52412 Federal Register / Vol. 73, No. 175 / Tuesday, September 9, 2008 / Notices 

The Committee is composed of 16 

individuals appointed by the Secretary. 

The membership of the Committee shall 

include equal representation of 

employers, education community, labor 

organizations, and the public/private 

sectors. The Secretary shall appoint one 

of the members as Chairperson to the 

Committee. A representative of the U.S. 

Department of Education, U.S. 

Department of Health and Human 

Services and the U.S. Department of 

Justice shall be invited to serve as nonvoting members to the Committee exofficio. The Deputy Secretaries of Labor, 

Agriculture, and Interior shall be nonvoting members to the Committee exofficio. The National Director, Office of 

Job Corps, Office of the Secretary 

(OSEC), shall be the designated Federal 

official to the Committee. 

Terms of members shall be 2 years, as 

designated by the Secretary, and all 

Committee members shall serve at the 

pleasure of the Secretary. Appointments 

to vacancies occurring during the terms 

of such appointments shall be for the 

unexpired portions of the terms. 

FOR FURTHER INFORMATION CONTACT: 

Crystal Woodard, Office of Job Corps, 

202–693–3000 (this is not a toll-free 

number). 

Signed at Washington, DC, this 3rd day of 

September 2008. 

Esther R. Johnson, 

Administrator, Office of Job Corps. 

[FR Doc. E8–20870 Filed 9–8–08; 8:45 am] 

BILLING CODE 4510–23–P 

NATIONAL FOUNDATION ON THE 

ARTS AND THE HUMANITIES 

National Endowment for the Arts; Arts 

Advisory Panel 

Pursuant to Section 10(a)(2) of the 

Federal Advisory Committee Act (Pub. 

L. 92–463), as amended, notice is hereby 

given that three meetings of the Arts 

Advisory Panel to the National Council 

on the Arts will be held at the Nancy 

Hanks Center, 1100 Pennsylvania 

Avenue, NW., Washington, DC, 20506 

as follows (ending times are 

approximate): 

Music/Jazz (application review): 

September 29, 2008 by teleconference. 

This meeting, from 12 p.m. to 4 p.m. 

will be closed. 

Learning in the Arts (application 

review): October 3, 2008 in Room 716. 

This meeting, from 9 a.m. to 6 p.m., will 

be closed. 

Learning in the Arts (application 

review): October 14–15, 2008 in Room 

716. A portion of this meeting, from 

1:30 p.m. to 2 p.m. on October 15th, will 

be open to the public for a policy 

discussion. The remainder of the 

meeting, from 9 a.m. to 6 p.m. on 

October 14th, and from 9 a.m. to 1:30 

p.m. and 2 p.m. to 3 p.m. on October 

15th, will be closed. 

Learning in the Arts (application 

review): October 21–24, 2008 in Room 

716. A portion of this meeting, from 

1:15 p.m. to 1:45 p.m. on October 24th, 

will be open to the public for a policy 

discussion. The remainder of the 

meeting, from 9 a.m. to 6 p.m. on 

October 21st—23rd and from 9 a.m. to 

1:15 p.m. and 1:45 p.m. to 5 p.m. on 

October 24th, will be closed. 

The closed portions of meetings are 

for the purpose of Panel review, 

discussion, evaluation, and 

recommendations on financial 

assistance under the National 

Foundation on the Arts and the 

Humanities Act of 1965, as amended, 

including information given in 

confidence to the agency. In accordance 

with the determination of the Chairman 

of February 28, 2008, these sessions will 

be closed to the public pursuant to 

subsection (c)(6) of section 552b of Title 

5, United States Code. 

Any person may observe meetings, or 

portions thereof, of advisory panels that 

are open to the public, and if time 

allows, may be permitted to participate 

in the panel’s discussions at the 

discretion of the panel chairman. If you 

need special accommodations due to a 

disability, please contact the Office of 

AccessAbility, National Endowment for 

the Arts, 1100 Pennsylvania Avenue, 

NW., Washington, DC 20506, 202/682–

5532, TDY–TDD 202/682–5496, at least 

seven (7) days prior to the meeting. 

Further information with reference to 

these meetings can be obtained from Ms. 

Kathy Plowitz-Worden, Office of 

Guidelines & Panel Operations, National 

Endowment for the Arts, Washington, 

DC, 20506, or call 202/682–5691. 

