59082 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 

available in the NRC Public Document 

Room (PDR), One White Flint North, 

Room O–1F21, 11555 Rockville Pike, 

Rockville, MD 20852–2738. ACNW 

meeting agenda, transcripts, and letter 

reports are available through the NRC 

Public Document Room at pdr@nrc.gov, 

by calling the PDR at 1–800–394–4209, 

or from the Publicly Available Records 

System (PARS) component of NRC’s 

document system (ADAMS) which is 

accessible from the NRC Web site at 

http://www.nrc.gov/reading-rm/ 

adams.html or http://www.nrc.gov/ 

reading-rm/doc-collections/. A copy of 

the certified minutes of the meeting will 

be available at the same location up to 

three months following the meeting. 

Copies may be obtained upon payment 

of appropriate reproduction charges. 

(f) Video teleconferencing service is 

available for observing open sessions of 

some ACNW meetings. Those wishing 

to use this service for observing ACNW 

meetings should contact Mr. Theron 

Brown, ACNW Audio Visual 

Technician, (301–415–8066) between 

7:30 a.m. and 3:45 p.m. Eastern Time at 

least 10 days before the meeting to 

ensure the availability of this service. 

Individuals or organizations requesting 

this service will be responsible for 

telephone line charges and for providing 

the equipment and facilities that they 

use to establish the video 

teleconferencing link. The availability of 

video teleconferencing services is not 

guaranteed. 

(g) The meeting room is handicapped 

accessible. 

ACNW Working Group Meetings 

From time to time the ACNW may 

sponsor an in-depth meeting on a 

specific technical issue to understand 

staff expectations and review work in 

progress. Such meetings are called 

Working Group meetings. These 

Working Group meetings will also be 

conducted in accordance with these 

procedures noted above for the ACNW 

meeting, as appropriate. When Working 

Group meetings are held at locations 

other than at NRC facilities, 

reproduction facilities may not be 

available at a reasonable cost. 

Accordingly, 50 additional copies of the 

materials to be used during the meeting 

should be provided for distribution at 

such meetings. 

Special Provisions When Proprietary 

Sessions Are To Be Held 

If it is necessary to hold closed 

sessions for the purpose of discussing 

matters involving proprietary 

information, persons with agreements 

permitting access to such information 

may attend those portions of the ACNW 

meetings where this material is being 

discussed upon confirmation that such 

agreements are effective and related to 

the material being discussed. 

The DFO should be informed of such 

an agreement at least five working days 

prior to the meeting so that it can be 

confirmed, and a determination can be 

made regarding the applicability of the 

agreement to the material that will be 

discussed during the meeting. The 

minimum information provided should 

include information regarding the date 

of the agreement, the scope of material 

included in the agreement, the project 

or projects involved, and the names and 

titles of the persons signing the 

agreement. Additional information may 

be requested to identify the specific 

agreement involved. A copy of the 

executed agreement should be provided 

to the DFO prior to the beginning of the 

meeting for admittance to the closed 

session. 

Dated: October 5, 2005. 

Annette L. Vietti-Cook, 

Secretary of the Commission. 

[FR Doc. 05–20317 Filed 10–7–05; 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 section 189a.(2) of the 

Atomic Energy Act of 1954, as amended 

(the Act), the U.S. Nuclear Regulatory 

Commission (the Commission or NRC 

staff) is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and 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 September 

6, 2005, to September 29, 2005. The last 

biweekly notice was published on 

September 27, 2005 (70 FR 56499). 

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. Within 60 days after the 

date of publication of this notice, 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. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 59083 

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 Commission’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. The filing of 

requests for a hearing and petitions for 

leave to intervene is discussed below. 

Within 60 days after the date of 

publication of this notice, 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 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

reading-rm/doc-collections/cfr/. If a 

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, 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 general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

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/ 

requestor to relief. A petitioner/ 

requestor who fails to satisfy 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. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, 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 by: 

(1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, Attention: Rulemaking and 

Adjudications Staff; (2) courier, express 

mail, and expedited delivery services: 

Office of the Secretary, Sixteenth Floor, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland, 20852, 

Attention: Rulemaking and 

Adjudications Staff; (3) E-mail 

addressed to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

HearingDocket@nrc.gov; or (4) facsimile 

transmission addressed to the Office of 

the Secretary, U.S. Nuclear Regulatory 

Commission, Washington, DC, 

Attention: Rulemakings and 

Adjudications Staff at (301) 415–1101, 

verification number is (301) 415–1966. 

A copy of the request for hearing and 

petition for leave to intervene should 

also be sent to the Office of the General 

Counsel, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and it is requested that copies be 

transmitted either by means of facsimile 

transmission to (301) 415–3725 or by email to OGCMailCenter@nrc.gov. A copy 

of the request for hearing and petition 

for leave to intervene should also be 

sent to the attorney for the licensee. 

Nontimely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission or the presiding officer of 

the Atomic Safety and Licensing Board 

that the petition, request and/or the 

contentions should be granted based on 

a balancing of the factors specified in 10 

CFR 2.309(a)(1)(i)–(viii). 

For further details with respect to this 

action, see the application for 

amendment which is available for 

public inspection at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the ADAMS Public 

Electronic Reading Room on the Internet 

at the NRC Web site, http:// 

www.nrc.gov/reading-rm/adams.html. If 

you do not have access to ADAMS or if 

there are problems in accessing the 

documents located in ADAMS, contact 

the PDR Reference staff at 1 (800) 397–

4209, (301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

59084 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 

Carolina Power & Light Company, et al., 

Docket No. 50–400, Shearon Harris 

Nuclear Power Plant, Unit 1, Wake and 

Chatham Counties, North Carolina 

Date of amendment request: April 6, 

2005, as supplemented by letter dated 

August 8, 2005. 

Description of amendment request: 

The proposed amendment will modify 

Technical Specification (TS) 6.8.4.k, 

‘‘Containment Leakage Rate Testing 

Program,’’ and TS Surveillance 

Requirement (SR) 4.6.1.6.1, 

‘‘Containment Vessel Surfaces.’’ The 

proposed amendment would modify the 

TS to allow for a one-time extension of 

the containment Type A test interval 

from once in 10 years to once in 15 

years. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

This change does not involve a 

significant hazards consideration for the 

following reasons: 

1. The proposed amendment does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

The proposed change to HNP [Harris 

Nuclear Plant] TS 6.8.4.k and TS SR 4.6.1.6.1 

provide a one-time extension of the 

containment Type A test interval from 10 

years to 15 years and specifies that additional 

visual inspections are done in accordance 

with Subsections IWE and IWL of the ASME 

[American Society of Mechanical Engineers] 

Section XI Code. The existing 10-year test 

interval is based on past test performance. 

The proposed TS change does not involve a 

physical change to the plant or a change in 

the manner in which the plant is operated or 

controlled. The containment vessel is 

designed to provide a leak-tight barrier 

against the uncontrolled release of 

radioactivity to the environment in the 

unlikely event of postulated accidents. As 

such, the containment vessel is not 

considered as the initiator of an accident. 

Therefore, the proposed TS change does not 

involve a significant increase in the 

probability of an accident previously 

evaluated. 

The proposed change involves only a onetime change to the interval between 

containment Type A tests. Type B and C 

leakage testing will continue to be performed 

at the intervals specified in 10 CFR Part 50, 

Appendix J, Option A, as required by the 

HNP TS. As documented in NUREG–1493, 

‘‘Performance-Based Containment LeakageTest Program,’’ industry experience has 

shown that Type B and C containment leak 

rate tests have identified a very large 

percentage of containment leak paths, and 

that the percentage of containment leak paths 

that are detected only by Type A testing is 

very small. In fact, an analysis of 144 

integrated leak rate tests, including 23 

failures, found that none of the failures 

involved a containment liner breach. 

NUREG–1493 also concluded, in part, that 

reducing the frequency of containment Type 

A testing to once per 20 years results in an 

imperceptible increase in risk. The HNP test 

history and risk-based evaluation of the 

proposed extension to the Type A test 

interval supports this conclusion. The design 

and construction requirements of the 

containment vessel, combined with the 

containment inspections performed in 

accordance with the American Society of 

Mechanical Engineers (ASME) Code, Section 

XI, and the Maintenance Rule (10 CFR 50.65) 

provide a high degree of assurance that the 

containment vessel will not degrade in a 

manner that is detectable only by Type A 

testing. Therefore, the proposed TS change 

does not involve a significant increase in the 

consequences of an accident previously 

evaluated. 

Therefore, this change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. The proposed amendment does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

The proposed change to HNP TS 6.8.4.k 

and TS SR 4.6.1.6.1 provide a one-time 

extension of the containment Type A test 

interval to 15 years and specifies that 

additional visual inspections are done in 

accordance with Subsections IWE and IWL of 

the ASME Section XI Code. The existing 10- 

year test interval is based on past test 

performance. The proposed change to the 

Type A test interval does not result in any 

physical changes to HNP. In addition, the 

proposed test interval extension does not 

change the operation of HNP such that a 

failure mode involving the possibility of a 

new or different kind of accident from any 

accident previously evaluated is created. 

