31700 Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices

provisions of 10 CFR part 20, subpart K,

General License for Storage of Spent

Fuel at Power Reactor Sites, as specified

in 10 CFR 72.212(b)(5), MYAPS is

required to meet the physical protection

requirements of 10 CFR 73.55 for an

ISFSI at a reactor site. However, MYAPC

has proposed to be exempted from the

requirements of 10 CFR 72.212(b)(5) to

‘‘protect the spent fuel against the

design basis threat of radiological

sabotage in accordance with the same

provisions and requirements as are set

forth’’ in 10 CFR 73.55 (with certain

exceptions provided by 10 CFR

72.212(b)(5)). Instead, MYAPC has

proposed alternative approaches to meet

the provisions of portions of 10 CFR

73.55 related to the security

organization, physical barriers, access

requirements, detection aids,

communications, and response

requirements. The alternative measures

for protection against radiological

sabotage would meet the same high

assurance objectives and the general

performance requirements of 10 CFR

73.55 related to spent fuel storage at an

ISFSI.

Environmental Impacts of the Proposed

Action

The NRC has completed its evaluation

of the proposed action and concludes

that granting an exemption from the

requirements of 10 CFR 72.212(b)(5) to

protect the spent fuel against the design

basis threat of radiological sabotage in

accordance with the requirements of 10

CFR 75.55, thereby enabling MYAPC to

implement alternative provisions of 10

CFR 73.55, would not have a significant

impact on the environment.

The proposed action will not

significantly increase the probability or

consequences of accidents, no changes

are being made in the types of any

effluents that may be released off site,

and there is no significant increase in

occupational or public radiation

exposure. Therefore, there are no

significant radiological environmental

impacts associated with the proposed

action.

With regard to potential

nonradiological impacts, the proposed

action does not involve any historic

sites. It does not affect nonradiological

plant effluents and has no other

environmental impact. Therefore, there

are no significant nonradiological

environmental impacts associated with

the proposed action.

Accordingly, the NRC concludes that

there are no significant environmental

impacts associated with the proposed

action.

Alternatives to the Proposed Action

As an alternative to the proposed

action, the staff considered denial of the

proposed action (i.e., the ‘‘no-action’’

alternative). Denial of the application

would result in no change in current

environmental impacts. The

environmental impacts of the proposed

action and the alternative action are

similar.

Alternative Use of Resources

This action does not involve the use

of any resources not previously

considered in the Final Environmental

Statement related to Operation of Maine

Yankee Atomic Power Station (July

1972).

Agencies and Persons Contacted

In accordance with its stated policy,

on April 19, 2001, the staff consulted

with Mr. Patrick Dostie of the State of

Maine, Department of Human Services,

Division of Health Engineering,

regarding the environmental impact of

the proposed action. The State official

had no comments.

Finding of No Significant Impact

On the basis of the environmental

assessment, the NRC concludes that the

proposed action will not have a

significant effect on the quality of the

human environment. Accordingly, the

NRC has determined not to prepare an

environmental impact statement for the

proposed action.

For further details with respect to the

proposed action, see the licensee’s

letters dated January 4, 2001, March 12,

2001, and April 4, 2001, which may be

examined, and/or copied for a fee, at the

NRC’s Public Document Room, located

at One White Flint North, 11555

Rockville Pike (first floor), Rockville,

Maryland. Publicly available records

will be accessible electronically from

the ADAMS Public Library component

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

(the Electronic Reading Room).

Dated at Rockville, Maryland, this 5th day

of June 2001.

For the Nuclear Regulatory Commission.

Michael K. Webb,

Project Manager, Section 1, Project

Directorate IV & Decommissioning, Division

of Licensing Project Management, Office of

Nuclear Reactor Regulation.

[FR Doc. 01–14753 Filed 6–11–01; 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 May 21,

2001 through June 1, 2001. The last

biweekly notice was published on May

30, 2001 (66 FR 29349).

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.

Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices 31701

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules and

Directives Branch, Division of

Administrative Services, Office of

Administration, U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001, and should cite the publication

date and page number of this Federal

Register notice. Written comments may

also be delivered to Room 6D22, Two

White Flint North, 11545 Rockville

Pike, Rockville, Maryland from 7:30

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

Copies of written comments received

may be examined at the NRC Public

Document Room, 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 July 13, 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.

Publicly available records will be

accessible from the Agencywide

Documents Access and Management

Systems (ADAMS) Public Electronic

31702 Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices

Reading Room on the internet at the

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

ADAMS/index.html. If you do not have

access to ADAMS or if there are

problems in accessing the documents

located in ADAMS, contact the NRC

Public Document Room (PDR) Reference

staff at 1–800–397–4209, 301–415–4737

or by email to pdr@nrc.gov.

AmerGen Energy Company, LLC,. et al.,

Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of amendment request: April 6,

2001.

Description of amendment request:

The existing Oyster Creek Technical

Specification (TS) Section 4.7.B.5

requires capacity testing of the Station

Batteries and the Diesel Generator

Starting Batteries at least once per 24

months during a plant shutdown. The

proposed amendment request will allow

the 24-month capacity test for the Diesel

Generator Starting Batteries to be

performed during plant shutdowns or

during the 24-month on-line Diesel

Generator inspection (TS 4.7.A.3). The

proposed revision to Section 4.7.B.5.b

also reflects this change in specified

frequency.

Additionally, TS 4.7.A.5 is revised to

delete the statement that the battery

capacity test need not be performed if

the installed batteries were replaced

during the previous Diesel Generator

on-line biennial inspection. This

exception is no longer necessary

because the battery capacity testing is

not restricted to refueling outages based

on the proposed change to Section

4.7.B.5.

TS 4.7.B.5.a is revised to delete the

phrase ‘‘* * * to be considered

operable’’ because all of the specified

surveillances constitute operability

criteria. The title of Section 4.7.B is

revised to identify applicability to the

Diesel Generating Starting Batteries.

These additional proposed revisions are

considered administrative changes,

which clarify the existing TS.

TS 4.7 Bases is also revised to reflect

the above specification changes. Section

4.7 Bases contained on page 4.7–3 are

being relocated to Bases page 4.7–4.

This relocation of the Bases is a purely

administrative change.

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 change to allow the batteries to be

tested during the 24-month Diesel Generator

inspection outage does not increase the

probability of occurrence of an accident

previously evaluated. No change is being

made to equipment, equipment operation, or

equipment requirements. If a Diesel

Generator battery were to fail during the 24-

month inspection, the availability of the

Diesel Generator will not be affected because

the Diesel Generator will already be out of

service for the inspection. The change will

allow the Diesel Generator out of service time

during refueling outages to be reduced or

eliminated, thereby reducing risk.

The change to allow the batteries to be

tested during the 24-month Diesel Generator

inspection outage does not increase the

consequences of an accident previously

evaluated. No change is being made to

equipment, equipment operation, or

equipment requirements. If a Diesel

Generator battery were to fail during the 24-

month inspection, the consequences of the

battery failing are not affected.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The change to allow the batteries to be

tested during the 24-month Diesel Generator

inspection outage does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Moving the testing will not create a new

possible failure type, it will only move the

detection of a battery failure from the

refueling outage to the 24-month Diesel

Generator inspection outage.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The change to allow the batteries to be

tested during the 24-month Diesel Generator

inspection outage does not reduce a margin

of safety. Since the Diesel Generator will

already be out of service for the 24-month

inspection, the margin of safety for the Diesel

Generator 24-month inspection outage will

not be affected. The change will allow the

Diesel Generator out of service time during

the refueling outage to be reduced or

eliminated, thereby increasing the margin of

safety during the refueling outage.

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: Kevin P. Gallen,

Morgan, Lewis & Bockius, LLP, 1800 M

Street, NW., Washington, DC 20036–

5869.

NRC Section Chief: Richard P.

Correia, Acting.

AmerGen Energy Company, LLC,

Docket No. 50–289, Three Mile Island

Nuclear Station, Unit 1, Dauphin

County, Pennsylvania

Date of amendment request: January

23, 2001.

Description of amendment request:

The proposed amendment revises the

requirements for containment integrity

associated with the personnel and

emergency air locks and other

penetrations during fuel movement and

refueling operations to allow these

penetrations to remain open. One door

in each of the emergency and personnel

air locks must be capable of being

closed and each penetration providing

direct access from the containment

atmosphere to the outside atmosphere

shall be capable of being closed by an

isolation valve, blind flange, or manual

valve. The supporting revised design

basis fuel handling accident inside

containment analysis will also

incorporate alternative source term

methodology in accordance with Title

10 of the Code of Federal Regulations

(10 CFR) Section 50.67 and Regulatory

Guide 1.183, ‘‘Alternative Radiological

Source Terms For Evaluating Design

Basis Accidents At Nuclear Power

Reactors,’’ July 2000. Technical

Specification (TS) 3.8.7 is also revised

to provide equivalent isolation methods

for other penetrations consistent with

Babcock & Wilcox Owner’s Group

(BWOG) Standard Technical

Specifications (STSs), Section 3.9.3.c.1,

NUREG–1430, April 1995. TS 3.8.11 is

added to specify the requirement to

maintain at least 23 feet of water over

the top of the reactor vessel flange and

the actions required if this level is not

maintained. TS Bases 3.8 is revised to

provide a description of the plant

conditions under which the personnel

and emergency air locks and other

penetrations including those consistent

with the BWOG STSs, Section 3.9., may

be open during fuel movement, and the

administrative controls that would be in

place. The surveillance requirements of

TS 4.4.1.3 are also revised to identify

the exception allowed by TS 3.8.6 under

which both doors of the personnel and

emergency air locks can be open.

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

Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices 31703

consideration, which is presented

below:

1. Will the operation of the facility in

accordance with this proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Response: The proposed change would

allow the personnel and emergency air lock

doors and other penetrations to remain open

during fuel loading and refueling operations.

These penetrations were previously closed

during this time period in order to prevent

the escape of radioactive material in the

event of a fuel handling accident inside

containment (FHA). These penetrations are

not initiators of any accident. The probability

of a FHA is unaffected by the position of

these penetrations.

The new FHA analysis utilizing an

Alternative Source Term with an open

containment demonstrates that the maximum

doses are well within the acceptance criteria

specified in 10 CFR 50.67 and Regulatory

Guide 1.183. In the event of a fuel handling

accident, actual control room and offsite

doses will be less than analyzed values

because containment integrity will be

restored following an evacuation of

containment. As noted above, with the

Alternative Source Term implementation, the

acceptance criteria are also being revised. A

direct comparison of the new Alternative

Source Term dose consequences with the

existing licensing basis FHA source term

dose consequences is not practical due to the

significant differences in methodology and

assumptions.

However, a comparison of the previous

thyroid and whole body dose results for the

postulated TMI Unit 1 FHA Inside

Containment documented in the TMI Unit

1UFSAR [updated final safety analysis

report] Chapter 14 with the new dose results

expressed in terms of Total Effective Dose

Equivalent (TEDE), using the guidance in

Regulatory Guide 1.183 Footnote 7, indicates

that the new doses are not significantly

higher than the previous dose results. The

revised Alternative Source Term calculated

doses remain well within the allowable

acceptance criteria.

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 change does not

involve the addition or modification of any

plant equipment. Also, the proposed change

would not alter the design or method of

operation of the plant beyond the standard

functional capabilities of the equipment. The

proposed change involves a change to the

Technical Specifications that would allow

the personnel and emergency air lock doors

and other penetrations to be open during fuel

loading and refueling operations within the

containment. Having these doors and

penetrations open does not create the

possibility of a new accident. Administrative

provisions will be made to ensure the

capability to close the containment in the

event of a FHA inside containment.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Will operation of the facility in

accordance with this proposed change

involve a significant reduction in a margin of

safety?

Response: This proposed change has the

potential for an increased postulated accident

dose due to a FHA Inside Containment;

However, the analysis demonstrates that the

resultant doses are well within the

appropriate acceptance criteria. The margin

of safety, as defined by 10 CFR 50.67 and

Regulatory Guide 1.183, has been

maintained. The offsite and control room

doses due to a FHA with an open

containment have been evaluated with

conservative assumptions, which ensure the

calculation bounds the postulated accident

dose. Closing at least one door in each of the

personnel and emergency air locks following

the evacuation of the containment and

closure of other open penetrations would

reduce the control room and offsite doses in

the event of a FHA inside containment and

provides additional margin to the calculated

doses.