Dated: August 8, 2008. 

Kathy Plowitz-Worden, 

Panel Coordinator, Panel Operations, 

National Endowment for the Arts. 

[FR Doc. E8–20788 Filed 9–8–08; 8:45 am] 

BILLING CODE 7537–01–P 

NATIONAL TRANSPORTATION 

SAFETY BOARD 

Sunshine Act Meeting; Agenda 

TIME AND DATE: 9:30 a.m., Tuesday, 

September 16, 2008. 

PLACE: NTSB Conference Center, 429 

L’Enfant Plaza, SW., Washington, DC 

20594. 

STATUS: The three items are open to the 

public. 

MATTERS TO BE CONSIDERED: 

8042 Special Investigation Report on 

the Safety of Parachute Jump 

Operations. 

8040 Aircraft Accident Summary 

Report on Crash of Skydive Quantum 

Leap, de Havilland DHC–6–100, 

N203E, Sullivan, Missouri, July 29, 

2006. 

8041 Highway Accident Report—

Truck-Tractor Semitrailer Rollover 

and Motorcoach Collision With 

Overturned Truck, Interstate Highway 

94, Near Osseo, Wisconsin, October 

16, 2005. 

NEWS MEDIA CONTACT: Telephone: (202) 

314–6100. 

Individuals requesting specific 

accommodations should contact 

Rochelle Hall at (202) 314–6305 by 

Friday, September 12, 2008. 

The public may view the meeting via 

a live or archived Web cast by accessing 

a link under ‘‘News & Events’’ on the 

NTSB home page at http:// 

www.ntsb.gov. 

FOR MORE INFORMATION CONTACT: Vicky 

D’Onofrio, (202) 314–6410. 

Dated: September 5, 2008. 

Vicky D’Onofrio, 

Federal Register Liaison Officer. 

[FR Doc. E8–21024 Filed 9–5–08; 4:15 pm] 

BILLING CODE 7533–01–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 August 14, 

2008, to August 27, 2008. The last 


jlentini on PROD1PC65 with NOTICES
Federal Register / Vol. 73, No. 175 / Tuesday, September 9, 2008 / Notices 52413 

biweekly notice was published on 

August 26, 2008 (73 FR 50356). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

10 CFR 50.92, this means that operation 

of the facility in accordance with the 

proposed amendment would not (1) 

Involve a significant increase in the 

probability or consequences of an 

accident previously evaluated; or (2) 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated; or (3) 

involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 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. 

Written comments may be submitted 

by mail to the Chief, Rulemaking, 

Directives and Editing 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, person(s) 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 

via electronic submission through the 

NRC E-Filing system 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 person(s) 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. 


jlentini on PROD1PC65 with NOTICES
52414 Federal Register / Vol. 73, No. 175 / Tuesday, September 9, 2008 / Notices 

A request for hearing or a petition for 

leave to intervene must be filed in 

accordance with the NRC E-Filing rule, 

which the NRC promulgated in August 

28, 2007 (72 FR 49139). The E-Filing 

process requires participants to submit 

and serve documents over the internet 

or in some cases to mail copies on 

electronic storage media. Participants 

may not submit paper copies of their 

filings unless they seek a waiver in 

accordance with the procedures 

described below. 

To comply with the procedural 

requirements of E-Filing, at least five (5) 

days prior to the filing deadline, the 

petitioner/requestor must contact the 

Office of the Secretary by e-mail at 

hearingdocket@nrc.gov, or by calling 

(301) 415–1677, to request (1) a digital 

ID certificate, which allows the 

participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and/or (2) creation of an 

electronic docket for the proceeding 

(even in instances in which the 

petitioner/requestor (or its counsel or 

representative) already holds an NRCissued digital ID certificate). Each 

petitioner/requestor will need to 

download the Workplace Forms 

ViewerTM to access the Electronic 

Information Exchange (EIE), a 

component of the E-Filing system. The 

Workplace Forms ViewerTM is free and 

is available at http://www.nrc.gov/sitehelp/e-submittals/install-viewer.html. 

Information about applying for a digital 

ID certificate is available on NRC’s 

public Web site at http://www.nrc.gov/ 

site-help/e-submittals/applycertificates.html. 