Thus, this change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. The proposed amendment does not 

involve a significant reduction in a margin of 

safety. 

The proposed change to HNP TS 6.8.4.k 

and TS SR 4.6.1.6.1 provide a one-time 

extension of the containment Type A test 

interval from 10 years to 15 years and 

specifies that additional visual inspections 

are done in accordance with Subsections IWE 

and IWL of the ASME Section XI Code. The 

existing 10-year test interval is based on past 

test performance. The NUREG–1493 study of 

the effects of extending containment leak rate 

testing found that a 20 year extension for 

Type A testing resulted in an imperceptible 

increase in risk to the public. NUREG–1493 

found that, generically, the design 

containment leak rate contributes a very 

small amount to the individual risk and that 

the decrease in Type A testing frequency 

would have a minimal affect on this risk 

since most potential leak paths are detected 

by Type B and C testing. The proposed 

change involves only a one-time extension of 

the interval for containment Type A testing; 

the overall containment leak rate specified by 

the HNP TS is being maintained. Type B and 

C testing will continue to be performed at the 

frequency required by the HNP TS. The 

regular containment inspections being 

performed in accordance with the ASME 

Code, Section XI, and the Maintenance Rule 

(10 CFR 50.65) provide a high degree of 

assurance that the containment will not 

degrade in a manner that is only detectable 

by Type A testing. In addition, a plantspecific risk evaluation has demonstrated 

that the one-time extension of the Type A test 

interval from 10 years to 15 years results in 

a very small increase in risk for those 

accident sequences influenced by Type A 

testing. 

Therefore, this change does not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David T. 

Conley, Associate General Counsel II—

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, North Carolina 27602. 

NRC Section Chief: Michael L. 

Marshall, Jr. 

Carolina Power & Light Company, et al., 

Docket No. 50–400, Shearon Harris 

Nuclear Power Plant (HNP), Unit 1, 

Wake and Chatham Counties, North 

Carolina 

Date of amendment request: June 20, 

2005. 

Description of amendment request: 

The amendment would revise Technical 

Specifications (TS) 3/4.4.7, ‘‘Reactor 

Coolant System Chemistry.’’

Specifically, the proposed amendment 

would revise the footnotes in Tables 

3.4–2 and 4.4–3 of the TS to increase the 

temperature limit from 180 °F to 250 °F 

above which reactor coolant sampling 

and analysis for dissolved oxygen is 

required and dissolved oxygen limits 

apply. 

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: 

This amendment does not involve a 

significant hazards consideration for the 

following reasons: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

Operation of HNP in accordance with the 

proposed amendment does not increase the 

Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 59085 

probability or consequences of accidents 

previously evaluated. The Final Safety 

Analysis Report (FSAR) documents the 

analyses of design basis accidents (DBA) at 

HNP. Any scenario or previously analyzed 

accident that results in offsite dose were 

evaluated as part of this analysis. The 

proposed amendment does not change or 

affect any accident previously evaluated in 

the FSAR. The proposed amendment does 

not modify any plant equipment. In addition, 

the proposed amendment does not result in 

a change to a structure, system, or component 

(SSC), or adversely affect its design function. 

The purpose of the temperature limit for 

RCS [Reactor Coolant System] oxygen control 

is to minimize corrosion at high temperatures 

on RCS components. Increasing the 

temperature at which oxygen levels are 

required to be maintained within specified 

limits from 180 °F to 250 °F is supported by 

industry and vendor data which indicates 

that the influence of dissolved oxygen at or 

below 250 °F is not significant with regard 

to stress corrosion cracking and general 

corrosion of RCS components. The proposed 

amendment is consistent with the Electric 

Power Research Institute’s (EPRI’s) 

guidelines for Pressurized Water Reactor 

(PWR) Primary Water Chemistry. This 

amendment places HNP in line with standard 

industry specifications for reactors of similar 

size and vintage. HNP’s proposed 

amendment to increase the temperature limit 

for applicability to 250 °F would decrease the 

time needed to achieve compliance with the 

dissolved oxygen limit and decrease the 

overall time to restart the plant from cold 

shutdown. Removing oxygen in a more 

expeditious fashion enhances RCS chemistry. 

Based on the above, RCS integrity is 

maintained by this amendment. 

Therefore, this amendment does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Operation of HNP in accordance with the 

proposed amendment does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. The FSAR documents the analyses 

of design basis accidents (DBA) at HNP. Any 

scenario or previously analyzed accident that 

results in offsite dose were evaluated as part 

of this analysis. The proposed amendment 

does not change or affect any accident 

previously evaluated in the FSAR, and no 

new or different scenarios are created by the 

proposed amendment to the TS. The 

proposed amendment does not modify any 

plant equipment. In addition, the proposed 

amendment does not result in a change to an 

SSC [structure, system, or component] or 

adversely affect its design function. 

The purpose of the temperature limit for 

RCS oxygen control is to minimize corrosion 

at high temperatures on RCS components. 

Increasing the temperature at which oxygen 

levels are required to be maintained within 

specified limits from 180 °F to 250 °F is 

supported by industry and vendor data 

which indicates that the influence of 

dissolved oxygen at or below 250 °F is not 

significant with regard to stress corrosion 

cracking and general corrosion of RCS 

components. The proposed amendment is 

consistent with EPRI’s guidelines for PWR 

Primary Water Chemistry. This amendment 

places HNP in line with standard industry 

specifications for reactors of similar size and 

vintage. HNP’s proposed amendment to 

increase the temperature limit for 

applicability to 250 °F would decrease the 

time needed to achieve compliance with the 

dissolved oxygen limit and decrease the 

overall time to restart the plant from cold 

shutdown. Removing oxygen in a more 

expeditious fashion enhances RCS chemistry. 

Based on the above, RCS integrity is 

maintained by this amendment. 

Therefore, this amendment 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? 

Response: No. 

Operation of HNP in accordance with the 

proposed amendment does not involve a 

significant reduction in a margin of safety. 

Existing TS operability and surveillance 

requirements are not reduced by the 

proposed amendment. The proposed 

amendment does not modify any plant 

equipment. In addition, the proposed 

amendment does not result in a change to a 

structure, system, or component (SSC), or its 

design function. The proposed amendment 

does not adversely affect existing plant safety 

margins or the reliability of equipment 

assumed to mitigate accidents in the FSAR. 

The purpose of the temperature limit for 

RCS oxygen control is to minimize corrosion 

at high temperatures on RCS components. 

Increasing the temperature at which oxygen 

levels are required to be maintained within 

specified limits from 180 °F to 250 °F is 

supported by industry and vendor data 

which indicates that the influence of 

dissolved oxygen at or below 250 °F is not 

significant with regard to stress corrosion 

cracking and general corrosion of RCS 

components. The proposed amendment is 

consistent with EPRI’s guidelines for PWR 

Primary Water Chemistry. This amendment 

places HNP in line with standard industry 

specifications for reactors of similar size and 

vintage. HNP’s proposed amendment to 

increase the temperature limit for 

applicability to 250 °F would decrease the 

time needed to achieve compliance with the 

dissolved oxygen limit and decrease the 

overall time to restart the plant from cold 

shutdown. Removing oxygen in a more 

expeditious fashion enhances RCS chemistry. 

Based on the above, RCS integrity is 

maintained by this amendment. 

Therefore, this amendment 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: David T. 

Conley, Associate General Counsel II—

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, North Carolina 27602. 

NRC Section Chief: Michael L. 

Marshall, Jr. 

Nebraska Public Power District, Docket 

No. 50–298, Cooper Nuclear Station, 

Nemaha County, Nebraska 

Date of amendment request: June 20, 

2005. 

Description of amendment request: 

The proposed amendment would revise 

Cooper Nuclear Station (CNS) Technical 

Specification (TS) 5.3, ‘‘Unit Staff 

Qualifications,’’ to upgrade the 

qualification standard for the Shift 

Manager, Senior Operator, Licensed 

Operator, and Shift Technical Engineer 

from Regulatory Guide (RG) 1.8, 

Revision 2 ‘‘Qualification and Training 

of Personnel for Nuclear Power Plants,’’

to RG 1.8, Revision 3. It also clarifies 

qualification requirements applicable to 

the Operations Manager position. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

These changes are administrative in nature 

and do not require any physical 

modifications, affect any plant components, 

or result in any changes in plant operation. 

They provide clarity and consistency to the 

CNS licensing basis. 