Therefore, the proposed change will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Edward J.

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

2301 Market Street, S23–1,

Philadelphia, PA 19103.

NRC Section Chief: Richard P.

Correia, Acting.

AmerGen Energy Company, LLC,

Docket No. 50–289, Three Mile Island

Nuclear Station, Unit 1, Dauphin

County, Pennsylvania

Date of amendment request: January

29, 2001.

Description of amendment request:

The proposed amendment revises the

Technical Specifications (TSs) to

remove the note from TS 4.5.4.1 that

restricts the applicability of the

specified engineered safeguards feature

(ESF) systems leakage rate limit of 15

gallons per hour to the current operating

Cycle 13 and establish this value as the

permanent TS limit. This limit had

previously been approved with the

issuance of Amendment No. 215 on

August 24, 1999, for Cycle 13 only. The

proposed amendment also would

implement a full scope alternative

source term for Three Mile Island

Nuclear Station, Unit 1, in accordance

with Title 10 of the Code of Federal

Regulations (10 CFR) Section 50.67 and

the guidance contained in Regulatory

Guide 1.183.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed permanent Technical

Specification limit on ESF Systems leak rate

is identical with the existing licensing basis

value and is conservatively reevaluated for

the limiting design basis Maximum

Hypothetical Accident (MHA) using

alternative source term methodology.

Implementation of the alternative source

term in accordance with Regulatory Guide

1.183 does not affect the design or operation

of the facility, and therefore, does not

significantly increase the probability of an

accident previously evaluated. Based on the

results of this reanalysis, it has been

demonstrated that with the requested

Technical Specification change, the offsite

and control room dose consequences for this

limiting event remain within the allowable

dose criteria specified in 10 CFR 50.67 and

Regulatory Guide 1.183.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed permanent Technical

Specification limit on ESF leak rate and

implementation of the alternative source

term in accordance with Regulatory Guide

1.183 does not affect the design, functional

performance, or operation of the facility or of

any equipment within the facility.

Modifications supporting the proposed

change have been evaluated and determined

not to create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed change involves

implementation of the alternative source

term in accordance with 10 CFR 50.67 and

Regulatory Guide 1.183, and maintains the

current Technical Specification limit on ESF

Systems leak rate. The reanalysis of the

limiting design basis MHA has been

performed using conservative methodologies

as specified in Regulatory Guide 1.183.

Margin has been maintained to ensure that

31704 Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices

the accident analysis dose consequences

bound the postulated event scenarios. The

calculated offsite and control room dose

consequences for this limiting event are

within the acceptance criteria as specified in

10 CFR 50.67 and Regulatory Guide 1.183.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Edward J.

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

2301 Market Street, S23–1,

Philadelphia, PA 19103.

NRC Section Chief: Richard P.

Correia, Acting.

Consolidated Edison Company of New

York, Docket No. 50–247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of amendment request: May 8,

2001.

Description of amendment request:

The proposed amendment would revise

the frequency of the Technical

Specification (TS) surveillance

requirement to check the movement of

the control rods. Specifically, the

frequency listed for this requirement in

TS Table 4.1–3, ‘‘Frequencies for

Equipment Tests,’’ would be changed

from ‘‘every 31 days’’ to ‘‘quarterly’’

during reactor critical operations.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability [...] or consequences of an

accident previously evaluated.

This change to the frequency of

performance of surveillance does not result

in any hardware changes or nor does it

change the response of control rods in

performing their specified function.

Therefore the change cannot affect the

probability of occurrence of previously

evaluated accidents.

The proposed frequency has been

determined to be adequate to assure the

reliability of reactor trip based on the

conclusions in NUREG 1366 [‘‘Improvements

to Technical Specification Surveillance

Requirements’’] and the recommendations of

GL [Generic Letter] 93–05 [‘‘Line-Item

Technical Specifications Improvements to

Reduce Surveillance Requirements for

Testing During Power Operation’’].

Therefore, operation of the facility in

accordance with the proposed amendment

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not introduce a

new failure mechanism or a new or different

type of accident than those previously

evaluated since there are no physical changes

being made to the facility. Performance of the

surveillance on the revised frequency will

not have an adverse affect on the ability of

the control rods to perform their intended

function. The proposed change does not

degrade the reliability of systems, structures,

or components or create a new accident

initiator or precursor. Therefore, the change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in [a] margin

of safety.

The proposed reduction in surveillance

testing reduces the risk for causing dropped

rods or reactor trips. This results in a slight

improvement in the margin of safety by

decreasing challenges to reactor components

and safety systems.

The proposed surveillance frequency, as

supported by the industry experience

described in NUREG–1366, continues to

provide the required assurance of control rod

operability, such that safety margins

established through the design and facility

license, including the Technical

Specifications, remain unchanged.

Therefore, operation of the facility in

accordance with the proposed amendment is

expected to result in a slight net

improvement in [a] margin of safety. Hence

the proposed change 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: Brent L.

Brandenburg, Esq., 4 Irving Place, New

York, New York 10003.

NRC Section Chief: Richard P.

Correia, Acting.

Consolidated Edison Company of New

York, Docket No. 50–247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of amendment request: May 10,

2001.

Description of amendment request:

The proposed amendment would

remove Technical Specification (TS)

surveillance requirement (SR) 4.6.A.4

that requires each emergency diesel

generator (EDG) to be given a thorough

inspection at least annually following

the manufacturer’s recommendations.

The requirement for the EDG inspection

will be relocated to the Updated Final

Safety Analysis Report and will be in

accordance with the licensee controlled

maintenance program. The inspection

period required by the maintenance

program will also be changed to specify

that it will be ‘‘in accordance with the

manufacturer’s recommendations.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability [...] or consequences of an

accident previously evaluated.

There is no change to the design, function,

or capability of the EDGs as a result of this

change. Hence there is no change in the

probability of occurrence of an accident

previously evaluated.

The change does not affect the ability of

the EDGs to mitigate the consequences of any

accident previously evaluated; including the

loss of coolant accident coupled with loss of

offsite power. To the contrary, this change is

structured to enhance the availability and

reliability of the EDGs by tailoring the actual

EDG maintenance program to the EDGs’

operational history and experience. In

addition, the surveillance testing

requirements of TS Surveillance

Requirements 4.6.A.1, 2 & 3 have not

changed and are adequate to verify the

operability of the EDG system. And, the

Maintenance Rule Program at IP2 [Indian

Point Unit 2] has established specific

performance criteria for the EDGs. These

performance criteria, and requirements to

ensure the criteria are met, are not affected

by this change.

The deletion of the surveillance

requirement and controlling EDG

maintenance using a licensee-controlled

maintenance program does not alter or

prevent the ability of the EDGs to perform

their intended functions.

Therefore, operation of the facility in

accordance with the proposed amendment

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

2. Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The EDG is not an accident initiator. The

proposed change does not involve any

physical design change or operational

change. Thus a new failure mode is not

introduced. In addition, the proposed change

has been evaluated to not degrade the

reliability of any existing system, structure,

or component. Therefore, the proposed

Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices 31705

change does not create a new accident

initiator or precursor, or create the possibility

of a new or different kind of accident from

any accident previously evaluated.

3. Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in [a] margin

of safety.

As a result of this change, there are no

changes to IP2’s design or to the IP2 TS safety

limits, limiting safety system settings, or

limiting conditions [for] operation. A single

SR is replaced by a performance-based

maintenance program.

The substitution of the performance-based

maintenance program for the prescriptive SR

is expected to increase the availability of the

EDGs because the amount of time the EDGs

are out-of-service for on-line maintenance

will decrease. Reducing the number of plant

operating hours that the unit is exposed to an

out-of-service EDG improves rather than

reduces the margin of safety. The substitution

of the performance-based maintenance

program for the prescriptive SR is expected

to improve the reliability of the EDGs by

minimizing the possibility of adverse results

that may result from intrusive maintenance

activities. The expected reliability

improvement improves rather than reduces

[a] margin of safety.

The transfer of control of EDG maintenance

from the TS to a licensee-controlled EDG

maintenance program is an administrative

change. But the change is structured so that

maintenance program changes must be

evaluated using the 10 CFR 50.59 process.

Use of the 10 CFR 50.59 process assures that

future changes to the EDG maintenance

program cannot significantly increase the

likelihood of a malfunction of the EDGs. And

use of the 10 CFR 50.59 process, instead of

the license amendment process, allows Con

Edison to optimize EDG maintenance in a

timely manner to meet the intent of 10 CFR

50.65.

The proposed changes do not adversely

affect the EDG’s ability to function when

required to mitigate any accident or licensing

basis event. Therefore, operation of the

facility in accordance with the proposed

amendment 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: Brent L.

Brandenburg, Esq., 4 Irving Place, New

York, New York 10003.

NRC Section Chief: Richard P.

Correia, Acting.

Dominion Nuclear Connecticut, Inc.,

Docket No. 50–336, Millstone Nuclear

Power Station, Unit No. 2, New London

County, Connecticut

Date of amendment request: April 11,

2001.

Description of amendment request:

The proposed amendment would revise

Technical Specification definitions 1.12,

‘‘Core Alteration;’’ 3.9.1, ‘‘Refueling

Operations—Boron Concentration;’’

3.9.2, ‘‘Refueling Operations—

Instrumentation;’’ and 3.9.11,

‘‘Refueling Operations—Water Level—

Reactor Vessel.’’ The Bases for these

Technical Specifications would also be

modified to reflect the proposed

changes to these definitions.

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 Technical Specification

changes associated with the definition for

Core Alteration and LCO [limiting condition

for operation], applicability, action

requirements and surveillance requirements

of Sections 3.9.1, 3.9.2 and 3.9.11 will not

cause an accident to occur and will not result

in any change in operation of the associated

accident mitigation equipment. The design

basis accidents (fuel handling and boron

dilution event) remain the same postulated

events described in the Millstone Unit No. 2

Final Safety

Analysis Report (FSAR). Therefore, the

proposed changes will not increase the

probability of an accident previously

evaluated.

The proposed LCO and Applicability

changes are consistent with the design basis

accident analyses of record. This will ensure

that the accident mitigation equipment

functions and associated equipment are

available for accident mitigation as assumed

in the associated accident analyses. The

proposed surveillance requirement changes

will continue to provide reasonable

assurance of equipment operability. As a

result, the accident assumptions and

mitigation methods will not be adversely

affected by the changes. Therefore, the

proposed changes will not result in [an]

increase in the consequences of accident[s]

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

The proposed changes to the Technical

Specifications do not impact any system or

component that could cause an accident. The

proposed changes will not alter the plant

configuration (no new or different type of

equipment will be installed) or require any

new or unusual operator actions. The

proposed changes will not alter the way any

structure, system, or component functions,

and will not significantly alter the manner in

which the plant is operated. There will be no

adverse effect on plant operation or accident

mitigation equipment. The response of the

plant and the operators following an accident

will not be different. In addition, the

proposed changes do not introduce any new

failure modes. Therefore, the proposed

changes will not create the possibility of a

new or different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed LCO and Applicability

changes are consistent with the design basis

accident analyses of record. The proposed

surveillance requirement changes will

continue to provide assurance of equipment

operability. The proposed changes do not

involve any changes in the accident analyses,

therefore, the proposed changes do not

involve a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Lillian M.

Cuoco, Senior Nuclear Counsel,

Dominion Nuclear Connecticut, Inc.,

Rope Ferry Road, Waterford, CT 06385.

NRC Section Chief: James W. Clifford.

Dominion Nuclear Connecticut, Inc.,

Docket No. 50–336, Millstone Nuclear

Power Station, Unit No. 2, New London

County, Connecticut

Date of amendment request: April 23,

2001.

Description of amendment request:

The proposed amendment would

remove the surveillance requirement to

perform inspections of the Emergency

Diesel Generators (EDGs) during

shutdown conditions from Technical

Specifications; although, inspections of

the EDGs would continue to be

performed in accordance with

procedures prepared in conjunction

with the recommendations of the

manufacturer.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s analysis which is based on

the representations made by the licensee

in the April 23, 2001 application, is

presented below:

1. Involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The Technical Specification change is

associated with the surveillance

requirement to perform inspections of

the EDGs during shutdown conditions.

