendment:

November 22, 2000, as supplemented on

January 30 and February 2, 2001.

Brief description of amendment: This

amendment changes the pressuretemperature limit curves of Figures

3.6.1, 3.6.2, and 3.6.3 of the Technical

Specifications (TS) over operation

between 20, 32, and 48 Effective Full

Power Years. However, these curves

will only apply for the remainder of

operating cycles 13 and 14. The Bases

section has been modified to reflect

these TS changes.

Date of issuance: April 13, 2001.

Effective date: As of the date of

issuance, and shall be implemented

within 30 days.

Amendment No.: 190.

Facility Operating License No. DPR–

35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: December 27, 2000 (65 FR

81915).

The January 30 and February 2, 2001,

letters provided clarifying information

that did not change the initial proposed

no significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 13, 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:

August 10, 2000, as supplemented by

letter dated March 22, 2001.

Brief description of amendment: The

amendment revised Technical

Specification 3/4.9.4, ‘‘Refueling

Operations, Containment Building

Penetrations,’’ by deleting the

requirements for the containment purge

and exhaust system and by revising the

closure requirements for containment

building penetrations to require that

containment penetrations are capable of

being closed during the handling of

irradiated fuel within the containment.

Date of issuance: April 18, 2001.

Effective date: As of the date of

issuance to be implemented within 60

days from the date of issuance.

Amendment No.: 230.

Facility Operating License No. NPF–6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 20, 2000 (65 FR

56950).

The March 22, 2001, supplemental

letter provided clarifying information

that was within the scope of the original

Federal Register notice and did not

change the staff’s initial no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 18, 2001.

No significant hazards consideration

comments received: No.

Exelon Generation Company, LLC,

Docket No. 50–353, Limerick Generating

Station, Unit 2, Montgomery County,

Pennsylvania

Date of application for amendment:

February 1, 2001, as supplemented

March 6 and 23, 2001.

Brief description of amendment: This

amendment revises the minimum

critical power ratio safety limits for

operating cycle 7.

Date of issuance: April 12, 2001.

Effective date: As of date of issuance,

and shall be implemented within 30

days.

Amendment No.: 114.

Facility Operating License No. NPF–

85. This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 21, 2001 (66 FR

11061).

The March 6 and 23, 2001, letters

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 12, 2001.

No significant hazards consideration

comments received: No.

Florida Power and Light Company,

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

No. 1, St. Lucie County, Florida

Date of application for amendment:

December 4, 2000, as supplemented

February 9, 2001.

Brief description of amendment: This

amendment changes the licensing bases

to incorporate a revised analysis of the

Main Steam Line Break inside

containment.

Date of Issuance: April 20, 2001.

Effective Date: April 20, 2001.

Amendment No.: 175.

Facility Operating License No. DPR–

67: Amendment revised the Updated

Final Safety Analysis Report.

Date of initial notice in Federal

Register: February 7, 2001 (66 FR

9383).

The February 9, 2001, Supplement

did not affect the original proposed no

significant hazards determination, or

expand the scope of the request as

noticed in the Federal Register.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 20, 2001.

No significant hazards consideration

comments received: No.

North Atlantic Energy Service

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

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request:

December 18, 2000.

Description of amendment request:

The amendment deletes Technical

Specifications Section 6.7.6.e, ‘‘PostAccident Sampling,’’ for Seabrook

Station, Unit No. 1 and thereby

eliminates the requirements to have and

maintain the post-accident sampling

system.

Date of issuance: April 17, 2001.

Effective date: As of its date of

issuance, and shall be implemented

within 60 days.

Amendment No.: 78.

Facility Operating License No. NPF–

86: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 24, 2000 (66 FR

7683).

22040 Federal Register / Vol. 66, No. 85 / Wednesday, May 2, 2001 / Notices

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 17, 2001.

No significant hazards consideration

comments received: No.

Nuclear Management Company, LLC,

Docket No. 50–331, Duane Arnold

Energy Center, Linn County, Iowa

Date of application for amendment:

October 19, 2000, as supplemented

November 16, 2000, and April 9, 2001,

and as limited in scope by letter dated

March 23, 2001.