Upgrading the unit staff qualifications for 

the Shift Manager, Senior Operator, Licensed 

Operator, and Shift Technical Engineer from 

Regulatory Guide 1.8, Revision 2, to 

Regulatory Guide 1.8, Revision 3, is an 

administrative change that will clarify the 

current requirements for qualification and 

training of operations personnel. The changes 

are consistent with the application of a 

systems approach to training in an accredited 

training program. By promulgation of the 10 

CFR Part 55 rule change, the NRC determined 

that an accredited licensed operator training 

program based on a systems approach to 

training provides an acceptable means of 

qualifying licensed operating personnel. 

The addition of qualification requirements 

for the Operations Manager position clarifies 

SRO [Senior Reactor Operator] license 

requirements for Operations management 

personnel by specifying that the Operations 

Supervisor is the member of Operations 

management required to have a current SRO 

license at CNS. The Operations Manager is 

required to hold or have previously held a 

59086 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 

SRO license. This will ensure an acceptable 

level of operations knowledge to perform in 

a managerial oversight role. This approach is 

consistent with current guidance in ANSI/ 

ANS [American Nuclear Standards Institute/ 

American Nuclear Society] 3.1–1993. This 

change is administrative in nature and has no 

impact on previously evaluated accidents. 

Therefore, these changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

These changes are administrative in nature 

and do not involve a physical alteration of 

the plant or a change to plant operations. No 

new failure mechanisms, malfunctions, or 

accident initiators are introduced. The 

proposed changes provide clarity and 

consistency to the CNS licensing basis in 

regard to training and qualification of 

operations personnel and SRO license 

requirements for Operations management 

personnel. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

Response: No. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

These changes are administrative in nature 

and do not affect any Technical Specification 

safety limit or limiting condition for 

operation. No safety margins are affected by 

these changes. The proposed changes do not 

involve a change in plant design or operation 

for the mitigation of postulated accidents. 

The proposed changes provide clarity and 

consistency to the CNS licensing basis in 

regard to training and qualification of 

operations personnel and SRO license 

requirements for Operations management 

personnel. 

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: Mr. John C. 

McClure, Nebraska Public Power 

District, Post Office Box 499, Columbus, 

NE 68602–0499. 

NRC Section Chief: David Terao. 

Nebraska Public Power District, Docket 

No. 50–298, Cooper Nuclear Station, 

Nemaha County, Nebraska 

Date of amendment request: August 

25, 2005. 

Description of amendment request: 

The proposed amendment would revise 

the definitions of Channel Calibration, 

Channel Function Test, and Logic 

System Functional Test in accordance 

with the Technical Specification Task 

Force Traveler 205–A. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The definitions of Channel Calibration, 

Channel Functional Test, and Logic System 

Functional Test specified in Technical 

Specifications (TS) provide basic information 

regarding what the test involves, the 

components involved in the test, and general 

information regarding how the test is to be 

performed. These definitions and their 

specific wording are not precursors to any 

accident. As a result these revised definitions 

result in no increase in the probability of an 

accident previously evaluated. 

The proposed revisions of these definitions 

involve no changes to plant design, 

equipment, or operation related to mitigation 

of accidents. The proposed revisions of these 

definitions do not change their meaning or 

intent. The proposed revisions clarify the 

definitions and do not result in a reduction 

of required testing of instrumentation used to 

mitigate accidents. 

Based on the above NPPD [Nebraska Public 

Power District] concludes that the proposed 

changes do not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed revisions of the definitions 

do not involve a change to the design or 

operation of any plant structure, system, or 

component (SSC). As a result the plant will 

continue to be operated in the same manner. 

The proposed revisions will not result in a 

change to how the instrumentation used to 

monitor plant operation and to mitigate 

accidents is tested. Operating the plant and 

testing the plant’s instrumentation in the 

same manner as is currently done will not 

create the possibility of a new or different 

kind of accident. 

Based on the above NPPD concludes that 

the proposed changes do not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The affected definitions involve testing of 

instrumentation used in the mitigation of 

accidents to ensure that the instrumentation 

will perform as assumed in safety analyses. 

The proposed revisions of these definitions 

will not change their meaning or intent. As 

a result, the instrumentation will continue to 

be tested in the same manner as is currently 

done. Revising these definitions as proposed 

will not result in a change to the design or 

operation of any plant SSC used to shutdown 

the plant, initiate the Emergency Core 

Cooling Systems, or isolate primary or 

secondary containment. As a result the 

ability of the plant to respond to and mitigate 

accidents is unchanged by the revised 

definitions. 

Based on the above, NPPD concludes that 

the proposed changes do not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. John C. 

McClure, Nebraska Public Power 

District, Post Office Box 499, Columbus, 

NE 68602–0499. 

NRC Section Chief: David Terao. 

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: July 29, 

2005. 

Description of amendment requests: 

The proposed amendments would 

revise Technical Specification 3.7.5, 

‘‘Auxiliary Feedwater (AFW) System,’’

to change the frequency of Surveillance 

Requirement 3.7.5.6 from 92 days to 24 

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 proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change to increase [the] 

frequency interval for Surveillance 

Requirement (SR) 3.7.5.6 from 92 days to 24 

months has no impact on the probability of 

accidents previously evaluated. The valves 

controlled by SR 3.7.5.6 are used to provide 

an alternate supply of water to the auxiliary 

feedwater (AFW) system from the fire water 

storage tank (FWST) and are only operated 

after an accident has occurred. They are not 

accident initiators. 

Misoperation, or failure of a[n] FWST 

supply to be correctly positioned following 

an accident, could result in an inadequate 

supply of water to the AFW system. Failure 

to provide adequate core cooling could 

increase the radiological consequences of an 

accident. However, operating and 

maintenance histories of the FWST supply 

valves show that these valves have been 

Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 59087 

capable of full stroke cycling each time they 

have been tested. There is no evidence of any 

time-related degradation mechanism that 

would prevent the valves from performing 

their design function. Thus[,] the proposed 

change has no impact on the consequences 

of an accident. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different [kind of] 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change to increase frequency 

interval for SR 3.7.5.6 from 92 days to 24 

months has no impact on the probability of 

accidents of the type evaluated in the Final 

Safety Analysis Report, as updated. The 

valves are used to provide an alternate 

supply of water to the AFW system from the 

FWST, and are only operated after an 

accident has occurred. They are not accident 

initiators. Review of the operating and 

maintenance histories of the FWST supply 

valves show that they are highly reliable in 

maintaining their capability to perform their 

design function. 

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? 

Response: No. 

The proposed change to SR 3.7.5.6 

involves only an increase in the frequency 

interval. No physical changes are required to 

the facility or to the plant operating or 

emergency procedures as a result of the 

change. Based on review of the operating and 

maintenance histories of the FWST supply 

valves, they have been capable of full stroke 

cycling each time they have been tested. 

There is no evidence of any time-related 

degradation mechanism that would prevent 

the valves from performing their design 

function. This evidence supports the 

conclusion that there will be no impact in the 

operation of these valves following an 

accident. 

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 

requests involve no significant hazards 

consideration. 

Attorney for licensee: Richard F. 

Locke, Esq., Pacific Gas and Electric 

Company, P.O. Box 7442, San 

Francisco, California 94120. 

NRC Section Chief: Daniel S. Collins 

(Acting). 

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: August 

23, 2005. 

Description of amendment requests: 

The proposed amendments would 

revise the expiration dates of the Units 

1 and 2 facility-operating licenses to 

recapture low-power testing time, and to 

reflect a 40-year term measured from the 

date of issuance of each unit’s fullpower operating license. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed additional operating license 

periods do not affect the probability or 

consequences of an accident previously 

evaluated since they require no physical 

change in the plant equipment or operating 

procedures and the Final Safety Analysis 

Report (FSAR) Update safety analyses are 

based on [a] 40-year full[-]power operation. 

Surveillance and maintenance practices, as 

well as other programs such as 

environmental qualification of equipment, 

ensure timely identification and correction of 

any degradation of safety-related plant 

equipment. The long-term integrity of the 

reactor vessels has been evaluated using 

currently acceptable NRC calculational 

methods and best available Diablo Canyon 

Power Plant (DCPP) specific data. The 

evaluation results demonstrate that both 

reactor vessels are safe for normal operations 

in excess of 40 years. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different [kind of] 

accident from any accident previously 

evaluated? 

Response: No. 

The possibility of a new or different kind 

of accident is not created by the proposed 

additional operating periods since at least 40 

years of full[-]power operation was assumed 

in the design and construction of DCPP Units 

1 and 2. The plant maintenance programs are 

also designed to both maintain and 

determine the need to replace safety-related 

components. These programs will continue 

to be applied as they are presently to assure 

safe operation. 

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? 

Response: No. 

The proposed additional operating periods 

do not involve a significant reduction in a 

margin of safety since, as is the case with 

present operation, degradation of safetyrelated equipment will be identified and 

corrected by ongoing surveillance and 

maintenance practices. Existing programs, 

routine maintenance, and compliance with 

Technical Specifications assure that an 

adequate margin of safety is maintained. 