The proposed change will remove this

surveillance requirement from

Technical Specifications; although,

inspections of the EDGs will continue to

31706 Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices

be performed in accordance with

procedures prepared in conjunction

with the recommendations of the

manufacturer.

Removal of the EDG inspection

surveillance requirement from

Technical Specifications does not verify

operability or EDG functions assumed in

the safety analysis. EDG inspections,

which are maintenance activities that

can be adequately controlled by plant

procedures, will still be performed in

accordance with the recommendations

of the manufacturer. This will provide

continued assurance the EDGs will be

available when required.

The proposed Technical Specification

change will have no adverse effect on

plant operation or the operation of

accident mitigation equipment, and will

not impact the availability of accident

mitigation equipment. The plant

response to the design basis accidents

will not change. In addition, the

equipment covered by this specification

change is not an accident initiator and

cannot cause an accident. Therefore, the

proposed Technical Specification

change will not result in an increase in

the probability or consequences of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed change does not impact

any system or component which could

cause an accident. The proposed change

will not alter the plant configuration (no

new or different type of equipment will

be installed) or require any unusual

operator actions. The proposed change

will not alter the way any structure,

system, or component functions, and

will not alter the manner in which the

plant is operated. There will be no

adverse effect on plant operation or

accident mitigation equipment. The

proposed change does not introduce any

new failure modes. Also, the response of

the plant and the operators following an

accident will not be different as a result

of this change. In addition, the accident

mitigation equipment affected by the

proposed change is not an accident

initiator. Therefore, the proposed

change will not create the possibility of

a new or different kind of accident from

any previously analyzed.

3. Involve a significant reduction in a

margin of safety.

The proposed change will have no

adverse effect on plant operation or

equipment important to safety. The

plant response to the design basis

accidents will not change and the

accident mitigation equipment will

continue to function as assumed in the

design basis accident analysis.

Therefore, there will be no reduction in

a margin of safety.

Based on the NRC staff’s analysis, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Lillian M.

Cuoco, Senior Nuclear Counsel,

Dominion Nuclear Connecticut, Inc.,

Rope Ferry Road, Waterford, CT 06385.

NRC Section Chief: James W. Clifford.

Entergy Operations, Inc., Docket No.

50–368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: May 2,

2001.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TSs) to not

require the moderator temperature

coefficient (MTC) determination in TS

4.1.1.4.2.c if the results of the MTC

determinations required in TSs

4.1.1.4.2.a and 4.1.1.4.2.b are within a

certain tolerance of the corresponding

design values.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Will operation of the facility in

accordance with this proposed change

involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Under the proposed change, compliance

with the TS[s] is maintained by measuring

the beginning[-]of[-]cycle [(BOC)]

temperature coefficients.

This change does not require a

modification to any of the assumptions used

in the input to the safety analyses. The

assumptions were based on the current range

of MTC allowed by TSs. The proposed

change does not include a revision to the TS

allowed range of MTC.

Therefore, this change does not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

2. Will operation of the facility in

accordance with this proposed change create

the possibility of a new or different kind of

accident from any accident previously

evaluated?

This change does not result in changing

plant operation or any TS limits. The MTC

will continue to be acceptably verified within

specified limits. As described in the

Combustion Engineering topical report, if the

BOC MTC measurements are within the

specified tolerance when compared to the

design value, then the EOC [end-of-cycle]

value is expected to fall within the design

margin.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. Will operation of the facility in

accordance with this proposed change

involve a significant reduction in a margin of

safety?

This change does not modify the range of

allowed temperature coefficients. The

surveillance program consisting of BOC

measurements, of plant parameter

monitoring, and of explicit EOC predictions

will ensure that the MTC remains within the

range of acceptable values.

Therefore, this change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC,

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois Docket Nos. STN

50–456 and STN 50–457, Braidwood

Station, Unit Nos. 1 and 2, Will County,

Illinois

Date of amendment request: April 27,

2001.

Description of amendment request:

The proposed amendment would revise

TS 5.5.7, ‘‘Reactor Coolant Pump

Flywheel Inspection Program,’’ which

requires the inspection of each reactor

coolant pump (RCP) flywheel in general

conformance with the recommendations

of Regulatory Position C.4.b of NRC

Regulatory Guide (RG) 1.14, Revision 1,

‘‘Reactor Coolant Pump Flywheel

Integrity,’’ dated August 1975. The

proposed change revises TS 5.5.7 to

provide an exception to the

recommendations of Regulatory Position

C.4.b which would allow either a

qualified in-place ultrasonic volumetric

examination (UT) over the volume from

the inner bore of the flywheel to the

circle of one-half the outer radius or a

surface examination (i.e., magnetic

particle testing (MT) and/or liquid

penetrant testing (PT)) of exposed

surfaces of the removed flywheel to be

conducted at approximately 10-year

intervals. The proposed change is in

accordance with the NRC approved

Improved Standard TS Generic Change

Traveler TSTF–237, Revision 1.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices 31707

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?

An integral part of the Reactor Coolant

System (RCS) in a Pressurized Water Reactor

(PWR) is the Reactor Coolant Pump (RCP).

The RCP ensures an adequate cooling flow

rate by circulating large volumes of the

primary coolant water at high temperature

and pressure through the RCS. Following an

assumed loss of power to the RCP motor, the

flywheel, in conjunction with the impeller

and motor assembly, provide sufficient

rotational inertia to assure adequate core

cooling flow during RCP coastdown.

Westinghouse Electric Corporation Topical

Report WCAP–14535A, ‘‘Topical Report on

Reactor Coolant Pump Flywheel Inspection

Elimination,’’ dated November 1996,

provides the technical basis for the

elimination of inspection requirements for

RCP flywheels for all domestic Westinghouse

plants. In the Safety Evaluation for WCAP–

14535A, dated September 1996, the NRC

stated that the evaluation methodology

described in WCAP–14535A is appropriate

and the criteria are in accordance with the

design criteria of RG 1.14. RCP flywheel

inspections have been performed for 20 years

with no indications of service induced flaws.

Flywheel integrity evaluations show a very

high flaw tolerance for the RCP flywheels.

Crack extension over a 60-year service life is

negligible. Structural reliability studies have

shown that eliminating inspections after 10

years of plant life will not significantly

change the probability of failure.

The proposed change does not adversely

affect accident initiators or precursors nor

alter the design assumptions, conditions, and

configuration of the facility or the manner in

which the plant is operated and maintained.

The proposed change does not alter or

prevent the ability of structures, systems, and

components (SSC) from performing their

intended function to mitigate the

consequences of an initiating event within

the acceptance limits assumed in the

Braidwood and Byron Stations’ Updated

Final Safety Analysis Report (UFSAR). The

proposed changes do not affect the source

term, containment isolation, or radiological

release assumptions used in evaluating the

radiological consequences of an accident

previously evaluated in the Braidwood and

Byron Stations’ UFSAR.

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 accident

from any accident previously evaluated?

The proposed change does not modify the

design or function of the RCP flywheels.

Based upon the results of WCAP–14535A, no

new failure mechanisms will be introduced

by the revised RCP Flywheel Inspection

Program. As presented in WCAP–14535A,

detailed stress analysis and risk assessments

have been performed that indicate that there

would be no change in the probability of

failure for RCP flywheels if all inspections

were eliminated. Flywheel integrity

evaluations show that RCP flywheels exhibit

a very high tolerance for the presence of

flaws.

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?

There is no significant mechanism for inservice degradation of the flywheels since

they are isolated from the primary coolant

environment. Additionally WCAP–14535A

analyses have shown there is no significant

deformation of the flywheels even at

maximum overspeed conditions. Likewise,

the results of RCP flywheel inspections

performed throughout the industry and at the

Braidwood Station and the Byron Station

identified no indications that would affect

flywheel integrity.

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

requested amendments involve no

significant hazards consideration.

Attorney for licensee: Mr. Edward J.

Cullen, Vice President, General Counsel,

Exelon Generation Company, LLC, 300

Exelon Way, Kennett Square, PA 19348.

NRC Section Chief: Anthony J.

Mendiola.

Exelon Generation Company, LLC,

Docket Nos. 50–373 and 50–374,

LaSalle County Station, Units 1 and 2,

LaSalle County, Illinois

Date of amendment request: April 16,

2001.

Description of amendment request:

The proposed amendments would

change the reference in Technical

Specification 5.5.6, ‘‘Pre-Stressed

Concrete Containment Tendon

Surveillance Program,’’ from Regulatory

Guide 1.35, ‘‘Inservice Inspection of

Ungrouted Tendons in Prestressed

Concrete Containments,’’ Revision 3,

1989, to a reference to Subsection IWL,

‘‘Requirements of Class CC Concrete

Components of Light-Water Cooled

Power Plants,’’ of Section XI, ‘‘Inservice

Inspection,’’ of the American Society of

Mechanical Engineers (ASME) Boiler

and Pressure Vessel Code, and to delete

the applicability of Surveillance

Requirement (SR) 3.0.2 to TS Section

5.5.6. SR 3.0.2 allows the surveillance to

be performed within 1.25 times the

interval specified in the surveillance’s

frequency.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed changes to Technical

Specifications (TS) Section 5.5.6, ‘‘PreStressed Concrete Containment Tendon

Surveillance Program,’’ change the reference

in TS Section 5.5.6 from Regulatory Guide

(RG) 1.35, ‘‘Inservice Inspection of Ungrouted

Tendons in Prestressed Concrete

Containments,’’ Revision 3, 1989, to a

reference to Subsection IWL, ‘‘Requirements

of Class CC Concrete Components of LightWater Cooled Power Plants,’’ of Section XI,

‘‘Inservice Inspection,’’ of the American

Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel (B&PV) Code, and

to delete the applicability of Surveillance

Requirement (SR) 3.0.2 to TS Section 5.5.6.

SR 3.0.2 allows the surveillance to be

performed within 1.25 times the interval

specified in the surveillance’s frequency. The

proposed changes do not significantly effect

the Tendon Surveillance Program, inspection

frequencies, and acceptance criteria which

provide the requirements for the performance

of the primary containment tendon

inspections at LaSalle County Station, Unit 1

and Unit 2.

The performance of a primary containment

tendon inspection is not a precursor to any

accident previously evaluated. Thus, the

proposed changes to the performance of a

primary containment tendon inspection do

not have any effect on the probability of an

accident previously evaluated.

The performance of primary containment

tendon inspections does provide assurance

that the primary containment will perform as

designed. Thus, the radiological

consequences of any accident previously

evaluated are not increased.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Does the change create the possibility of a

new or different kind of an accident from any

accident previously evaluated?

The proposed changes to TS Section 5.5.6,

provide assurance that the primary

containment will perform as designed and do

not introduce any new modes of primary

containment operation of failure

mechanisms.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any previously

evaluated.

Does the change involve a significant

reduction in a margin of safety?

On August 8, 1996, the NRC published a

final rule in the Federal Register (i.e., 61

Federal Register 41303) to amend 10 CFR

50.55a, ‘‘Codes and standards,’’ to

incorporate by reference Subsection IWL of

Section XI, of the ASME B&PV Code.

Subsection IWL of Section XI, of the ASME

31708 Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices

B&PV Code, provides rules for the inservice

inspection and repair of the reinforced

concrete and post tensioning systems of Class

CC components. LaSalle County Station, Unit

1 and Unit 2, primary containments are Class

CC components. The amended 10 CFR 50.55a

required incorporation of Subsection IWL of

Section XI, of the ASME B&PV Code, into

inspection programs by September 9, 2001.

We have developed an inspection program to

implement Subsection IWL of Section XI, of

the ASME B&PV Code. The proposed TS

changes support this program.

The revised Tendon Surveillance Program,

inspection frequencies, and acceptance

criteria developed to implement Subsection

IWL of Section XI, of the ASME B&PV Code,

as required by 10 CFR 50.55a, provide

acceptable requirements to perform

inspections of the tendons in the LaSalle

County Station, Unit 1 and Unit 2, primary

containments. Thus, the proposed change to

TS Section 5.5.6 will continue to ensure the

integrity of the Unit 1 and Unit 2 primary

containment tendons as required by the

current TS.

Thus, 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 19 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. Edward J.

Cullen, Vice President, General Counsel,

Exelon Generation Company, LLC, 300

Exelon Way, Kennett Square, PA 19348.