Once a petitioner/requestor has 

obtained a digital ID certificate, had a 

docket created, and downloaded the EIE 

viewer, it can then submit a request for 

hearing or petition for leave to 

intervene. Submissions should be in 

Portable Document Format (PDF) in 

accordance with NRC guidance 

available on the NRC public Web site at 

http://www.nrc.gov/site-help/esubmittals.html. A filing is considered 

complete at the time the filer submits its 

documents through EIE. To be timely, 

an electronic filing must be submitted to 

the EIE system no later than 11:59 p.m. 

Eastern Time on the due date. Upon 

receipt of a transmission, the E-Filing 

system time-stamps the document and 

sends the submitter an e-mail notice 

confirming receipt of the document. The 

EIE system also distributes an e-mail 

notice that provides access to the 

document to the NRC Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/ 

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically may 

seek assistance through the ‘‘Contact 

Us’’ link located on the NRC Web site 

at http://www.nrc.gov/site-help/esubmittals.html or by calling the NRC 

technical help line, which is available 

between 8:30 a.m. and 4:15 p.m., 

Eastern Time, Monday through Friday. 

The help line number is (800) 397–4209 

or locally, (301) 415–4737. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file a 

motion, in accordance with 10 CFR 

2.302(g), with their initial paper filing 

requesting authorization to continue to 

submit documents in paper format. 

Such filings must be submitted 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; or 

(2) courier, express mail, or expedited 

delivery service to the Office of the 

Secretary, Sixteenth Floor, One White 

Flint North, 11555 Rockville Pike, 

Rockville, Maryland 20852, Attention: 

Rulemaking and Adjudications Staff. 

Participants filing a document in this 

manner are responsible for serving the 

document on all other participants. 

Filing is considered complete by firstclass mail as of the time of deposit in 

the mail, or by courier, express mail, or 

expedited delivery service upon 

depositing the document with the 

provider of the service. 

Non-timely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission, the presiding officer, or 

the Atomic Safety and Licensing Board 

that the petition and/or request should 

be granted and/or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

CFR 2.309(c)(1)(i)–(viii). To be timely, 

filings must be submitted no later than 

11:59 p.m. Eastern Time on the due 

date. 

Documents submitted in adjudicatory 

proceedings will appear in NRC’s 

electronic hearing docket which is 

available to the public at http:// 

ehd.nrc.gov/EHD_Proceeding/home.asp, 

unless excluded pursuant to an order of 

the Commission, an Atomic Safety and 

Licensing Board, or a Presiding Officer. 

Participants are requested not to include 

personal privacy information, such as 

social security numbers, home 

addresses, or home phone numbers in 

their filings. With respect to copyrighted 

works, except for limited excerpts that 

serve the purpose of the adjudicatory 

filings and would constitute a Fair Use 

application, participants are requested 

not to include copyrighted materials in 

their submission. 

For further details with respect to this 

amendment 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. 

Dominion Energy Kewaunee, Inc. Docket 

No. 50–305, Kewaunee Power Station 

(KPS), Kewaunee County, Wisconsin 

Date of amendment request: August 

14, 2008. 

Description of amendment request: 

The proposed amendment would 

modify Specification 4.4.f.1, 

‘‘Containment Isolation Device 

Verification,’’ of the Technical 

Specifications (TS) to require 

verification that the 36-inch 

containment purge and vent isolation 

valves are sealed closed when the 

reactor is at greater than Cold Shutdown 

conditions. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

(1) Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The Design Bases Accidents (DBA) that 

result in a release of radioactive material 

within containment are a steam line break, 

rupture of a rod cluster control assembly, and 

loss-of-coolant accident (LOCA). In the 

analyses for each of these accidents, it is 

assumed that containment isolation valves 

are either closed or function to close within 

the required isolation time following 

accident initiation. This ensures that 

potential leakage paths to the environment 


jlentini on PROD1PC65 with NOTICES
Federal Register / Vol. 73, No. 175 / Tuesday, September 9, 2008 / Notices 52415 

through containment isolation valves 

(including containment purge and vent 

isolation valves) are minimized. The safety 

analyses assume that the containment purge 

and vent isolation valves are closed at 

accident initiation. 