These activities will remain in effect for the 

duration of the proposed additional operating 

periods. 

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 requests involve no 

significant hazards consideration. 

Attorney for licensee: Richard F. 

Locke, Esq., Pacific Gas and Electric 

Company, P.O. Box 7442, San 

Francisco, California 94120. 

NRC Section Chief: Daniel S. Collins 

(Acting). 

South Carolina Electric & Gas Company, 

South Carolina Public Service 

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

Summer Nuclear Station, Unit No. 1, 

Fairfield County, South Carolina 

Date of amendment request: June 30, 

2005. 

Description of amendment request: 

The proposed changes would revise the 

Administrative Control section of the 

Technical Specifications (TSs) to permit 

the Westinghouse best estimate 

methodology for loss-of-coolantaccident (LOCA) analysis methodology 

to be utilized for analyses as required by 

Title 10 of the Code of Federal 

Regulations, Part 50, Section 46, 

‘‘Acceptance criteria for emergency core 

cooling systems [ECCS] for light water 

nuclear power reactors’ (10 CFR 50.46). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

Implementation of the best-estimate large 

break LOCA methodology and associated TS 

changes is proposed to increase margin to the 

peak clad temperature limits defined in 10 

CFR 50.46. There are no physical plant 

changes or changes in manner in which the 

plant will be operated as a result of this 

59088 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 

change. Since the plant conditions and ECCS 

performance assumed in the analysis are 

consistent with the plant’s current design, 

the proposed change in methodology will 

thus have no impact on the probability of a 

LOCA. When applied, the best estimate 

methodology shows that the ECCS is more 

effective than previously evaluated in 

mitigating the consequences of a LOCA, as 

lower peak clad temperatures are predicted 

relative to current 10 CFR 50.46 Appendix K 

results. Since the proposed best-estimate 

methodology is only applicable to a large 

break LOCA and since the application of the 

proposed methodology shows there is a high 

probability that all of the acceptance criteria 

contained in 10 CFR 50.46, Paragraph b are 

met, the proposed change does not increase 

the consequences of an accident previously 

evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

There are no physical changes being made 

to the plant. No new modes of plant 

operation are being introduced. The 

parameters assumed in the analysis remain 

within the design limits of the existing plant 

equipment. All plant systems will perform as 

designed during the response to a potential 

accident. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

analyzed. 

3. Does this change involve a significant 

reduction in a margin of safety? 

Response: No. 

It has been shown that the methodology 

used in the analysis would more realistically 

describe the expected behavior of V. C. 

Summer Nuclear Station systems during a 

postulated loss of coolant accident. 

Uncertainties have been accounted for as 

required by 10 CFR 50.46. A sufficient 

number of loss of coolant accidents with 

different break sizes, different locations and 

other variations in properties are analyzed to 

provide assurance that the most severe 

postulated loss of coolant accidents are 

calculated. It has been shown by analysis that 

there is a high level of probability that all 

criteria contained in 10 CFR 50.46, Paragraph 

b are met. 

Pursuant to 10 CFR 50.91, the preceding 

analyses provide a determination that the 

proposed Technical Specifications change 

poses no significant hazard as delineated by 

10 CFR 50.92. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92 (c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Thomas G. 

Eppink, South Carolina Electric & Gas 

Company, Post Office Box 764, 

Columbia, South Carolina 29218. 

NRC Section Chief: Evangelos C. 

Marinos. 

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: August 

30, 2005. 

Description of amendment request: 

The proposed amendment would 

change the Technical Specifications 

(TSs) to reflect the use of the 

Westinghouse Best Estimate Analyzer 

for Core Operations—Nuclear 

(BEACON) to augment the functional 

capability of the flux mapping system 

for the purpose of power distribution 

surveillances. In addition, editorial 

changes to the TSs are proposed. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The PDMS [power distribution monitoring 

system] performs continuous core power 

distribution monitoring. This system utilizes 

the NRC-approved Westinghouse proprietary 

computer code BEACON to provide data 

reduction for incore flux maps, core 

parameter analysis, load follow operation 

simulation, and core prediction. It in no way 

provides any protection or control system 

function. Fission product barriers are not 

impacted by these proposed changes. The 

proposed changes occurring with PDMS will 

not result in any additional challenges to 

plant equipment that could increase the 

probability of any previously evaluated 

accident. The changes associated with the 

PDMS do not affect plant systems such that 

their function in the control of radiological 

consequences is adversely affected. These 

proposed changes will therefore not affect the 

mitigation of the radiological consequences 

of any accident described in the Updated 

Final Safety Analysis Report Update 

(UFSAR). 

Continuous on-line monitoring through the 

use of PDMS provides significantly more 

information about the power distributions 

present in the core than is currently 

available. This results in more time (i.e., 

earlier determination of an adverse condition 

developing) for operator action prior to 

having an adverse condition develop that 

could lead to an accident condition or to 

unfavorable initial conditions for an 

accident. 

Each accident analysis addressed in the 

UFSAR is examined with respect to changes 

in cycle-dependent parameters, which are 

obtained from application of the NRCapproved reload design methodologies, to 

ensure that the transient evaluations of 

reload cores are bounded by previously 

accepted analyses. This examination, which 

is performed in accordance with the 

requirements set forth in 10 CFR [Title 10 of 

the Code of Federal Regulations] 50.59, 

ensures that future reloads will not involve 

a significant increase in the probability or 

consequences of any accident previously 

evaluated. 

The three editorial changes only correct 

typographical errors made in previously 

approved TS changes. They do not affect 

plant operation or structures, systems, and 

components important to safety. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequence of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The implementation of the PDMS has no 

influence or impact on plant operations or 

safety, nor does it contribute in any way to 

the probability or consequences of an 

accident. No safety-related equipment, safety 

function, or plant operation will be altered as 

a result of this proposed change. The 

possibility for a new or different type of 

accident from any accident previously 

evaluated is not created since the changes 

associated with implementation of the PDMS 

do not result in a change to the design basis 

of any plant component or system. The 

evaluation of the effects of using the PDMS 

to monitor core power distribution 

parameters shows that all design standards 

and applicable safety criteria limits are met. 

The proposed changes do not result in any 

event previously deemed incredible being 

made credible. Implementation of the PDMS 

will not result in more adverse conditions 

and will not result in any increase in the 

challenges to safety systems. The cyclespecific variables required by the PDMS are 

calculated using NRC-approved methods. 

The TS will continue to require operation 

within the required core operating limits and 

appropriate actions will be taken if limits are 

exceeded. 

The three editorial changes only correct 

typographical errors made in previously 

approved TS changes. They do not affect 

plant operation or structures, systems, and 

components important to safety. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The margin of safety is not affected by 

implementation of the PDMS. The margin of 

safety provided by current TS is unchanged. 

The proposed changes continue to require 

operation within the core limits that are 

based on NRC-approved reload design 

methodologies. Appropriate measures exist 

to control the values of these cycle-specific 

limits. The proposed changes continue to 

ensure that appropriate actions will be taken 

Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 59089 

if limits are violated. These actions remain 

unchanged. 

The three editorial changes only correct 

typographical errors made in previously 

approved TS changes. They do not affect 

plant operation or structures, systems, and 

components important to safety. 

Therefore, the proposed changes do 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the 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: A. H. 

Gutterman, Esq., Morgan, Lewis & 

Bockius, 1111 Pennsylvania Avenue, 

NW., Washington, DC 20004. 

NRC Section Chief: David Terao. 

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 (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), 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/ 

reading-rm/adams.html. If you do not 

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail 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: 

April 3, 2003, as supplemented 

December 23, 2003, December 9 and 17, 

2004, and March 30 and August 19, 

2005. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications (TSs) to support the 

application of an alternative source term 

methodology in accordance with Title 

10 of the Code of Federal Regulations, 

Section 50.67, ‘‘Accident Source Term,’’

with the exception that Technical 

Information Document 14844, 

‘‘Calculation of Distance Factors for 

Power and Test Reactor Sites,’’ was used 

as the radiation dose basis for 

equipment qualification. 

Date of issuance: September 19, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 167. 

Facility Operating License No. NPF–

62: The amendment revised the TSs. 

Date of initial notice in Federal 

Register: September 2, 2003 (68 FR 

52234). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 19, 

2005. 

The supplements dated December 23, 

2003, December 9 and 17, 2004, and 

March 30 and August 19, 2005 provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. 

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: 

November 11, 2003, as supplemented 

April 16 and September 10, 2004, and 

March 30 and September 21, 2005. 

Brief description of amendment: The 

amendment revised the instrument 

channel trip setpoint allowable values 

for thirteen Technical Specification (TS) 

functions at Clinton Power Station, Unit 

1. 

Date of issuance: September 27, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 168. 

Facility Operating License No. NPF–

62: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: March 16, 2004 (69 FR 

12363). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 21, 

2005. The supplements dated April 16 

and September 10, 2004, and March 30 

and September 21, 2005, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. No 

significant hazards consideration 

comments received: No. 