NRC Section Chief: Anthony J.

Mendiola.

Exelon Generation Company, LLC,

PSEG Nuclear LLC, and Atlantic City

Electric Company, Dockets Nos. 50–277

and 50–278, Peach Bottom Atomic

Power Station, Units Nos. 2 and 3, York

County, Pennsylvania

Date of application for Amendments:

April 3, 2001.

Description of amendment request:

The proposed amendment would revise

the Peach Bottom Atomic Power Station

(PBAPS), Units 2 and 3, technical

specifications (TSs) in accordance with

Technical Specification Task Force

(TSTF) item TSTF–258, Revision 4. This

TSTF has been previously reviewed and

approved by the NRC as generically

applicable to nuclear plants with

improved standard TSs, such as PBAPS.

The proposed amendment revises TS

Section 5.0, ‘‘Administrative Controls,’’

to delete details of staffing

requirements, eliminate specific details

for working hour limits, clarify

requirements for the Shift Technical

Advisor position, add regulatory

definitions for Senior Reactor Operators

and Reactor Operators, revise the

Radioactive Effluents Control Program

to be consistent with the intent of Title

10 of the Code of Federal Regulations

(10 CFR) Part 20, delete periodic

reporting requirements for main stream

relief valve openings, and revise

radiological control requirements for

radiation areas to be consistent with

those specified in 10 CFR 20.1601(c).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed TS changes do not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed TS changes are

administrative in nature and do not impact

the operation, physical configuration, or

function of plant equipment or systems. The

changes do not impact the initiators or

assumptions of analyzed events, nor do they

impact mitigation of accidents or transient

events. Therefore, these proposed changes do

not increase the probability of occurrence or

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. The proposed TS changes are

administrative in nature and do not alter

plant configuration, require that new

equipment be installed, alter assumptions

made about accidents previously evaluated,

or impact the operation or function of plant

equipment. The proposed changes do not

introduce any new modes of plant operation

or make any changes to system setpoints.

Therefore, these proposed changes do not

create the possibility of a new or different

kind of accident than previously evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The proposed TS changes are

administrative in nature and do not involve

physical changes to plant structures, systems,

or components (SSCs), or the manner in

which these SSCs are operated, maintained,

modified, tested, or inspected. The proposed

changes do not involve a change to any safety

limits, limiting safety system settings,

limiting conditions for operation, or design

parameters for any SSC. The proposed

changes do not impact any safety analysis

assumptions and do not involve a change in

initial conditions, system response times, or

other parameters affecting any accident

analysis. Therefore, these changes do no

involve any 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. Edward

Cullen, Vice President and General

Counsel, Exelon Generation Company,

LLC, 300 Exelon Way, Kennett Square,

PA 19348.

NRC Section Chief: James W. Clifford.

FirstEnergy Nuclear Operating

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

Ottawa County, Ohio

Date of amendment request: April 4,

2001.

Description of amendment request:

The proposed amendment would delete

Technical Specification (TS) Section

1.7, Definitions—Reportable Events, and

TS 6.6, Reportable Event Action, from

the Davis-Besse Nuclear Power Station

Operating License, and revise TS 6.5.3,

Technical Review and Control—

Activities, and TS Bases 4.0.3,

Applicability. These changes are being

proposed to delete TS requirements

already required by Title 10 of the Code

of Federal Regulations Part 50 (10 CFR

50), update the TS Bases to reflect

recent changes made to 10 CFR 50.73,

revise the approval authorizations for

procedures, plant modifications, tests

and experiments, and reflect recent

changes made 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 Davis-Besse Nuclear Power Station has

reviewed the proposed changes and

determined that a significant hazards

consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station, Unit No. 1, in accordance with these

changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no accident initiators,

conditions or assumptions are affected by the

proposed changes to delete Technical

Specification (TS) 1.7, Definitions—

Reportable Event, and TS 6.6, Reportable

Event Action, from the Davis-Besse Nuclear

Power Station (DBNPS) Operating License;

and revise TS Bases 4.0.3, Applicability.

Reportable Events are addressed by 10 CFR

50.73 and it is not necessary for the TS to

include items already required by federal

regulation. The proposed changes to TS

Bases 4.0.3 would make these Bases

consistent with the recent revision to 10 CFR

50.73. The proposed changes to the TS Index

reflect the deletion of TS 1.7 and TS 6.6,

Reportable Event Action, and are

administrative changes.

The proposed changes to TS 6.5.3,

Technical Review and Control—Activities,

provide for the approval of activities affecting

nuclear safety by personnel authorized by

procedure. These changes continue to

implement the DBNPS Quality Assurance

Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices 31709

Program commitments. Qualification

requirements for individuals performing

reviews of activities affecting nuclear safety

are not affected. Accordingly, there is no

increase in the probability of an accident.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because no accident conditions or

assumptions are affected by the proposed

changes. The proposed changes do not alter

the source term, containment isolation, or

allowable releases. The proposed changes,

therefore, will not increase the radiological

consequences of a previously evaluated

accident.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because no new

accident initiators or assumptions are

introduced by the proposed changes. The

proposed changes do not alter any existing

accident scenarios, or involve a modification

or change in operation of any plant systems,

structures, or components.

3. Not involve a significant reduction in a

margin of safety because the proposed

changes are administrative in-nature and do

not reduce or adversely affect the capabilities

of any plant structures, systems or

components to perform their nuclear safety

functions.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Mary E.

O’Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street,

Akron, OH 44308.

NRC Section Chief: Anthony J.

Mendiola.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Units 1 and 2,

Berrien County, Michigan

Date of amendment requests: May 15,

2001.

Description of amendment requests:

The proposed amendments would

replace the current Technical

Specification (TS) requirement to

establish containment integrity within 8

hours if less than the specified

minimum complement of A.C. or D.C.

busses and equipment is operable in

Modes 5 and 6. The proposed TS would

require immediate suspension of

operations involving core alterations,

positive reactivity changes, and

movement of irradiated fuel assemblies,

and immediately initiate actions to

restore the required busses and

equipment to operable status, and to

immediately declare the associated

required residual heat removal loop(s)

inoperable. The current Action

requirement presents a scheduling and

administrative burden during outages

and extended shutdowns. In the

addition, the proposed amendment

would add options to the TS to allow

containment penetration closure

methods that are equivalent to those

that are currently required during core

alterations or movement of irradiated

fuel in containment, and allow

unisolation of some penetrations under

administrative control. The additional

options will allow flexibility in

scheduling outage activities.

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 determination that the criteria set forth

in 10 CFR 50.92 are met for this amendment

request is indicated below.

1. Does the change involve a significant

increase in the probability of occurrence or

consequences of an accident previously

evaluated?

Probability of Occurrence of an Accident

Previously Evaluated

The proposed changes to Action statements

for T/S 3/4.8.2.2 and T/S 3/4.8.2.4 will

eliminate current compensatory requirements

that can only mitigate the consequences of

accidents. The current requirements will be

replaced with requirements that include

measures to reduce the likelihood of

accidents and assist in responding to

malfunctions. The proposed requirements to

immediately suspend operations involving

core alterations, positive reactivity changes,

and movement of irradiated fuel assemblies

provide assurance that the applicable

accidents, fuel handling and shutdown

dilution accidents, will not occur by

requiring cessation of activities that may

cause them. The proposed requirements to

immediately initiate actions to restore the

required busses and equipment to operable

status and to immediately declare associated

required RHR loop(s) inoperable provide

assurance operators can take timely

corrective action for malfunctions that may

lead to a dilution accident, and will take

appropriate corrective actions for RHR

malfunctions. Therefore, there is no adverse

effect on accident initiators or precursors.

The proposed change to the Applicability

requirements for T/S 3/4.8.2.2 and T/S

3/4.8.2.4 expands the conditions under

which the T/S are invoked. The proposed

change will assure that the electrical power

is available for mitigation of a fuel handling

accident, regardless of the operational mode

of the plant. The proposed change only

involves accident mitigation capabilities and

does not affect any accident initiators or

precursors.

The proposed changes to the LCO for T/S

3/4.9.4 will provide additional options for

assuring closure of containment penetrations

during core alterations or movement of

irradiated fuel in containment. Containment

closure provides only mitigation for the

consequences of a fuel handling accident and

does not affect the initiators or precursors of

the accident.

The proposed change to the Surveillance

requirements for T/S 3/4.9.4 allows the LCO

to define the penetration status that is to be

periodically verified. The effect of the

proposed Surveillance change is bounded by

the effect of the proposed LCO change as

described above. Therefore, the proposed

Surveillance change does not adversely affect

any accident initiators or precursors.

Consequences of an Accident Previously

Evaluated

The proposed changes to the Action

requirements for T/S 3/4.8.2.2 and T/S

3/4.8.2.4 provide assurance that fuel

handling and dilution accidents will not

occur and that timely and appropriate

responses can and will be taken for

malfunctions, thereby reducing the

likelihood that radioactive material will be

released.

The proposed change to the Applicability

requirements for Unit 1 T/S 3/4.8.2.2 and T/

S 3/4.8.2.4 provides assurance that electrical

power is available for mitigation of a fuel

handling accident (FHA), regardless of the

operational mode of the plant. Since the

current Applicability requirement only

provides this assurance in Modes 5 and 6, the

proposed change will not increase the

consequences of the accident.

The additional options provided by the

proposed changes to the LCO for T/S 3/4.9.4

will mitigate the consequences of a fuel

handling accident in containment as

effectively as those specified by the current

LCO. Additionally, the consequences of a

FHA in containment determined by the

accident analyses will not increase since the

analyses do not credit mitigation by closure

of containment penetrations.

The proposed change to the Surveillance

requirements for T/S 3/4.9.4 only reflects the

change proposed for the LCO. The effect of

the proposed Surveillance change is bounded

by the effect of the proposed LCO change as

described above. Therefore, the proposed

Surveillance change does not adversely affect

the consequences of an accident.

The proposed changes to the Bases for the

above identified T/S only provide

explanatory information regarding the intent

of the specifications and how they are to be

implemented. The proposed Bases changes

do not alter requirements of the associated T/

S. Therefore, the effect of the Bases changes

on accident initiators and precursors and on

the consequences of an accident is bounded

by the effect of the associated Action or LCO

change as described above. The format

changes do not alter any requirements.

Therefore, the probability of occurrence or

the consequences of accidents previously

evaluated are not increased.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed changes to Action statements

for T/S 3/4.8.2.2 and T/S 3/4.8.2.4 to

eliminate requirements to establish

containment integrity does not affect

existing, or create new, accident initiators or

precursors because only existing passive

31710 Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices

accident mitigation features are involved.

Implementation of the proposed new

requirements to suspend operations

involving core alterations, positive reactivity

changes, and movement of irradiated fuel

assemblies does not affect existing, or create

new, accident initiators or precursors

because these activities do not require the

operation of existing equipment in a new or

different manner, or involve the operation of

new or different equipment. Implementation

of the proposed new requirements to initiate

actions to restore the required busses and

equipment to operable status and to declare

associated required RHR loop(s) inoperable

does not affect or create new accident

initiators or precursors because these

activities are currently required by existing

procedures and other T/S.

The proposed change to the Applicability

requirements for T/S 3/4.8.2.2 and T/S 3/

4.8.2.4 does not affect or create new accident

initiators or precursors because it only

expands the conditions under which the T/

S are invoked.

The proposed changes to the LCO for T/S

3/4.9.4 to provide additional options for

assuring closure of containment penetrations

during core alterations or movement of

irradiated fuel in containment does not affect

or create new accident initiators or

precursors because the changes involve only

containment penetrations which are passive

accident mitigation measures.

The proposed change to the Surveillance

requirements for T/S 3/4.9.4 allows the LCO

to define the penetration status that is to be

periodically verified. The effect of the

proposed Surveillance change is bounded by

the effect of the proposed LCO change as

described above. Therefore, the proposed

Surveillance change does not affect or create

new accident initiators or precursors.

The proposed changes to the Bases for the

above identified T/S only provide

explanatory information regarding the intent

of the specifications and how they are to be

implemented. The proposed Bases changes

do not alter requirements of the associated T/

S. Therefore, the effect of the Bases changes

on accident initiators or precursors is

bounded by the effect of the associated

Action or LCO change as described above.

The format changes do not alter any

requirements.