The safety function of the containment 

purge and vent isolation valves is to support 

the Containment Isolation system by 

confining fission products within the 

Primary Containment system boundary 

during a DBA. The proposed amendment 

would require verification that the 

containment purge and vent isolation valves 

are sealed closed when the reactor is at 

greater than Cold Shutdown conditions. This 

requirement ensures the valves are in their 

required DBA post-accident position when 

the reactor is at greater than Cold Shutdown 

conditions. 

Verifying the containment purge and vent 

isolation valves are sealed closed at 31-day 

intervals does not add, delete, or modify any 

KPS system, structure, or component (SSC). 

Verifying that the containment purge and 

vent isolation valves are sealed closed when 

the reactor is at greater than Cold Shutdown 

conditions has no adverse effect on the 

ability of the plant to mitigate the effects of 

DBAs. The subject surveillance requirement 

constitutes a verification of isolation valve 

position and has no effect on equipment. 

Verification of valve closure only ensures the 

previous assumptions made in evaluating the 

consequences of DBAs remain valid. 

Therefore, there is no increase in the 

probability of an accident by performing the 

surveillance in additional modes of plant 

operation. 

Therefore, the proposed amendment does 

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. 

Verifying the containment purge and vent 

isolation valves are sealed closed when the 

reactor is at greater than Cold Shutdown 

conditions at 31-day intervals ensures these 

valves are in their required DBA postaccident position when the design function 

is required. The proposed amendment does 

not change the manner in which these valves 

are operated when the reactor is at or below 

Cold Shutdown or their design function. The 

proposed amendment does not create any 

new failure mechanisms or malfunctions for 

plant equipment or the nuclear fuel. 

In addition, the containment purge and 

vent isolation valves are not accident 

initiators. Their function is only for 

mitigation of accidents. 

Therefore, the proposed amendment does 

not create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

(3) Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Verifying the containment purge and vent 

isolation valves are sealed closed when the 

reactor is at greater than Cold Shutdown 

conditions at 31-day intervals ensures these 

valves are in their required DBA postaccident position when the design function 

is required. The proposed amendment does 

not change the manner in which these valves 

are operated when the reactor is at or below 

Cold Shutdown condition. 

The proposed amendment would align the 

KPS TS with applicable NRC requirements 

stated in NUREG–0800 [‘‘Standard Review 

Plan,’’], Section 6.2.4 and NUREG–0737 

[‘‘Clarification of Three Mile Island Action 

Plan Requirements,’’], Item II.E.4.2. The 

proposed amendment does not result in 

altering or exceeding a design basis or safety 

limit for the plant. The safety analysis of 

record, including evaluations of the 

radiological consequences of design basis 

accidents, will remain applicable and 

unchanged. 

Therefore, the proposed 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: Lillian M. 

Cuoco, Senior Counsel, Dominion 

Resources Services, Inc., Counsel for 

Dominion Energy Kewaunee, Inc., 120 

Tredegar Street, Richmond, VA 23219. 

NRC Branch Chief: Lois James. 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–269, 50–270, and 50–287, 

Oconee Nuclear Station, Units 1, 2, and 

3, Oconee County, South Carolina 

Date of amendment request: August 1, 

2008. 

Description of amendment request: 

The proposed amendments would 

authorize changes to the Updated Final 

Safety Analysis Report (UFSAR) to 

account for small areas of carbon steel 

(CS) and low alloy steel that may be 

exposed to the reactor coolant system 

(RCS). 

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? 

No. 

The Pressurizer vent nozzle and 

thermowell, as components of the RCS, must 

maintain system pressure boundary. RCS 

design pressure is 2500 psig and design 

temperature is 670 °F. The vent nozzle and 

thermowell replacements are designed for the 

RCS pressure and temperature. As described 

above, the material of the new Pressurizer 

vent nozzle and thermowell is an 

improvement in the PWSCC [primary water 

stress corrosion cracking] resistance of those 

components as compared to the original 

components. The design of the new 

Pressurizer vent nozzle and thermowell 

exposes small areas of the Pressurizer shell 

carbon steel to a stagnant reactor coolant 

environment. However, the corrosion of the 

Pressurizer shell is considered negligible. 

Therefore, the replacement of the Pressurizer 

vent nozzle and thermowell do not more than 

minimally increase the likelihood of 

occurrence of a malfunction. Corrosion 

evaluations performed show that all 

applicable ASME Code requirements are met. 