Calvert Cliffs Nuclear Power Plant, Inc., 

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

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of application for amendments: 

August 3, 2004, as supplemented on 

July 8 and August 26, 2005. 

Brief description of amendments: The 

amendments extend the surveillance 

frequency interval from monthly to 

quarterly for Technical Specification 

surveillance requirement (SR) 3.3.3.1, 

which involves a channel functional test 

of each reactor trip circuit breaker 

(RTCB). SRs 3.3.3.1 and 3.3.3.2 will be 

scheduled such that the RTCBs testing 

is performed every 6 weeks, which 

meets the vendor-recommended interval 

for cycling each RTCB. 

Date of issuance: September 26, 2005. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment Nos.: 275 and 252. 

Renewed Facility Operating License 

Nos. DPR–53 and DPR–69: Amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: January 4, 2005 (70 FR 400). 

59090 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 

The July 8 and August 26, 2005, 

supplemental letters provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of these amendments is contained in a 

Safety Evaluation dated September 26, 

2005. 

No significant hazards consideration 

comments received: No. 

Dominion Nuclear Connecticut, Inc., 

Docket No. 50–245, Millstone Power 

Station Unit No. 1, New London County, 

Connecticut 

Date of application for amendment: 

September 8, 2004, as supplemented by 

letters dated May 5 and July 27, 2005. 

Brief description of amendment: The 

amendment revised the Millstone Power 

Station, Unit No. 1 Technical 

Specifications (TSs) to support the 

implementation of the proposed 

Dominion Nuclear Facility Quality 

Assurance Program (Topical Report 

DOM–QA–1). Implementation of this 

Topical Report would create a common 

quality assurance program for all sites 

owned by Dominion Nuclear 

Connecticut, Inc. Review of this 

proposed amendment was requested in 

concert with the review of the Topical 

Report. 

Date of issuance: September 15, 2005. 

Effective date: As of the date of 

issuance, and shall be implemented by 

February 28, 2006. 

Amendment No.: 115. 

Facility Operating License No. DPR–

21: The amendment revised the TSs. 

Date of initial notice in Federal 

Register: January 18, 2005 (70 FR 2888). 

The additional information provided 

in the supplemental letters dated May 5 

and July 27, 2005, did not expand the 

scope of the application as noticed and 

did not change the NRC staff’s original 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 15, 

2005. 

No significant hazards consideration 

comments received: No. 

Dominion Nuclear Connecticut, Inc., 

Docket No. 50–336, Millstone Power 

Station, Unit No. 2, New London 

County, Connecticut 

Date of application for amendment: 

July 15, 2004, as supplemented by letter 

dated August 23, 2004. 

Brief description of amendment: The 

amendment revised the Facility 

Operating License DPR–65 to address 

the resolution of a non-conservative 

Technical Specifications (TSs) 

associated with control room isolation 

radiation monitoring instrumentation. 

Specifically, the amendment would 

revise the TSs to require two operable 

channels of control room isolation 

radiation monitoring instrumentation. 

Date of issuance: September 23, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment No.: 289. 

Facility Operating License No. DPR–

65: The amendment revised the TSs. 

Date of initial notice in Federal 

Register: January 18, 2005 (70 FR 2887). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 23, 

2005. 

No significant hazards consideration 

comments received: No. 

Dominion Nuclear Connecticut, Inc., et 

al., Docket No. 50–423, Millstone Power 

Station, Unit No. 3, New London 

County, Connecticut 

Date of application for amendment: 

April 15, 2004, as supplemented on 

June 23, 2005. 

Brief description of amendment: The 

amendment approves modifications to 

the Fire Protection Program. 

Specifically, the modifications involve 

converting the existing automatic 

carbon dioxide fire suppression systems 

installed in the cable spreading room to 

manual actuation. 

Date of issuance: September 22, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 227. 

Facility Operating License No. NPF–

49: The amendment allows for 

conversion from an automatic to a 

manual carbon dioxide suppression 

system in the cable spreading area. 

Date of initial notice in Federal 

Register: July 6, 2004 (69 FR 40672). 

The supplement dated June 23, 2005, 

provided clarifying information and did 

not change the initial proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 22, 

2005. 

No significant hazards consideration 

comments received: No. 

Dominion Nuclear Connecticut, Inc., 

Docket Nos. 50–336 and 50–423, 

Millstone Power Station, Unit Nos. 2 

and 3, New London County, Connecticut 

Date of application for amendments: 

September 8, 2004, as supplemented by 

letters dated May 5 and July 27, 2005. 

Brief description of amendments: The 

amendments revised the Millstone 

Power Station, Unit Nos. 2 and 3 

Technical Specifications (TSs) to 

support the implementation of the 

proposed Dominion Nuclear Facility 

Quality Assurance Program (Topical 

Report DOM–QA–1). Implementation of 

this Topical Report would create a 

common quality assurance program for 

all sites owned by Dominion Nuclear 

Connecticut, Inc. Review of these 

proposed amendments was requested to 

be done in concert with the review of 

the Topical Report. 

Date of issuance: September 15, 2005. 

Effective date: As of the date of 

issuance, and shall be implemented by 

February 28, 2006. 

Amendment Nos.: 288 and 226. 

Facility Operating License Nos. DPR–

65 and NPF–49: The amendments 

revised the TSs. 

Date of initial notice in Federal 

Register: January 18, 2005 (70 FR 

2888). The additional information 

provided in the supplemental letters 

dated May 5, and July 27, 2005, did not 

expand the scope of the application as 

noticed and did not change the NRC 

staff’s original proposed no significant 

hazards consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 15, 

2005. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of application for amendment: 

October 6, 2004, as supplemented on 

February 16, and August 9, 2005. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) surveillance 

requirement 4.5.B.1 related to air testing 

of the drywell spray headers and 

nozzles. Specifically, the amendment 

changes the test frequency from once 

every five years to following 

maintenance that could result in nozzle 

blockage. 

Date of Issuance: September 20, 2005. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 59091 

Amendment No.: 228. 

Facility Operating License No. DPR–

28: The amendment revised the TSs. 

Date of initial notice in Federal 

Register: December 21, 2004 (69 FR 

76492). The supplements contained 

clarifying information only, and did not 

change the initial no significant hazards 

consideration determination or expand 

the scope of the initial Federal Register 

notice. 

The Commission’s related evaluation 

of this amendment is contained in a 

Safety Evaluation dated September 20, 

2005. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50–

313, Arkansas Nuclear One, Unit No. 1, 

Pope County, Arkansas 

Date of amendment request: 

September 30, 2004, as supplemented 

by letter dated May 20, 2005. 

Brief description of amendment: The 

amendment revises the Technical 

Specifications to allow the use of M5 

fuel cladding and of Mark-B-high 

thermal performance fuel in Arkansas 

Nuclear One, Unit 1, during its fuel 

Cycle 20 and beyond. 

Date of issuance: September 12, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment No.: 226. 

Renewed Facility Operating License 

No. DPR–51: Amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: November 9, 2004 (69 FR 

64988). The supplement dated May 20, 

2005, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 12, 

2005. 

No significant hazards consideration 

comments received: No. 

Exelon Generating Company, LLC, 

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Units 1 and 2, 

Will County, Illinois 

Date of application for amendment: 

December 17, 2004. 

Brief description of amendment: The 

amendments revised Appendix B, 

Environmental Protection Plan (nonradiological), of the Braidwood Station 

Facility Operating Licenses. 

Date of issuance: September 19, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 138. 

Facility Operating License Nos. NPF–

72 and NPF–77: The amendments 

revised the Environmental Protection 

Plan. 

Date of initial notice in Federal 

Register: April 12, 2005 (70 FR 19115). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 19, 

2005. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket No. STN 50–455, Byron Station, 

Unit No. 2, Ogle County, Illinois 

Date of application for amendment: 

May 24, 2005. 

Brief description of amendment: The 

amendment modifies the inspection 

requirements for portions of the steam 

generator (SG) tubes within the hot leg 

tubesheet region of the SGs. 

Date of issuance: September 19, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 144. 

Facility Operating License No. NPF–

66: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: July 5, 2005 (70 FR 38718). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 19, 

2005. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Units 1 

and 2, Rock Island County, Illinois 

Date of application for amendments: 

February 27, 2004, as supplemented by 

letters dated October 11, 2004, January 

3, 2005, August 11, 2005, and 

September 12, 2005. 

Brief description of amendments: The 

amendments add the Oscillation Power 

Range Monitor (OPRM) instrumentation 

to the Technical Specifications. 

Date of issuance: September 22, 2005. 

Effective date: As of the date of 

issuance and shall be implemented by 

December 31, 2005. 

Amendment Nos.: 227, 222. 