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 change involve a significant

reduction in a margin of safety?

The margins of safety pertinent to the

proposed changes to Action statements for T/

S 3/4.8.2.2 and T/S 3/4.8.2.4 are those

associated with a FHA, a shutdown dilution

event, and a RHR system malfunction. The

applicable margin of safety for a FHA is that

defined by the off site dose analyses for the

accident. Since the analyses do not credit

mitigation by the containment, the margin of

safety is unaffected. The applicable margin of

safety for a shutdown dilution event is the

time available for operators to take action to

preclude violating shutdown margin

requirements. The proposed new Action

requirements to immediately suspend

operations involving positive reactivity

changes, and to immediately initiate actions

to restore the required electrical busses and

equipment to operable status, would not

decrease the margin of safety for a shutdown

dilution event. The applicable margin of

safety for a RHR system malfunction is the

time available for operators to take action to

restore decay heat removal capabilities. The

proposed new actions requirements to

immediately initiate actions to restore the

required electrical busses and equipment to

operable status and to immediately declare

associated required RHR loop(s) inoperable

would not decrease the margin of safety for

a RHR system malfunction.

The margin of safety pertinent to the

proposed changes to LCO for T/S 3/4.9.4 is

that associated with a FHA. The applicable

margin of safety for a FHA is that defined by

the off site dose analyses for the accident.

Since the analyses do not credit mitigation by

the containment, the margin of safety is

unaffected.

There is no margin of safety pertinent to

the proposed changes to associated

Applicability requirements, Surveillance

requirements, and Bases for the above

identified T/S. The format changes do not

alter any requirements.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

In summary, based upon the above

evaluation, [Indiana Michigan Power

Company (I&M)] has concluded that the

proposed amendment involves no 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

amendment requests involve no

significant hazards consideration.

Attorney for licensee: David W.

Jenkins, Esq., 500 Circle Drive,

Buchanan, MI 49107.

NRC Section Chief: Claudia M. Craig.

Nebraska Public Power District, Docket

No. 50–298, Cooper Nuclear Station,

Nemaha County, Nebraska

Date of amendment request: February

28, 2001.

Description of amendment request:

The proposed amendment would

change the Technical Specification (TS)

to incorporate laboratory testing

recommendations of Generic Letter 99–

02, ‘‘Laboratory Testing of NuclearGrade Activated Charcoal,’’ June 3,

1999.

The proposed charcoal testing

changes and explicit reference to

American Society for Testing and

Materials (ASTM) D3803–1989 nucleargrade activated charcoal test protocol do

not affect engineered safety feature

(ESF) ventilation system operation or

performance, reliability, actuation

setpoints, or accident mitigation

capabilities. The proposed changes also

do not affect the operation and

performance of any other equipment

important to safety at Cooper Nuclear

Station (CNS).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed charcoal testing changes and

explicit reference to ASTM D3803–1989

nuclear-grade activated charcoal test protocol

do not affect ESF ventilation system

operation or performance, reliability,

actuation setpoints, or accident mitigation

capabilities. The proposed changes also do

not affect the operation and performance of

any other equipment important to safety at

CNS. ASTM D3803–1989 is a more accurate

and demanding test which ensures that the

charcoal filter efficiencies assumed in the

CNS accident dose analysis are maintained.

The proposed changes involve ESF

ventilation system charcoal testing only and

do not affect accident initiators. Therefore

the proposed changes do not significantly

increase the probability or consequences of

an accident previously evaluated in the

Updated Safety Analysis Report (USAR), as

revised by the Design Basis Accident (DBA)

radiological assessment calculational

methodology revisions submitted to the U. S.

Nuclear Regulatory Commission (NRC) under

Reference 2.

2. Does not create the possibility for a new

or different kind of accident from any

accident previously evaluated.

The charcoal testing changes, and explicit

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

affect ESF ventilation system operation or

performance, or the operation and

performance of any other equipment

important to safety at CNS. The proposed

changes clarify and explicitly identify the

testing of the ESF ventilation system charcoal

samples. No new or different accident

scenarios, transient precursors, failure

mechanisms, plant operating modes, or

limiting single failures are introduced as a

result of these changes. Therefore, the

possibility of a new or different kind of

accident from that previously evaluated in

the USAR, as revised by the DBA radiological

assessment calculational methodology

revision submitted to the NRC under

Reference 2, is not created by this change.

3. Does not create a significant reduction

in the margin of safety.

The required performance of the ESF

ventilation systems following a DBA is not

impacted by utilizing a more demanding

protocol for charcoal testing. Thus, the

margin of safety assumed in the CNS

accident analysis, as revised by the DBA

radiological assessment calculational

methodology revision submitted to the NRC

Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices 31711

under Reference 2, is maintained. Revising

the TS to clarify charcoal testing

methodology and explicitly referencing the

charcoal absorber testing being performed

does not affect ESF ventilation system

performance or operation, or the operation

and performance of any other equipment

important to safety at CNS. Therefore, these

changes do not result in a significant

reduction in 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: Mr. John R.

McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus,

NE 68602–0499.

NRC Section Chief: Robert A. Gramm.

Nuclear Management Company, LLC,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: May 25,

2001.

Description of amendment request:

The proposed amendment would

change the Kewaunee Nuclear Power

Plant Technical Specification 4.2 to

remove the steam generator tube

alternate repair criteria, because these

alternate repair criteria, as approved, are

not compatible with the replacement

steam generators scheduled to be

installed in the fall of 2001. In addition,

the proposed amendment would make

administrative changes revising the

phrasing of text without altering

technical content.

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.

Changing the technical specification

within limits of the bounding accident

analyses cannot change the probability of an

accident previously evaluated or the

currently licensed radiological consequence

predicted by the analyses of record. Removal

of an allowance for alternate repair criteria

defaults to the more conservative repair

criteria of plugging degraded tubes. Thus,

nothing in this proposal will cause an

increase in the probability or consequence of

an accident previously evaluated.

(2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Removal of alternate repair criteria from

[Technical Specification] TS leaves in its

place the more conservative, more restrictive

criteria for plugging degraded steam

generator tubes. Plugging degraded steam

generator tubes is a currently licensed repair

methodology for [Kewaunee Nuclear Power

Plant] KNPP, is consistent with current plant

design bases, and does not adversely affect

any fission product barrier, nor does it alter

the safety function of safety significant

systems, structures and components or their

roles in accident prevention or mitigation.

Currently, licensed design basis accident and

transient analyses of record bound the effect

of plugging tubes. Thus, this proposal does

not create the possibility of a new or different

kind of accident.

(3) Involve a significant reduction in the

margin of safety.

The proposed change does not alter the

manner in which Safety Limits, Limiting

Safety System Setpoints, or Limiting

Conditions for Operation are determined. It

places TS 4.2 in a more conservative

configuration than that previously approved

for use by the [Nuclear Regulatory

Commission] NRC. It conforms to plant

design bases, is consistent with current safety

analyses, and limits actual plant operation

within analyzed and licensed boundaries.

Removal of reference to use of alternate

repair criteria from TS 4.2 and its Bases

leaves existing and more conservative criteria

in place. Thus, changes proposed by this

request do not involve a significant reduction

in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Bradley D.

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

Box 1497, Madison, WI 53701–1497.

NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC,

Docket No. 50–263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of amendment request: May 18,

2001.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications to delete a

redundant requirement for valving out a

control rod drive, revise control rod

accumulator operability requirements,

add the option to hydraulically isolate

control rod drives, and correct an

inconsistency in core monitoring

describing when source range monitors

are required to be operable.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Deleting the paragraph which specifies one

specific pattern of control rod inoperability

does not degrade the safe operation of the

plant as inoperable control rods must still be

analyzed to meet shutdown requirements.

Revising the operability requirements for

control rod accumulators from ‘‘a nine-rod

square array’’ to: ‘‘provided that no other

control rod within two control rod cells in

any direction has a:’’ is a clarification. No

technical requirements are changed,

therefore, the probability or consequences of

previous evaluations of accidents have not

been affected. This change will assure

conformance with the Banked Position

Withdrawal Sequence (BPWS) analysis

documented in General Electric (GE) report

NEDO–21231. No changes in plant

equipment will occur.

The proposed change adds the option to

hydraulically isolate the drive to prevent

inadvertent drive withdrawal and not

consider the accumulator inoperable. This

provides a method of isolating a control rod

drive with an inoperable accumulator in

addition to electrical isolation when the

control rod is fully inserted. A statement on

when an inoperable accumulator is allowed

is being relocated so that it also applies

during refueling. Since in refueling, the plant

is already shutdown, the accumulators are

not required. As such, this change does not

increase the probability or consequences of

an accident previously evaluated.

A qualifier is being added that source range

monitors (SRMs) only need to be functionally

tested when there are more than two fuel

assemblies present in any reactor quadrant.

Criticality is not considered possible with

two or less fuel bundles in each quadrant and

adjacent to an SRM. Since this change will

only allow bypassing SRM functional checks

when two fuel bundles or less are present in

each quadrant, this change cannot result in

an inadvertent criticality. This proposed

change would reduce surveillance testing to

that time when the instrument is required to

be operable and provide consistency between

specifications.

The proposed Technical Specification

changes do not introduce new equipment or

new equipment operating modes, nor do the

proposed changes alter existing system

relationships. The proposed amendment does

not introduce new failure modes. Based on

the above justification, the proposed

amendment will have no impact on the

probability or consequences of an accident.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously analyzed.

This change does not degrade the safe

operation of the plant as inoperable control

rods must still be analyzed to meet existing

shutdown reactivity requirements. It will

assure conformance with the Banked Position

Withdrawal Sequence analysis documented

in General Electric report NEDO–21231. No

changes in plant equipment will occur.

Adding hydraulic isolation will not create

the possibility of a new or different kind of

31712 Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices

accident from any accident previously

analyzed.

Since this change will only allow

bypassing SRM functional checks with two

fuel bundles or less present in each quadrant,

this change cannot result in an inadvertent

criticality.

The proposed Technical Specification

changes do not introduce new equipment or

new equipment operating modes, nor do the

proposed changes alter existing system

relationships. The proposed amendment does

not introduce new failure modes. Based on

the above justification, the proposed

amendment will not create the possibility of

a new or different kind of accident from any

accident previously analyzed.

3. The proposed amendment will not

involve a significant reduction in the margin

of safety.

Revising the control rod operability

requirement does not degrade the safe

operation of the plant.

Hydraulic isolation provides a method of

isolating the drive in addition to the current

electrical isolation. Both methods disarm the

control rod drive and preclude the possibility

of inadvertent drive withdrawal during

subsequent operations. Adding applicability

during refueling has little impact on safety as

the drive is required to be fully inserted prior

to isolation. As such, they do not involve a

significant reduction in the margin of safety.

Since this change will only allow

bypassing SRM functional checks with two

fuel bundles or less present in each quadrant,

this change cannot result in an inadvertent

criticality.

Based on the above justification, the

proposed Technical Specification change

does not involve a significant reduction in

the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Jay E. Silberg,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW,

Washington, DC 20037.

NRC Section Chief: Claudia M. Craig.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of amendment requests: May 4,

2001.

Description of amendment requests:

The proposed amendments delete

requirements from the Technical

Specifications (and, as applicable, other

elements of the licensing bases) to

maintain a Post Accident Sampling

System (PASS). Licensees were

generally required to implement PASS

upgrades as described in NUREG–0737,

‘‘Clarification of TMI [Three Mile

Island] Action Plan Requirements,’’ and

Regulatory Guide 1.97,

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

Plant and Environs Conditions During

and Following an Accident.’’

Implementation of these upgrades was

an outcome of the lessons learned from

the accident that occurred at TMI, Unit

2. Requirements related to PASS were

imposed by Order for many facilities

and were added to or included in the

technical specifications (TS) for nuclear

power reactors currently licensed to

operate. Lessons learned and

improvements implemented over the

last 20 years have shown that the

information obtained from PASS can be

readily obtained through other means or

is of little use in the assessment and

mitigation of accident conditions.