Brief description of amendment: The

amendment revises the Technical

Specifications regarding operability

requirements during core alterations

and while moving irradiated fuel

assemblies within the secondary

containment. The amendment also

provides for a change in design and

licensing bases for a selective

application of the alternate radiological

source term in accordance with 10 CFR

50.67, ‘‘Accident Source Term,’’ and

revised meteorology dispersion values,

both being limited to a design-basis fuel

handling accident.

Date of issuance: April 16, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 60 days.

Amendment No.: 237.

Facility Operating License No. DPR–

49: The amendment revised the

Technical Specifications and the

licensing and design bases regarding a

design-basis fuel handling accident.

Date of initial notice in Federal

Register: March 6, 2001 (66 FR 13598).

NMC’s letters dated March 23 and

April 9, 2001, are within the scope of

the changes proposed in NMC’s letter of

October 19, 2000, that was noticed in

the Federal Register on March 6, 2001.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 16, 2001.

No significant hazards consideration

comments received: No.

Nuclear Management Company, LLC,

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

Island Nuclear Generating Plant, Units

1 and 2, Goodhue County, Minnesota

Date of application for amendments:

April 17, 2000, as supplemented

February 2, 2001.

Brief description of amendments: The

amendments change the Technical

Specifications (TSs) for removal of boric

acid storage tanks from the safety

injection (SI) system. These changes

accomplish two objectives: (1) Eliminate

high concentration boric acid from the

SI system and (2) align this specific TS

section with the standard TSs.

Date of issuance: April 16, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 90 days.

Amendment Nos.: 156 and 147.

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: March 7, 2001 (66 FR 13806).

The February 2, 2001, supplement

provided clarifying information that was

within the scope of the original

application and did not change the

staff’s initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 16, 2001.

No significant hazards consideration

comments received: No.

PPL Susquehanna, LLC, Docket Nos. 50–

387 and 50–388, Susquehanna Steam

Electric Station, Units 1 and 2, Luzerne

County, Pennsylvania

Date of application for amendments:

October 4, 2000, as supplemented

March 12, April 2, and April 5, 2001.

Brief description of amendments: The

amendments revise the surveillance test

requirements for excess flow check

valves (EFCVs) to allow testing of a

representative sample at 24-month

intervals such that each EFCV is tested

at least once every 10 years.

Date of issuance: April 11, 2001.

Effective date: As of date of issuance

to be implemented within 30 days.

Amendment Nos.: 193 and 168.

Facility Operating License Nos. NPF–

14 and NPF–22. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: January 10, 2001 (66 FR

2021).

The March 12, April 2, and April 5,

letters provided clarifying information

that did not change the initial proposed

no significant hazards consideration

determination or expand the

amendment beyond the scope of the

initial notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated April 11, 2001.

No significant hazards consideration

comments received: No.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

January 18, 2001, as supplemented by

letter dated February 21, 2001.

Brief description of amendment: The

amendment authorizes the installation

of new engineered safety feature

transformers as an improvement. This

amendment will allow the installation

and use of the new transformers

equipped with automatic load tap

changers and an update to the Final

Safety Analysis Report (FSAR) to reflect

their installation.

Date of issuance: April 6, 2001.

Effective date: April 6, 2001, and shall

be implemented in the next periodic

update to the FSAR in accordance with

10 CFR 50.71(e).

Amendment No.: 143.

Facility Operating License No. NPF–

30: The amendment revised the FSAR.

Date of initial notice in Federal

Register: February 21, 2001 (66 FR

11063).

The February 21, 2001, supplemental

letter provided additional clarifying

information, did not expand the scope

of the application as originally noticed,

and did not change the staff’s original

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 6, 2001.

No significant hazards consideration

comments received: No.

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

January 18, 2001.

Brief description of amendment: The

amendment deletes Section 5.5.3, ‘‘Post

Accident Sampling,’’ of the Technical

Specifications for the Callaway Plant

and thereby eliminates the requirements

to have and maintain the post-accident

sampling system (PASS).

Date of issuance: April 6, 2001.

Effective date: April 6, 2001, to be

implemented within 60 days of the date

of issuance.

Amendment No.: 144.

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: March 7, 2001 (66 FR 13808).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated April 6, 2001.

No significant hazards consideration

comments received: No.

Dated at Rockville, Maryland this 25th day

of April 2001.

For The Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation

[FR Doc. 01–10822 Filed 5–1–01; 8:45 am]

BILLING CODE 7590–01–P