Facility Operating License Nos. DPR–

19, DPR–25, DPR–29 and DPR–30. The 

amendments revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: December 7, 2004 (69 FR 

70718). The October 11, 2004, and 

January 3, 2005, August 11, 2005, and 

September 12, 2005, submittals 

provided clarifying information that did 

not change the initial proposed no 

significant hazards consideration. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated: September 22, 

2005. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–352 and 50–353, 

Limerick Generating Station, Units 1 

and 2, Montgomery County, 

Pennsylvania 

Date of application for amendments: 

July 22, 2004, as supplemented 

December 3, 2004, and September 20, 

2005. The September 20, 2005, 

supplement withdrew a portion of the 

original application from consideration. 

Brief description of amendments: The 

amendments modified the operability 

and surveillance requirements in 

Technical Specification (TS) 3/4.1.3, 

‘‘Control Rods.’’ Specifically, the 

changes (1) exclude a fully-inserted 

immovable control rod from the 

shutdown action statement, and (2) 

limit the 24-hour exercise test of other 

control rods to a one-time occasion 

following detection of an immovable 

control rod. 

Date of issuance: September 27, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 178 and 140. 

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments 

revised the TSs. 

Date of initial notice in Federal 

Register: May 24, 2005 (70 FR 29794). 

The September 20, 2005, supplement 

withdrew a portion of the original 

application from consideration and did 

not change the proposed no significant 

hazards consideration. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 27, 

2005. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–352 and 50–353, 

Limerick Generating Station, Units 1 

and 2, Montgomery County, 

Pennsylvania 

Date of application for amendments: 

June 1, 2004. 

59092 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 

Brief description of amendments: The 

amendments relocate the operability 

and surveillance requirements for the 

reactor coolant system safety/relief 

valve position instrumentation from the 

Limerick Generating Station (LGS) 

Technical Specifications (TSs) to the 

LGS Technical Requirements Manual 

(TRM) and plant procedures. 

Specifically, the amendments relocate 

TSs 3.4.2.c, 4.4.2.1, and the associated 

footnotes to the TRM. Additionally, the 

‘‘Safety/Relief Valve Position 

Indicators’’ instrumentation is relocated 

from Tables 3.3.7.5–1 and 4.3.7.5–1 of 

TSs 3.3.7.5 and 4.3.7.5, respectively, to 

the TRM. 

Date of issuance: September 27, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 179 and 141. 

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments 

revised the TSs. 

Date of initial notice in Federal 

Register: October 26, 2004 (69 FR 

62475). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 27, 

2005. 

No significant hazards consideration 

comments received: No. 

FirstEnergy Nuclear Operating 

Company, et al., Docket Nos. 50–334 

and 50–412, Beaver Valley Power 

Station, Unit Nos. 1 and 2 (BVPS–1 and 

2), Beaver County, Pennsylvania 

Date of application for amendments: 

June 2, 2004, as supplemented February 

23 and August 19, 2005. 

Brief description of amendments: The 

amendments revised the BVPS–1 and 2, 

Technical Specifications (TSs) 3/4 3.1, 

‘‘Reactor Trip System (RTS) 

Instrument,’’ and 3/4 3.2, ‘‘Engineered 

Safety Features Actuation System 

(ESFAS) Instrument,’’ to increase the 

surveillance interval from monthly to 

quarterly for certain RTS and ESFAS 

instrument channel functional tests. 

Date of issuance: September 19, 2005. 

Effective date: September 19, 2005. 

Amendment Nos.: 267 and 149. 

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised 

the TSs. 

Date of initial notice in Federal 

Register: July 6, 2004 (69 FR 40674). 

The supplements dated February 23 

and August 19, 2005, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the Nuclear 

Regulatory Commission (NRC) staff’s 

original proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 19, 

2005. 

No significant hazards consideration 

comments received: No. 

FirstEnergy Nuclear Operating 

Company, et al., Docket Nos. 50–334 

and 50–412, Beaver Valley Power 

Station, Unit Nos. 1 and 2 (BVPS–1 and 

2), Beaver County, Pennsylvania 

Date of application for amendments: 

May 26, 2004, as supplemented by 

letters dated October 29 and December 

3, 2004, and January 18, June 15, and 

August 15, 2005. 

Brief description of amendments: The 

amendments extended the allowable 

outage time for the BVPS–1 and 2 

emergency diesel generators (EDGs) 

from 72 hours to 14 days. The 

amendments also deleted surveillance 

requirement (SR) 4.8.1.1.2.b.1 

concerning periodic EDG inspections. 

Requirements for periodic EDG 

inspections will be specified in a 

licensee-controlled EDG maintenance 

program referenced in the Updated 

Final Safety Analysis Report. The 

amendments also revised footnote (1) of 

TS 3.8.1.1 to clarify the wording to 

allow actions to be delayed for up to 7 

days to allow time to restore fuel oil 

back to its specified limits when an EDG 

is inoperable solely due to failure to 

meet fuel oil property limits of SR 

4.8.1.1.2.d.2 or SR 4.8.1.1.2.e. 

Date of issuance: September 29, 2005. 

Effective date: Upon issuance to be 

implemented within 60 days. The 

implementation shall include the 

commitments as described in the 

licensee’s submittals dated May 26 and 

December 3, 2004, and January 18 and 

June 15, 2005, and as described in the 

NRC staff’s safety evaluation related to 

this amendment. 

Amendment Nos.: 268 and 150. 

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: July 6, 2004 (69 FR 40673). 

The supplements dated October 29 

and December 3, 2004, and January 18, 

June 15, and August 15, 2005, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 29, 

2005. 

No significant hazards consideration 

comments received: No. 

Florida Power and Light Company, 

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

No. 1, St. Lucie County, Florida 

Date of application for amendment: 

December 20, 2004. 

Brief description of amendment: This 

amendment revises Technical 

Specifications Figures 3.1–1b, 3.4–2a, 

3.4–2b and 3.4–3 to reflect an extension 

in the effectiveness of the pressure/ 

temperature (P/T) limit curves from 23.6 

to 35 effective full power years (EFPY). 

The low temperature overpressure 

protection requirements, which are 

based on the P/T limits, are also 

extended to 35 EFPY. 

Date of Issuance: September 21, 2005. 

Effective Date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No.: 196. 

Renewed Facility Operating License 

No. DPR–67: Amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: March 1, 2005 (70 FR 9993). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 21, 

2005. 

No significant hazards consideration 

comments received: No. 

Florida Power and Light Company, et 

al., Docket Nos. 50–335 and 50–389, St. 

Lucie Plant, Unit Nos. 1 and 2, St. Lucie 

County, Florida 

Date of application for amendments: 

September 18, 2003, as supplemented 

on August 25 and September 15, 2005. 

Brief description of amendments: The 

amendments revise Technical 

Specifications (TSs) for the control room 

ventilation systems to model the 

Combustion Engineering Standard 

Technical Specifications, NUREG–1432. 

In addition, Table 3.3–6, Radiation 

Monitoring Instrumentation, in each 

unit’s TSs is revised to resolve minor 

inconsistencies that resulted from 

changes associated with previously 

issued Amendments 184 (Unit 1) and 

127 (Unit 2). The amendments also 

correct some minor typographical 

errors. 

Date of Issuance: September 27, 2005. 

Effective Date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment Nos.: 197 and 139. 

Renewed Facility Operating License 

Nos. DPR–67 and NPF–16: Amendments 

revised the TSs. 

Date of initial notice in Federal 

Register: October 28, 2003 (68 FR 

61478). The August 25 and September 

15, 2005, supplements did not affect the 

original proposed no significant hazards 

Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 59093 

determination, or expand the scope of 

the request as noticed in the Federal 

Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 27, 

2005. 

No significant hazards consideration 

comments received: No. 

Florida Power Corporation, et al., 

Docket No. 50–302, Crystal River Unit 

No. 3 Nuclear Generating Plant, Citrus 

County, Florida 

Date of application for amendment: 

January 13, 2005, as supplemented by 

letters dated February 11, May 6, and 

June 9, 2005. 

Brief description of amendment: The 

amendment allows a one-time extended 

allowed outage time (AOT) change to 

Improved Technical Specifications (ITS) 

3.5.2, Emergency Core Cooling Systems 

(ECCS)—Operating; 3.6.6, Reactor 

Building Spray and Containment 

Cooling Systems; 3.7.8, Decay Heat 

Closed Cycle Cooling Water System 

(DC); and 3.7.10, Decay Heat Seawater 

System to allow the refurbishment of 

Decay Heat Seawater System Pump 

RWP–3B online. 

Date of issuance: September 15, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No.: 221. 

Facility Operating License No. 

DPR–72: Amendment revises the 

Technical Specifications. 