The NRC staff issued a notice of

opportunity for comment in the Federal

Register on August 11, 2000 (65 FR

49271) on possible amendments to

eliminate PASS, including a model

safety evaluation and model no

significant hazards consideration

(NSHC) determination, using the

consolidated line item improvement

process. The NRC staff subsequently

issued a notice of availability of the

models for referencing in license

amendment applications in the Federal

Register on October 31, 2000 (65 FR

65018). The licensee affirmed the

applicability of the following NSHC

determination in its application dated

May 4, 2001.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), an

analysis of the issue of no significant

hazards consideration is presented

below:

Criterion 1—The Proposed Change Does Not

Involve a Significant Increase in the

Probability or Consequences of an Accident

Previously Evaluated

The PASS was originally designed to

perform many sampling and analysis

functions. These functions were designed

and intended to be used in post accident

situations and were put into place as a result

of the TMI–2 accident. The specific intent of

the PASS was to provide a system that has

the capability to obtain and analyze samples

of plant fluids containing potentially high

levels of radioactivity, without exceeding

plant personnel radiation exposure limits.

Analytical results of these samples would be

used largely for verification purposes in

aiding the plant staff in assessing the extent

of core damage and subsequent offsite

radiological dose projections. The system

was not intended to and does not serve a

function for preventing accidents and its

elimination would not affect the probability

of accidents previously evaluated.

In the 20 years since the TMI–2 accident

and the consequential promulgation of post

accident sampling requirements, operating

experience has demonstrated that a PASS

provides little actual benefit to post accident

mitigation. Past experience has indicated that

there exists in-plant instrumentation and

methodologies available in lieu of a PASS for

collecting and assimilating information

needed to assess core damage following an

accident. Furthermore, the implementation of

Severe Accident Management Guidance

(SAMG) emphasizes accident management

strategies based on in-plant instruments.

These strategies provide guidance to the

plant staff for mitigation and recovery from

a severe accident. Based on current severe

accident management strategies and

guidelines, it is determined that the PASS

provides little benefit to the plant staff in

coping with an accident.

The regulatory requirements for the PASS

can be eliminated without degrading the

plant emergency response. The emergency

response, in this sense, refers to the

methodologies used in ascertaining the

condition of the reactor core, mitigating the

consequences of an accident, assessing and

projecting offsite releases of radioactivity,

and establishing protective action

recommendations to be communicated to

offsite authorities. The elimination of the

PASS will not prevent an accident

management strategy that meets the initial

intent of the post-TMI–2 accident guidance

through the use of the SAMGs, the

emergency plan (EP), the emergency

operating procedures (EOP), and site survey

monitoring that support modification of

emergency plan protective action

recommendations (PARs).

Therefore, the elimination of PASS

requirements from Technical Specifications

(TS) (and other elements of the licensing

bases) does not involve a significant increase

in the consequences of any accident

previously evaluated.

Criterion 2—The Proposed Change Does Not

Create the Possibility of a New or Different

Kind of Accident From Any Previously

Evaluated

The elimination of PASS related

requirements will not result in any failure

mode not previously analyzed. The PASS

was intended to allow for verification of the

extent of reactor core damage and also to

provide an input to offsite dose projection

calculations. The PASS is not considered an

accident precursor, nor does its existence or

elimination have any adverse impact on the

pre-accident state of the reactor core or post

accident confinement of radionuclides

within the containment building.

Therefore, this change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

Criterion 3—The Proposed Change Does Not

Involve a Significant Reduction in the Margin

of Safety

The elimination of the PASS, in light of

existing plant equipment, instrumentation,

procedures, and programs that provide

effective mitigation of and recovery from

reactor accidents, results in a neutral impact

Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices 31713

to the margin of safety. Methodologies that

are not reliant on PASS are designed to

provide rapid assessment of current reactor

core conditions and the direction of

degradation while effectively responding to

the event in order to mitigate the

consequences of the accident. The use of a

PASS is redundant and does not provide

quick recognition of core events or rapid

response to events in progress. The intent of

the requirements established as a result of the

TMI–2 accident can be adequately met

without reliance on a PASS.

Therefore, this change does not involve a

significant reduction in the margin of safety.

Based upon the reasoning presented above

and the previous discussion of the

amendment request, the requested change

does not involve a significant hazards

consideration.

The NRC staff proposes to determine

that the amendment request involves no

significant hazards consideration.

Attorney for licensee: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120.

NRC Section Chief: Stephen Dembek.

PPL Susquehanna, LLC, Docket Nos.

50–387 and 50–388, Susquehanna

Steam Electric Station, Units 1 and 2,

Luzerne County, Pennsylvania

Date of amendment request: August 8,

2000.

Description of amendment request:

The proposed change would add a new

condition and associated required

actions to Technical Specification (TS)

3.6.1.3, ‘‘Primary Containment Isolation

Valves (PCIVs).’’ The new condition and

actions address the unique containment

isolation features of the hydrogenoxygen (H2O2) analyzer penetrations.

The containment isolation barriers for

the H2O2 analyzer penetrations consist

of two PCIVs in series and a closed

piping system outside primary

containment. Editorial changes

necessary to accommodate the addition

of the proposed requirements were also

proposed.

The licensee also requested approval

for a proposed exception to the

Susquehanna Steam Electric Station

Final Safety Analysis Report

commitments regarding conformance of

the design of closed systems to the

criteria of Section 6.2.4 of NUREG–75/

087, Revision 1, 1975 (Standard Review

Plan). The exception is related to the

boundary valves between the H2O2

analyzer and the post-accident sampling

system (PASS) and is necessary to

permit the use of the H2O2 analyzer

piping system outside primary

containment as a redundant

containment isolation barrier.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change adds a condition to

LCO 3.6.1.3 to address the unique design of

the H2O2 analyzer penetration. The H2O2

analyzer penetration isolation design requires

that both PCIVs and the closed system be

operable in order to support the single failure

criteria and containment integrity. As part of

the proposed change, an exemption

[exception] to NUREG–75/087 guidance on

closed systems for having all closed system

boundary valves to be powered from a Class

1E power source is being requested. The

proposed changes to Technical Specifications

and Technical Specification Bases have no

impact upon the safety functions of the H2O2

Analyzer PCIVs and closed system. The

safety functions of these components are to

maintain primary containment integrity by

limiting leakage following an accident to

within that assumed in the DBA [design-basis

accident] LOCA [loss of coolant accident]

Dose Analysis and to open to permit use of

the H2O2 Analyzer systems post accident.

The H2O2 Analyzer PCIVs and closed system

will be maintained and leak rate tested in

accordance with the Leakage Rate Test

Program, thereby assuring that leakage from

these components is maintained within the

required limits. The design of these

components is such that they meet the

applicable design requirements with the

exception of the PASS closed system

boundary valves discussed above. However,

the potential for a consequential failure of

these valves has been evaluated and

determined to be not credible. Thus, the

proposed changes have no impact upon the

H2O2 Analyzer PCIVs and closed system to

perform their containment isolation function.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

As discussed above, the proposed change

to the Technical Specifications does not

impact upon the safety function of the H2O2

Analyzer PCIVs and closed system. The

safety functions of these components are to

maintain primary containment integrity by

limiting leakage following an accident to

within that assumed in the DBA LOCA Dose

Analysis and to open to permit use of the

H2O2 Analyzer systems post accident.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change does not affect the

safety function of any plant system or

component, and does not have any impact on

plant operation. The proposed change does

not involve a significant reduction in the

margin of safety as currently defined in the

bases of the applicable Technical

Specification section. Therefore, the

proposed change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee: Bryan A. Snapp,

Esquire, Assoc. General Counsel, PPL

Services Corporation, Inc., 2 North

Ninth St., GENTW3, Allentown, PA

18101–1179.

NRC Acting Section Chief: Richard

Correia.

Southern Nuclear Operating Company,

Inc., Georgia Power Company,

Oglethorpe Power Corporation,

Municipal Electric Authority of

Georgia, City of Dalton, Georgia, Docket

No. 50–366, Edwin I. Hatch Nuclear

Plant, Unit 2, Appling County, Georgia

Date of amendment request: May 21,

2001.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications to

eliminate the response time testing

requirements for the reactor protector

system (RPS) signals of reactor high

steam dome pressure and reactor vessel

water level low.

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:

Southern Nuclear Operating Company

(SNC) has reviewed the proposed Technical

Specifications changes described above and

determined they do not involve a significant

hazards consideration based on the

following:

1. The changes do not involve a significant

increase in the probability or consequences

of an accident previously evaluated. The

purpose of the proposed changes is to

eliminate response time testing requirements

for select components in the RPS. However,

because of the continued application of other

existing Technical Specifications

requirements, such as channel calibrations,

channel checks, channel functional tests, and

logic system functional tests, the response

time of the RPS will be maintained within

the acceptance limits assumed in plant safety

analyses. This will assure successful

mitigation of an initiating event. The

proposed Technical Specifications changes

do not affect the capability of the associated

systems to perform their intended function

31714 Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices

within their required response time. The

BWR Owners’ Group (BWROG) has

documented an evaluation in NEDO–32291,

Supplement 1, ‘‘System Analyses for the

Elimination of Selected Response Time

Testing Requirements’’, which was submitted

to the NRC for review and approval as a

Topical Report in December 1997. The

BWROG submitted additional information to

the staff in Addendum 1 to NEDO 32291,

Supplement 1 in November, 1998.

Subsequently, the NRC approved the Topical

Report by a Safety Evaluation Report (SER)

issued in June, 1999.

This evaluation demonstrates that response

time testing is redundant to the other

Technical Specifications requirements listed

in the proceeding paragraph. These other

tests are sufficient to identify failure modes

or degradation in instrument response time

and ensure operation of the associated

systems within acceptance limits.

Furthermore, Addendum 1 to NEDO 32291,

Supplement 1 clearly demonstrates defensein-depth, such that from a realistic basis,

there is no safety significance even if

instrumentation loop response times are

significantly longer than the loop bounding

response times.

2. The proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

analyzed. The proposed Technical

Specifications changes do not affect the

capability of the RPS to perform its intended

function within the acceptance limits

assumed in plant safety analyses. Periodic

surveillance of these RPS instrument loop

components will continue and may be used

to detect degradation that could cause the

response time characteristic to exceed the

BRT [bounting response time] allowance.

3. The proposed changes do not involve a

significant reduction in the margin of safety.

The current Technical Specifications

response times are based on the maximum

allowable values assumed in the plant safety

analyses, which conservatively establish the

margin of safety. As described above, the

proposed Technical Specifications changes

do not affect the capability of the associated

systems to perform their intended function

within the allowed response time used as the

basis for the plant safety analyses. Plant and

system responses to an initiating event will

remain in compliance with the assumptions

of the safety analyses; therefore, the margin

of safety is not affected.

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: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Richard L. Emch,

Jr.

Southern Nuclear Operating Company,

Inc., Georgia Power Company,

Oglethorpe Power Corporation,

Municipal Electric Authority of

Georgia, City of Dalton, Georgia, Docket

No. 50–366, Edwin I. Hatch Nuclear

Plant, Unit 2, Appling County, Georgia

Date of amendment request: May 23,

2001.

Description of amendment request:

The proposed amendment would

change the Safety Limit Minimum

Critical Power Ratios (SLMCPR) in

Technical Specification (TS) 2.1.1.2 to

reflect the results of a cycle-specific

calculation.

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 Technical Specification

changes do not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The derivation of the revised SLMCPRs for

Plant Hatch Unit 2 Cycle 17 for incorporation

into the TS, and their use to determine cyclespecific thermal limits, have been performed

using NRC-approved methods and

procedures. The procedures incorporate

cycle-specific parameters and reduced power

distribution uncertainties in the

determination of the value for SLMCPRs.

These calculations do not change the method

of operating the plant and have no effect on

the probability of an accident initiating event

or transient. The basis of the MCPR Safety

Limit is to ensure no mechanistic fuel

damage is calculated to occur if the limit is

not violated. The new SLMCPRs preserve the

existing margin to transition boiling and the

probability of fuel damage is not increased.

Therefore, the proposed changes do not

involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS changes do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes result only from a

cycle-specific application of NRC-approved

methods to the Unit 2 Cycle 17 core reload.

These changes do not involve any new

method for operating the facility and do not

involve any facility modifications. No new

initiating events or transients result from

these changes. Therefore, the proposed TS

changes do not create the possibility of a new

or different kind of accident from any

accident previously evaluated.

3. The proposed TS changes do not involve

a significant reduction in a margin of safety.