Date of initial notice in Federal 

Register: February 1, 2005 (70 FR 

5246). The February 11, May 6, and June 

9, 2005, supplements contained 

clarifying information only and did not 

change the initial no significant hazards 

consideration determination or expand 

the scope of the initial application. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 15, 

2005. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket Nos. 50–266 and 50–1, Point 

Beach Nuclear Plant, Units 1 and 2, 

Town of Two Creeks, Manitowoc 

County, Wisconsin 

Date of application for amendments: 

July 24, 2005. 

Brief description of amendments: The 

amendments incorporated a Point Beach 

Nuclear Plant (PBNP), Unit 1 reactor 

vessel head (RVH) drop accident 

analysis into the PBNP Final Safety 

Analysis Report and revised the PBNP, 

Unit 2 RVH drop accident analysis. 

Date of issuance: September 23, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 220, 226. 

Facility Operating License Nos. DPR–

24 and DPR–27: Amendments revised 

the License. 

Date of initial notice in Federal 

Register: August 16, 2005 (70 FR 

48198). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 23, 

2005. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket Nos. 50–266 and 50–301, Point 

Beach Nuclear Plant, Units 1 and 2, 

Town of Two Creeks, Manitowoc 

County, Wisconsin 

Date of amendment request: April 8, 

2004, as supplemented by letters dated 

November 15, 2004, July 15 and August 

8, 2005. 

Description of amendment request: 

The amendments revised technical 

specification surveillance requirements 

(SR) 3.8.4.6 and SR 3.8.4.7, ‘‘DC 

Sources—Operating.’’ Specifically, the 

amendments revised battery charger 

current values, added a new allowance 

for verifying battery charger capacity, 

and removed a restriction on the 

conduct of a modified performance 

discharge test. 

Date of issuance: September 27, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 45 days. 

Amendment Nos.: 221, 227. 

Facility Operating License Nos. DPR–

24 and DPR–27: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: August 19, 2004 (69 FR 

51489). The November 15, 2004, July 15 

and August 8, 2005, supplemental 

letters provided additional information 

that clarified the application, did not 

expand the scope of the application 

originally noticed, and did not change 

the NRC staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 27, 

2005. 

No significant hazards consideration 

comments received: No. 

R.E. Ginna Nuclear Power Plant, LLC, 

Docket No. 50–244, R.E. Ginna Nuclear 

Power Plant, Wayne County, New York 

Date of application for amendment: 

September 30, 2004, and May 28, 2005. 

Brief description of amendment: The 

amendment revises information in the 

Updated Final Safety Analysis Report 

(UFSAR) regarding the application of 

‘‘leak-before-break’’ methodology for the 

emergency core cooling system 

accumulator lines A and B and the 

pressurizer surge line. The amendment 

permits the exclusion of these lines 

from the evaluation of the dynamic 

effects associated with postulated highenergy line breaks in the analyzed 

segments of the accumulator lines 

piping system and the pressurizer surge 

line piping system. 

Date of issuance: September 22, 2005. 

Effective date: As of the date of 

issuance and shall be implemented with 

the next update of the UFSAR in 

accordance with 10 CFR 50.71(e). 

Amendment No.: 92. 

Renewed Facility Operating License 

No. DPR–18: Amendment revised the 

UFSAR. 

Date of initial notice in Federal 

Register: July 5, 2005 (70 FR 38721). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 22, 

2005. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–424 and 50–425, 

Vogtle Electric Generating Plant, Units 1 

and 2, Burke County, Georgia 

Date of application for amendments: 

August 12, 2005, as supplemented by 

letter dated August 24, 2005. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications to incorporate changes in 

the steam generator (SG) inspection 

scope for Vogtle, Unit 2 during 

Refueling Outage 11 and the subsequent 

operating cycle. The proposed changes 

modify the inspection requirements for 

portions of SG tubes within the hot leg 

tubesheet region of the SGs. The license 

for Vogtle, Unit 1 is affected only due 

to the fact that Unit 1 and Unit 2 use 

common Technical Specifications. 

Date of issuance: September 21, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 138/117. 

Facility Operating License Nos. 

NPF–68 and NPF–81: Amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: August 22, 2005 (70 FR 

48985). 

The supplement dated August 24, 

2005, provided clarifying information 

that did not change the scope of the 

August 12, 2005, application nor the 

59094 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 

initial proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 21, 

2005. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–424 and 50–425, 

Vogtle Electric Generating Plant, Units 1 

and 2, Burke County, Georgia 

Date of application for amendments: 

August 13, 2004, as supplemented by 

letters dated May 3 and July 7, 2005. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications (TSs) to reflect updated 

spent fuel rack criticality analyses for 

Units 1 and 2. The amendments also 

corrected a typographical error on Page 

vi of the TSs Table of Contents 

associated with the issuance of 

Amendments 130 and 109, for Units 1 

and 2 TSs, respectively. 

Date of issuance: September 22, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: 139/118. 

Facility Operating License Nos. 

NPF–68 and NPF–81: Amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: November 9, 2004 (69 FR 

64990). 

The supplements dated May 3 and 

July 7, 2005, provided clarifying 

information that did not change the 

scope of the August 13, 2004, 

application nor the initial proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 22, 

2005. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket No. 

50–390, Watts Bar Nuclear Plant, Unit 1, 

Rhea County, Tennessee 

Date of application for amendment: 

November 21, 2003, as supplemented by 

letters dated May 5 and August 19, 

2004, and July 11, 2005. 

Brief description of amendment: The 

amendment allows the position of the 

control and shutdown rods to be 

monitored by a means other than the 

movable incore detectors. 

Date of issuance: September 20, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 58. 

Facility Operating License No. NPF–

90: Amendment revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: December 23, 2003 (68 FR 

74267). The supplemental letters 

provided clarifying information that was 

within the scope of the initial notice 

and did not change the initial proposed 

no significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 20, 

2005. 

No significant hazards consideration 

comments received: No. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Unit Nos. 

1 and 2, Somervell County, Texas 

Date of amendment request: March 

24, 2005. 

Brief description of amendments: The 

amendments revise Technical 

Specification (TS) 3.3.1 entitled 

‘‘Reactor Trip System (RTS) 

Instrumentation’’ and TS 3.3.2 entitled 

‘‘Engineered Safety Feature Actuation 

System (ESFAS) Instrumentation’’, and 

Required Action Notes in the TSs to 

reflect wording in the Commissions 

Standard TSs incorporating the channel 

bypass capabilities as discussed in TS 

Task Force Traveler 418, Revision 2. 

Date of issuance: September 29, 2005. 

Effective date: Effective as of the date 

of issuance and shall be implemented in 

90 days from the date of issuance. 

Amendment Nos.: 121 and 121. 

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: April 26, 2005 (70 FR 21464). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 29, 

2005. 

No significant hazards consideration 

comments received: No. 

Virginia Electric and Power Company, 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Units 1 and 2, 

Louisa County, Virginia 

Date of application for amendment: 

September 15, 2004, as supplemented 

by letter dated May 5, 2005. 

Brief description of amendment: 

These amendments revise the Technical 

Specifications for North Anna Power 

Station, Units 1 and 2 to support the 

implementation of the proposed Topical 

Report DOM–QA–1, ‘‘Dominion Nuclear 

Facility Quality Assurance Program 

Description.’’ The implementation of 

this topical report would create a 

common quality assurance program for 

North Anna, Surry, and Millstone Power 

Stations. The review of these proposed 

amendments was requested to be done 

in concert with the review of the 

Topical Report. The Topical Report was 

submitted to the NRC staff for review on 

August 24, 2004, and supplemented by 

letter dated May 5, 2005. By letter dated 

September 9, 2005, the NRC staff 

approved of Topical Report DOM–QA–

1. 

Date of issuance: September 15, 2005. 

Effective date: As of the date of 

issuance and shall be implemented 

within 6 months from the date of 

issuance. 

Amendment Nos.: 243 and 224. 

Renewed Facility Operating License 

Nos. NPF–4 and NPF–7: Amendments 

change the Technical Specifications. 

Date of initial notice in Federal 

Register: November 23, 2004 (69 FR 

68187). The supplement dated May 5, 

2005, contained clarifying information 

only and did not change the initial no 

significant hazards consideration 

determination or expand the scope of 

the initial application. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 15, 

2005. 

No significant hazards consideration 

comments received: No. 

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: 

September 15, 2004, as supplemented 

by letter dated May 5, 2005. 

Brief description of amendments: 

These amendments revise the Technical 

Specifications for Surry Power Station, 

Units 1 and 2 to support the 

implementation of the proposed Topical 

Report DOM–QA–1, ‘‘Dominion Nuclear 

Facility Quality Assurance Program 

Description.’’ The implementation of 

this topical report would create a 

common quality assurance program for 

North Anna, Surry, and Millstone Power 

Stations. The review of these proposed 

amendments was requested to be done 

in concert with the review of the 

Topical Report. The Topical Report was 

submitted to the NRC staff for review on 

August 24, 2004, and supplemented by 

letter dated May 5, 2005. Subsequently, 

the NRC staff approved this Topical 

Report on September 9, 2005. 