The margin of safety as defined in the TS

bases will remain the same. Cycle-specfic

SLMCPRs are calculated using NRCapproved methods and procedures which are

in accordance with the current fuel design

and licensing criteria. The SLMCPRs remain

high enough to ensure that greater than

99.9% of all fuel rods in the core are

expected to avoid transition boiling if the

limit is not violated, thereby preserving the

fuel cladding integrity. Therefore, the

proposed TS changes do not involve a

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: Ernest L. Blake,

Jr., Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Section Chief: Richard L. Emch,

Jr.

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: April 27,

2001.

Description of amendment request:

The proposed amendments would

revise Technical Specification 3.3.6,

‘‘Containment Ventilation Isolation

Instrumentation,’’ to relax the slave

relay test frequency from every 92 days

to every 18 months.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The results of WCAP–13878 demonstrate

that slave relays are highly reliable. WCAP–

13878 also provides guidance to assure that

slave relays remain highly reliable. The aging

assessment concludes that the age/

temperature-related degradation of all ND

[normally deenergized] relays, and NE

[normally energized] relays produced after

1992, is sufficiently slow such that a

refueling frequency surveillance interval will

not significantly increase the probability of

slave relay failures. Finally, the evaluation of

the auxiliary relays actuated during slave

relay testing has concluded that based on the

tests of the auxiliary relays performed during

other equipment testing, reasonable

assurance is provided that failures will be

identified if the associated slave relays are

tested on a refueling frequency.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed changes do not alter the

performance of the CVI [containment

Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices 31715

ventilation isolation] systems assumed in the

plant safety analysis. Changing the interval

for periodically verifying CVI slave relays

(assuring equipment operability) will not

create any new accident initiators or

scenarios.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated for VEGP [Vogtle

Electric Generating Plant].

3. Does the change involve a significant

reduction in a margin of safety?

The proposed changes do not affect the

total CVI response assumed in the safety

analysis since the reliability of the slave

relays will not be significantly affected by the

decreased surveillance frequency.

Therefore, the proposed changes do not

involve a significant reduction in a margin of

safety.

Based on the above safety evaluation,

VEGP concludes that the changes proposed

by this submittal satisfy the no significant

hazards consideration standards of 10 CFR

50.92(c) and, accordingly, a no significant

hazards finding is justified.

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.

STP Nuclear Operating Company,

Docket Nos. 50–498 and 50–499, South

Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request:

December 20, 2000.

Description of amendment request:

The proposed change will delete

Condition 2.G, ‘‘Reporting to the

Commission,’’ and Technical

Specification 6.6.1.a, ‘‘Reportable Event

Action.’’

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:

Pursuant to 10 CFR 50.92, it has been

determined that this request involves no

significant hazards considerations. The

determination of no significant hazards was

made by applying the Nuclear Regulatory

Commission established standards contained

in 10 CFR 50.92. These standards assure that

any changes to the operation of South Texas

Project in accordance with this request

consider the following:

(1) Will the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

Response: No.

This request involves administrative

changes only. No actual plant equipment or

accident analyses will be affected by the

proposed changes. Therefore, this request

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

(2) Will the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

Response: No.

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, this request does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

(3) Will the change involve a significant

reduction in a margin of safety?

Response: No.

Margin of safety is associated with

confidence in the ability of the fission

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

Reactor Coolant System pressure boundary,

and containment structure) to limit the level

of radiation dose to the public. This request

involves administrative changes only.

No actual plant equipment or accident

analyses will be affected by the proposed

change. Additionally, the proposed changes

will not relax any criteria used to establish

safety limits, will not relax any safety

systems settings, or will not relax the bases

for any limiting conditions of operation.

Therefore, these proposed changes will not

impact the margin of safety.

Conclusion: Based upon the analysis

provided herein, the proposed amendments

will not increase the probability or

consequences of an accident previously

evaluated, create the possibility of a new or

different kind of accident from any accident

previously evaluated, or involve a reduction

in a margin of safety. Therefore, the proposed

amendments meet the requirements of 10

CFR 50.92 and do not involve a significant

hazards consideration.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the 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: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, NW.,

Washington, DC 20036–5869.

NRC Section Chief: Robert A. Gramm.

STP Nuclear Operating Company,

Docket Nos. 50–498 and 50–499, South

Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: February

12, 2001.

Description of amendment request:

The proposed amendment will revise

Technical Specifications surveillance

requirement 4.4.6.2.2.e,which refers to

American Society of Mechanical

Engineers (ASME) Boiler and Pressure

Vessel Code Section XI, paragraph IWV–

3427(b) as a requirement for

demonstrating that each Reactor Coolant

System Pressure Isolation Valve

specified in TS Table 3.4–1 is operable.

Part 10 of the ASME Operations and

Maintenance (OM) Standards, OMa1988, is currently the applicable code

for these valves and does not have these

requirements.

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:

Pursuant to 10 CFR 50.91, this analysis

provides a determination that the proposed

change to the Technical Specifications

described previously does not involve any

significant hazards consideration as defined

in 10 CFR 50.92.

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

This Technical Specification change only

affects trending of valve leakage rate test

results to anticipate the expected leakage rate

performance of Reactor Coolant System

pressure isolation valves. Redundant

pressure isolation valves are included in the

plant to ensure continued protection of lower

pressure systems from exposure to the higher

pressure of the Reactor Coolant System in the

event that excessive leakage develops in an

isolation valve. In addition, leakage rate tests

of Reactor Coolant System pressure isolation

valves will continue to be performed with no

change in the accepted amount of leakage or

frequency. Therefore, the proposed change

does not involve a significant increase in the

probability of an accident previously

evaluated.

The limiting event associated with these

valves is a Loss of Coolant Accident. This has

already been reviewed as part of the South

Texas Project Updated Final Safety Analysis

Report. Therefore, the proposed change does

not involve a significant increase in the

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 proposed change only removes a

requirement for trending of pressure isolation

valve leakage rates. The proposed change

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

There is no change in the design of the

plant associated with this proposed license

31716 Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices

amendment. The only impact of this change

is in the prediction of when a particular

pressure isolation valve may have a leakage

rate higher than what is allowed. Adverse

test results will be addressed under the

corrective action program and by application

of the Maintenance Rule. Engineering

analysis of test results can take into account

special circumstances associated with a test

that would affect the conclusions.

Leakage rate test measurements of South

Texas Project Reactor Coolant System

isolation valves will continue to be taken

pursuant to the surveillance requirements of

Technical Specification 4.4.6.2.2, which is

consistent with the requirements of code

OMa-1988, paragraph 4.2.2.3.e for analysis of

leakage rates. Code OMa-1988, paragraph 6.3,

requires records of tests, including analysis

of deviations in test values. 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 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: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, NW.,

Washington, DC 20036–5869.

NRC Section Chief: Robert A. Gramm.

STP Nuclear Operating Company,

Docket Nos. 50–498 and 50–499, South

Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: February

28, 2001.

Description of amendment request:

The proposed amendment will revise

the Technical Specifications (TS) to

eliminate periodic response time testing

requirements on selected sensors and

selected protection channels, and will

modify TS Section 1.0 Definitions for

‘‘ENGINEERED SAFETY FEATURE

(ESF) RESPONSE TIME’’ and

‘‘REACTOR TRIP SYSTEM (RTS)

RESPONSE TIME’’ to provide for

verification of response time for selected

components. Surveillances 4.3.1.2 and

4.3.2.2 will be modified consistent with

the new definitions. The associated

Bases will be revised.

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:

Pursuant to 10 CFR 50.92, it has been

determined that this request involves no

significant hazards consideration. The

determination of no significant hazards was

made by applying the Nuclear Regulatory

Commission established standards contained

in 10 CFR 50.92. These standards assure that

any changes to the operation of South Texas

Project in accordance with this request

consider the following:

(1) Will the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

Response: No.

This change to the Technical

Specifications does not result in a condition

where the design, material, and construction

standards that were applicable prior to the

change are altered. The same RTS [Reactor

Trip System] and ESFAS [Engineered Safety

Features Actuation System] instrumentation

is being used; the time response allocations/

modeling assumptions in the Chapter 15

analyses are still the same; only the method

of verifying time response is changed. The

proposed change will not modify any system

interface and could not increase the

likelihood of an accident since these events

are independent of this change. The

proposed activity will not change, degrade or

prevent actions or alter any assumptions

previously made in evaluating the

radiological consequences of an accident

described in the SAR [Safety Analysis

Report]. Therefore, the proposed amendment

does not result in any increase in the

probability or consequences of an accident

previously evaluated.

(2) Will the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

Response: No.

This change does not alter the performance

of the pressure and differential pressure

transmitters and switches, Process Protection

racks, Nuclear Instrumentation, and Logic

Systems used in the plant protection systems.

All sensors, Process Protection racks, Nuclear

Instrumentation, and Logic Systems will still

have response time verified by test before

placing the equipment into operational

service and after any maintenance that could

affect the response time. Changing the

method of periodically verifying instrument

response times for certain equipment

(assuring equipment operability) from time

response testing to calibration and channel

checks will not create any new accident

initiators or scenarios. Periodic surveillance

of these instruments will detect significant

degradation in the equipment response time

characteristics. Implementation of the

proposed amendment does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Will the change involve a significant

reduction in a margin of safety?

Response: No.

This change does not affect the total system

response time assumed in the safety analysis.

The periodic system response time

verification method for selected pressure and

differential pressure sensors and for Process

Protection racks, Nuclear Instrumentation,

and Logic Systems is modified to allow use

of actual test data or engineering data. The

method of verification still provides

assurance that the total system response time

is within that assumed in the safety analysis.

Based on the above, it is concluded that the

proposed license amendment request does

not result in a reduction in margin of safety.

Conclusion: Based on the preceding

analysis, it is concluded that elimination of

periodic equipment response time testing is

acceptable and the proposed license

amendment does not involve a Significant

Hazards Consideration as defined in 10 CFR

50.92.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the 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: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, NW.,

Washington, DC 20036–5869.

NRC Section Chief: Robert A. Gramm.

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),

Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices 31717

Rockville, Maryland. Publicly available

records will be accessible from the

Agencywide Documents Access and

Management Systems (ADAMS) Public

Electronic Reading Room on the internet

at the NRC Web site, http://

www.nrc.gov/NRC/ADAMS/index.html.

If you do not have access to ADAMS or

if there are problems in accessing the

documents located in ADAMS, contact

the NRC Public Document Room (PDR)

Reference staff at 1–800–397–4209, 301–

415–4737 or by email to pdr@nrc.gov.

AmerGen Energy Company, LLC,

Docket No. 50–461, Clinton Power

Station, Unit 1, DeWitt County, Illinois

Date of application for amendment:

March 1, 2001.

Brief description of amendment: The

amendment increases the reactor core

isolation cooling system surveillance

test upper pressure limit.

Date of issuance: May 31, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 139.

Facility Operating License No. NPF–62:

The amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 4, 2001 (66 FR 17964).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 31, 2001.

No significant hazards consideration

comments received: No.

AmerGen Energy Company, LLC,

Docket No. 50–461, Clinton Power

Station, Unit 1, DeWitt County, Illinois

Date of application for amendment:

October 6, 2000 (U–603329).

Brief description of amendment: The

amendment relocates Technical

Specification Figure 3.6.4.1–1,

‘‘Secondary Containment Drawdown

Time for 1500 cfm Boundary Leakage’’

to plant procedures.

Date of issuance: June 1, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 140.

Facility Operating License No. NPF–62:

The amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 29, 2000 (65 FR

71132).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 1, 2001.

No significant hazards consideration

comments received: No.

Dominion Nuclear Connecticut, Inc.,

Docket No. 50–336, Millstone Nuclear

Power Station, Unit No. 2, New London

County, Connecticut

Date of application for amendment:

December 21, 2000.

Brief description of amendment: The

amendment revised Technical

Specification (TS) 3.7.11 to allow plant

operation to continue if the temperature

of the Ultimate Heat Sink (UHS) exceeds

the TS limit of 75 °F provided the water

temperature, averaged over the previous

24-hour period, is at or below 75 °F.

This operational flexibility only applies

if the UHS temperature is between 75 °F

and 77 °F. The action time requirements

if the UHS temperature exceeds 77 °F,

or if the 24-hour averaged value exceeds

75 °F still apply. An associated footnote

that is no longer applicable was deleted,

and the associated TS Bases were

modified to reflect these changes.

Date of issuance: May 31, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days from the date of

issuance.

Amendment No.: 257.

Facility Operating License No. DPR–65:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 18, 2001 (66 FR 20007).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 31, 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:

February 6, 2001, as supplemented by

letter dated May 1, 2001.

Brief description of amendment: The

amendment revised the Technical

Specifications associated with the

reactor coolant system leakage detection

systems, to make them consistent with

the requirements in NUREG–1432,

‘‘Standard Technical Specifications,

Combustion Engineering Plants.’’

Date of issuance: May 29, 2001.

Effective date: As of the date of

issuance to be implemented within 60

days from the date of issuance.

Amendment No.: 231.

Facility Operating License No. NPF–6:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 7, 2001 (66 FR 13803).

The May 1, 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 May 29, 2001.

No significant hazards consideration

comments received: No.

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Entergy Mississippi, Inc., Docket

No. 50–416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of application for amendment:

January 24, 2000.

Brief description of amendment:

Entergy Operations, Inc. requests

revisions to the Grand Gulf Nuclear

Station Technical Specifications which

specify the minimum useable fuel oil

inventories to be maintained in the

Division 1, 2, and 3 Diesel Generator

Fuel Oil Storage Tanks.

Date of issuance: May 24, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 60 days of issuance.

Amendment No: 147.

Facility Operating License No. NPF–29:

The amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: March 22, 2000 (65 FR

15381).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 24, 2001.

No significant hazards consideration

comments received: No.

Exelon Generation Company, LLC,

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:

November 7, 2000 as supplemented by

letter dated March 23, 2001.

Brief description of amendments: The

amendments would revise the technical

specifications (TS) to extend the TS

surveillance test interval (STI) from a

92-day STI to an 18-month STI, for the

solid state protection system (SSPS)

slave relay types that meet the

acceptance criteria for the reliability

assessments performed in accordance

with the methodology described in the

NRC approved Westinghouse Electric

Corporation Topical Reports.

Date of issuance: May 31, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment Nos.: 121, 121, 115, and

115.

31718 Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices

Facility Operating License Nos. NPF–

37, NPF–66, NPF–72 and NPF–77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: February 21, 2001 (66 FR

11053).

The supplemental information

contained clarifying information and

did not change the initial no significant

hazards consideration determination

and did not expand the scope of the

original Federal Register notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 31, 2001.

No significant hazards consideration

comments received: No.

Exelon Generation Company, Docket

Nos. 50–352 and 50–353, Limerick

Generating Station, Units 1 and 2,

Montgomery County, Pennsylvania

Date of application for amendments:

October 25, 2000.

Brief description of amendments:

Revised the Action Statements

associated with Technical Specification

(TS) Table 3.3.7.5–1,‘‘Accident

Monitoring Instrumentation,’’

concerning the Drywell Hydrogen/

Oxygen (H2/O2) Concentration

Analyzers, and the associated TS Bases.

Date of issuance: As of date of

issuance and shall be implemented

within 30 days.

Effective date: May 24, 2001.

Amendment Nos.: 151 and 115.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 27, 2000 (65 FR

81929).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 24, 2001.

No significant hazards consideration

comments received: No.

FirstEnergy Nuclear Operating

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

Ottawa County, Ohio

Date of application for amendment:

August 7, 2000, as supplemented

February 6, 2001.

Brief description of amendment: This

amendment will change Technical

Specification (TS) Section Bases 3/4.3.1

and 3/4.3.2 to clarify the actions that

must be performed when Steam and

Feedwater Rupture Control System

(SFRCS) components and SFRCSactuated components are inoperable.

Specifically, the changes will provide

guidance on which TS actions are

applicable for SFRCS-actuated

components. The changes will also add

a new TS 3/4.7.1.8 which would

provide appropriate requirements for

the Main Feedwater Control Valves and

the Startup Feedwater Control Valves.

Additionally, the changes add TS 3/

4.7.1.9 which will provide requirements

for the Turbine Stop Valves. The

changes are consistent with the intent of

NUREG–1430, ‘‘Standard Technical

Specifications—Babcock and Wilcox

Plants,’’ Revision 1, April 1995.

Date of issuance: May 29, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 120 days.

Amendment No.: 246.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 1, 2000 (65 FR

65342).

The supplemental information

contained clarifying information and

did not change the initial no significant

hazards consideration determination

and did not expand the scope of the

original Federal Register notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 29, 2001.

No significant hazards consideration

comments received: No.

Nuclear Management Company, LLC,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of application for amendment:

April 6, 2001.

Brief description of amendment: The

amendment revises the Kewaunee

Nuclear Power Plant (KNPP) Technical

Specifications (TSs) Section 6.2,

‘‘Organization,’’ and Section 6.13, ‘‘High

Radiation Area’’ to reflect the title

change from Shift Supervisor to Shift

Manager.

Date of issuance: June 1, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment No.: 154.

Facility Operating License No. DPR–43:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 2, 2001 (66 FR 22031).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 1, 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:

October 30, 2000.

Brief description of amendments: The

amendments approve the insertion of

breakaway ceramic pins into the latches

of eight double-leaf doors in the

auxiliary building special ventilation

zone in order to restrain the doors and

reduce the frequency of open-door

position alarms. The ceramic latch pins

are designed to break at forces well

below the differential pressure that

would be generated in the auxiliary

building as a result of a postulated highenergy line break (HELB), and thereby

allow the doors to swing open and

create a relief path from the auxiliary

building. Therefore, the modification

provides the restraints needed to reduce

the frequency of open-door position

alarms; but without impeding the doors’

steam relief function that was assumed

in the design-basis HELB analysis.

Date of issuance: May 30, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days.

Amendment Nos.: 157 and 148.

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revise the

Updated Final Safety Analysis Report.

Date of initial notice in Federal

Register: March 21, 2001 (66 FR

15928).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 30, 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:

February 16, 2001 (ULNRC–04390).

Brief description of amendment: The

amendment revises Technical

Specification 5.2.1.c to replace the title

‘‘Vice President and Chief Nuclear

Officer’’ with ‘‘Senior Vice President

and Chief Nuclear Officer.’’

Date of issuance: May 30, 2001.

Effective date: May 30, 2001, to be

implemented within 60 days from the

date of issuance.

Amendment No.: 145.

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 4, 2001 (66 FR 17971)

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 30, 2001.

No significant hazards consideration

comments received: No.

Federal Register / Vol. 66, No. 113 / Tuesday, June 12, 2001 / Notices 31719

Virginia Electric and Power Company,

et al., Docket Nos. 50–280 and 50–281,

Surry Power Station, Units 1 and 2,

Surry County, Virginia

Date of application for amendments:

December 12, 2000, as supplemented by

letters dated January 8, and February 22,

2001.

Brief Description of amendments: The

amendments revise Technical

Specification Section 3.17 and

associated Bases. The proposed changes

will accommodate a vacuum-assisted

fill technique for backfilling isolated

reactor coolant system (RCS) loops from

the active volume of the RCS.

Date of issuance: May 22, 2001.

Effective date: As of the date of

issuance and shall be implemented

within 30 days of issuance.

Amendment Nos.: 226 and 226.

Facility Operating License Nos. DPR–

32 and DPR–37: Amendments change

the Technical Specifications.

Date of initial notice in Federal

Register: March 21, 2001 (66 FR 15932).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 22, 2001.

No significant hazards consideration

comments received: No.

Dated at Rockville, Maryland this 5th day

of June 2001.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–14755 Filed 6–11–01; 8:45 am]

BILLING CODE 7590–01–U

NUCLEAR REGULATORY

COMMISSION

Procedures for Providing Security

Support for NRC Public Meetings/

Hearings

The Nuclear Regulatory Commission

(NRC) is revising its procedures for

providing security support for all public

NRC forums. This revision will provide

a single set of procedures that will

ensure consistency and uniformity in

providing security support for these

meetings. These procedures will be used

by NRC headquarters and regional staff

and are applicable to public hearings/

meetings held at NRC headquarters

buildings, other NRC space in the

Washington, D.C. area, and/or regional

locations to include space leased for the

occasion. This Federal Register notice

supersedes the previous Federal

Register notice, entitled ‘‘Security

Support for NRC Meetings/Hearings,’’

that was published on November 1,

1991 (56 FR 19451).

In order to balance the orderly

conduct of government business with

the right of free speech, the following

procedures regarding attendance at NRC

public meetings and hearings have been

established:

Visitors (other than properly identified

Congressional, press, and government

personnel) may be subject to personnel

screening, such as passing through metal

detectors and inspecting visitors’ briefcases,

packages, etc.

Signs, banners, posters and displays will

be prohibited from all NRC adjudicatory

proceedings (Commission and Atomic Safety

and Licensing Board Panel hearings) because

they are disruptive to the conduct of the

adjudicatory process. Signs, banners, posters

and displays not larger than 18′′×18′′ will be

permitted at all other NRC proceedings, but

cannot be waved, held over one’s head or

generally moved about while in the meeting

room. Signs, banners, posters and displays

larger than 18′′×18′′ will not be permitted in

the meeting room because they are disruptive

both to the participants and the audience.

Additionally, signs, banners, posters and

displays affixed to any sticks, poles or other

similar devices will not be permitted in the

meeting room.

The presiding official will note, on the

record, any disruptive behavior and warn the

person to cease the behavior. If the person

does not cease the behavior, the presiding

official may call a brief recess to restore order

and/or ask one of the security personnel on

hand to remove the person.

For Further Information Contact:

Calvin O. Byrd, Chief, Physical Security

Branch, Division of Facilities and

Security, Office of Administration, U. S.

Nuclear Regulatory Commission,

Washington, DC 20555–0001, telephone:

301–415–7402.

Dated at Rockville, Maryland, this 6th day

of June 2001.

For the Nuclear Regulatory Commission.

Thomas O. Martin,

Director, Division of Facilities and Security,

Office of Administration.

[FR Doc. 01–14752 Filed 6–11–01; 8:45 am]

BILLING CODE 7590–01–P

OFFICE OF PERSONNEL

MANAGEMENT

Submission for OMB Review;

Comment Request for Reclearance of

a Revised Information Collection: SF

2802 and SF 2802B

AGENCY: Office of Personnel

Management.

ACTION: Notice.

SUMMARY: In accordance with the

Paperwork Reduction Act of 1995

(Public Law 104–13, May 22, 1995), this

notice announces that the Office of

Personnel Management (OPM) has

submitted to the Office of Management

and Budget (OMB) a request for

reclearance of a revised information

collection. SF 2802, Application for

Refund of Retirement Deductions (Civil

Service Retirement System), is used to

support the payment of monies from the

Retirement Fund. It identifies the

applicant for refund of retirement

contributions. SF 2802B, Current/

Former Spouse’s Notification of

Application for Refund of Retirement

Deductions, is used to comply with the

legal requirement that any spouse or

former spouse of the applicant has been

notified that the former employee is

applying for a refund.

Approximately 32,100 SF 2802 forms

are completed annually. We estimate it

takes approximately 45 minutes to

complete the form. The annual burden

is 24,075 hours. Approximately 28,890

SF 2802B forms are processed annually.

We estimate it takes approximately 15

minutes to complete this form. The

annual burden is 7,223 hours. The total

annual burden is 31,298 hours.

For copies of this proposal, contact

Mary Beth Smith-Toomey on (202) 606–

8358, or email to mbtoomey@opm.gov.

DATES: Comments on this proposal

should be received on or before July 12,

2001.

ADDRESSES: Send or deliver comments

to—

Ronald W. Melton, Chief, Operations

Support Division, Retirement and

Insurance Service, U.S. Office of

Personnel Management, 1900 E Street,

NW, Room 3349A, Washington, DC

20415–3540, and

Joseph Lackey, OPM Desk Officer,

Office of Information & Regulatory

Affairs, Office of Management and

Budget, New Executive Office

Building, NW., Room 10235,

Washington, DC 20503.

For Information Regarding

Administrative Coordination— Contact:

Donna G. Lease, Team Leader, Budget

and Administrative Services Division,

Forms Analysis and Design, (202) 606–

0623.

U.S. Office of Personnel Management.

Steven R. Cohen,

Acting Director.

[FR Doc. 01–14711 Filed 6–11–01; 8:45 am]

BILLING CODE 6325–50–P