Date of issuance: As of the date of 

issuance and shall be implemented 

within 6 months from the date of 

issuance. 

Effective date: September 15, 2005. 

Amendment Nos.: 244/243. 

Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 59095 

Renewed Facility Operating License 

Nos. DPR–32 and DPR–37: Amendments 

change the Technical Specifications. 

Date of initial notice in Federal 

Register: December 7, 2004 (69 FR 

70723). The supplement dated May 5, 

2005, contained clarifying information 

only and did not change the initial no 

significant hazards consideration 

determination or expand the scope of 

the initial application. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 15, 

2005. 

No significant hazards consideration 

comments received: No. 

Notice of Issuance of Amendments to 

Facility Operating Licenses and Final 

Determination of No Significant 

Hazards Consideration and 

Opportunity for a Hearing (Exigent 

Public Announcement or Emergency 

Circumstances) 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application for the 

amendment complies with the 

standards and requirements of the 

Atomic Energy Act of 1954, as amended 

(the Act), and the Commission’s rules 

and regulations. The Commission has 

made appropriate findings as required 

by the Act and the Commission’s rules 

and regulations in 10 CFR Chapter I, 

which are set forth in the license 

amendment. 

Because of exigent or emergency 

circumstances associated with the date 

the amendment was needed, there was 

not time for the Commission to publish, 

for public comment before issuance, its 

usual Notice of Consideration of 

Issuance of Amendment, Proposed No 

Significant Hazards Consideration 

Determination, and Opportunity for a 

Hearing. 

For exigent circumstances, the 

Commission has either issued a Federal 

Register notice providing opportunity 

for public comment or has used local 

media to provide notice to the public in 

the area surrounding a licensee’s facility 

of the licensee’s application and of the 

Commission’s proposed determination 

of no significant hazards consideration. 

The Commission has provided a 

reasonable opportunity for the public to 

comment, using its best efforts to make 

available to the public means of 

communication for the public to 

respond quickly, and in the case of 

telephone comments, the comments 

have been recorded or transcribed as 

appropriate and the licensee has been 

informed of the public comments. 

In circumstances where failure to act 

in a timely way would have resulted, for 

example, in derating or shutdown of a 

nuclear power plant or in prevention of 

either resumption of operation or of 

increase in power output up to the 

plant’s licensed power level, the 

Commission may not have had an 

opportunity to provide for public 

comment on its no significant hazards 

consideration determination. In such 

case, the license amendment has been 

issued without opportunity for 

comment. If there has been some time 

for public comment but less than 30 

days, the Commission may provide an 

opportunity for public comment. If 

comments have been requested, it is so 

stated. In either event, the State has 

been consulted by telephone whenever 

possible. 

Under its regulations, the Commission 

may issue and make an amendment 

immediately effective, notwithstanding 

the pendency before it of a request for 

a hearing from any person, in advance 

of the holding and completion of any 

required hearing, where it has 

determined that no significant hazards 

consideration is involved. 

The Commission has applied the 

standards of 10 CFR 50.92 and has made 

a final determination that the 

amendment involves no significant 

hazards consideration. The basis for this 

determination is contained in the 

documents related to this action. 

Accordingly, the amendments have 

been issued and made effective as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the application for 

amendment, (2) the amendment to 

Facility Operating License, and (3) the 

Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment, as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

reading-rm/adams.html. If you do not 

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

The Commission is also offering an 

opportunity for a hearing with respect to 

the issuance of the amendment. Within 

60 days after the date of publication of 

this notice, 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 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland, 

and electronically on the Internet at the 

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

reading-rm/doc-collections/cfr/. If there 

are problems in accessing the document, 

contact the PDR Reference staff at 1 

(800) 397–4209, (301) 415–4737, or by email to pdr@nrc.gov. If a request for a 

hearing or petition for leave to intervene 

is filed by the above date, the 

Commission or a presiding officer 

designated by the Commission or by the 

Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, 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 general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

59096 Federal Register / Vol. 70, No. 195 / Tuesday, October 11, 2005 / Notices 

1To the extent that the applications contain 

attachments and supporting documents that are not 

publicly available because they are asserted to 

contain safeguards or proprietary information, 

petitioners desiring access to this information 

should contact the applicant or applicant’s counsel 

and discuss the need for a protective order. 1 15 U.S.C. 78l(d). 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for 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. The 

petition must include sufficient 

information to show that a genuine 

dispute exists with the applicant on a 

material issue of law or fact.1

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/requestor who fails to satisfy 

these requirements with respect to at 

least one contention will not be 

permitted to participate as a party. 

Each contention shall be given a 

separate numeric or alpha designation 

within one of the following groups: 

1. Technical—primarily concerns/ 

issues relating to technical and/or 

health and safety matters discussed or 

referenced in the applications. 

2. Environmental—primarily 

concerns/issues relating to matters 

discussed or referenced in the 

environmental analysis for the 

applications. 

3. Miscellaneous—does not fall into 

one of the categories outlined above. 

As specified in 10 CFR 2.309, if two 

or more petitioners/requestors seek to 

co-sponsor a contention, the petitioners/ 

requestors shall jointly designate a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. If a petitioner/requestor 

seeks to adopt the contention of another 

sponsoring petitioner/requestor, the 

petitioner/requestor who seeks to adopt 

the contention must either agree that the 

sponsoring petitioner/requestor shall act 

as the representative with respect to that 

contention, or jointly designate with the 

sponsoring petitioner/requestor a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. 

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. Since the Commission has 

made a final determination that the 

amendment involves no significant 

hazards consideration, if a hearing is 

requested, it will not stay the 

effectiveness of the amendment. Any 

hearing held would take place while the 

amendment is in effect. 

A request for a hearing or a petition 

for leave to intervene must be filed by: 

(1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, Attention: Rulemaking and 

Adjudications Staff; (2) courier, express 

mail, and expedited delivery services: 

Office of the Secretary, Sixteenth Floor, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland, 20852, 

Attention: Rulemaking and 

Adjudications Staff; (3) E-mail 

addressed to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

HearingDocket@nrc.gov; or (4) facsimile 

transmission addressed to the Office of 

the Secretary, U.S. Nuclear Regulatory 

Commission, Washington, DC, 

Attention: Rulemakings and 

Adjudications Staff at (301) 415–1101, 

verification number is (301) 415–1966. 

A copy of the request for hearing and 

petition for leave to intervene should 

also be sent to the Office of the General 

Counsel, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and it is requested that copies be 

transmitted either by means of facsimile 

transmission to (301) 415–3725 or by email to OGCMailCenter@nrc.gov. A copy 

of the request for hearing and petition 

for leave to intervene should also be 

sent to the attorney for the licensee. 

Nontimely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission or the presiding officer or 

the Atomic Safety and Licensing Board 

that the petition, request and/or the 

contentions should be granted based on 

a balancing of the factors specified in 10 

CFR 2.309(a)(1)(I)–(viii). 

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 request: 

September 12, 2005. 

Description of amendment request: 

The amendments replace the paragraph 

of Improved Technical Specification 

(ITS) Surveillance Requirement (SR) 

3.8.1.18 with the wording of previous 

TS SR 4.8.1.1.2.e.11. 

Date of issuance: September 23, 2005. 

Effective date: Immediately. 

Amendment Nos.: 290, 272. 

Facility Operating License Nos. (DPR–

58 and DPR–74): Amendment revises 

the technical specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): Yes. HeraldPalladium on September 18, 2005. The 

notice provided an opportunity to 

submit comments on the Commission’s 

proposed NSHC determination. No 

comments have been received. 

The Commission’s related evaluation 

of the amendment, finding of exigent 

circumstances, state consultation, and 

final NSHC determination are contained 

in a safety evaluation dated September 

23, 2005. 

Attorney for licensee: James M. Petro, 

Jr., Esquire, One Cook Place, Bridgman, 

MI 49106. 

NRC Section Chief: L. Raghavan. 

Dated at Rockville, Maryland, this 3rd day 

of October 2005. 

For the Nuclear Regulatory Commission. 

Ledyard B. Marsh, 

Director, Division of Licensing Project 

Management, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 05–20168 Filed 10–7–05; 8:45 am] 

BILLING CODE 7590–01–P 

SECURITIES AND EXCHANGE 

COMMISSION 

[File No. 1–31514] 

Issuer Delisting; Notice of Application 

of Meredith Enterprises, Inc. to 

Withdraw Its Common Stock, $.01 Par 

Value, From Listing and Registration 

on the American Stock Exchange LLC 

October 4, 2005. 

On September 15, 2005, Meredith 

Enterprises, Inc., a Delaware corporation 

(‘‘Issuer’’), filed an application with the 

Securities and Exchange Commission 

(‘‘Commission’’), pursuant to Section 

12(d) of the Securities Exchange Act of 

1934 (‘‘Act’’) 1 and Rule 12d2–2(d)