It is concluded that the consequences of a 

Pressurizer vent nozzle or Pressurizer 

thermowell failure resulting in a LOCA [lossof-coolant accident] are bounded by existing 

analysis. Therefore, there is no increase in 

the probability or consequences of an 

accident. 

(2) Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. 

The only credible accident involving the 

failure of these components is bounded by 

existing LOCA analyses. There are no new 

accidents that need to be postulated due to 

the replacement of the Pressurizer vent 

nozzle and Pressurizer thermowell. 

Therefore, this proposed activity will not 

create the possibility of a new or different 

kind of accident from any kind of accident 

previously evaluated. 

(3) Does the proposed amendment involve 

a significant reduction in a margin of safety? 

No. 

The mitigation technique selected for the 

Pressurizer vent nozzle and the Pressurizer 

thermowell exposes a small area of CS to the 

RCS environment. As required by the ASME 

Code, Section III, a supporting corrosion 

evaluation was developed within each of the 

two component designs. The technical 

package for the replacement of the 

Pressurizer vent nozzle and the Pressurizer 

thermowell utilized calculations to support 

the evaluation of the acceptability of this 

repair/replacement activity. The corrosion 

evaluation for the Pressurizer vent resulted in 

a conservative general stagnant corrosion rate 

of 0.0018 inches per year and the corrosion 

evaluation for the Pressurizer thermowell 

resulted in a conservative general corrosion 

rate of 0.00142 inches per year. The critical 

corrosion distance is the radius from the 

exposed CS surface to the edge of the weld 

pad. This distance is at least 1.1 inches for 

both the vent and thermowell designs. With 

this distance, a corrosion rate of less than 2 

mils per year is not significant when 

compared to the 60 year component design 

life, which begins at the time of installation. 

The original Pressurizer was designed to 

meet Section III of the ASME Code, and the 

Pressurizer, as modified, meets Section III of 

the ASME code. Although this change does 

expose small areas of CS in the Pressurizer, 

the 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 


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review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ms. Lisa F. 

Vaughn, Associate General Counsel and 

Managing Attorney, Duke Energy 

Carolinas, LLC, 526 South Church 

Street, EC07H, Charlotte, NC 28202. 

NRC Branch Chief: Melanie C. Wong. 

Entergy Nuclear Operations, Inc., 

Docket Nos. 50–247 and 50–286, Indian 

Point Nuclear Generating Unit Nos. 2 

and 3, Westchester County, New York 

Date of amendment request: July 8, 

2008. 

Description of amendment request: 

The proposed amendment to Indian 

Point Nuclear Generating Units Nos. 2 

and 3 (IP2 and IP3) would require the 

licensee to submit information and 

analyses associated with extending the 

Reactor Vessel (RV) Inservice Inspection 

(ISI) Interval from 10 to 20 years for 

specific pressure retaining welds in the 

RV. 

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 will 

revise the license to require the submission 

of information and analyses to the NRC 

following completion of each ASME 

[American Society of Mechanical Engineers] 

[C]ode, Section XI, Category B–A and B–D 

Reactor Vessel weld inspection. The 

extension of the ISI from 10 to 20 years is 

being evaluated as part of the relief request 

independent from the license change. 

Submission of the information and analyses 

can have no effect on the consequences of an 

accident or the probability of an accident 

because the submission of information is not 

related to the operation of the plant or any 

equipment, the programs and procedures 

used to operate the plant, or the evaluation 

of accidents. The submittal of information 

and analyses provides the opportunity for the 

NRC to independently assess the information 

and analyses. 

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 will 

only affect the requirement to submit 

information and analyses when specified 

inspections are performed. There are no 

changes to plant equipment, operating 

characteristics or conditions, programs, and 

procedures or training. Therefore, there are 

no potential new system interactions or 

failures that could 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 will 

revise the license to require the submission 

of information and analyses to the NRC 

following completion of each ASME [C]ode, 

Section XI, Category B–A and B–D Reactor 

Vessel weld inspection which does not affect 

any Limiting Conditions for Operation used 

to establish the margin of safety. The 

requirement to submit information and 

analyses is an administrative tool to assure 

the NRC has the ability to independently 

review information developed by the 

[l]icensee. 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. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Mark G. Kowal. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit No. 2, 

Westchester County, New York 

Date of amendment request: July 9, 

2008. 

Description of amendment request: 

The proposed amendment will revise 

the test acceptance criteria specified in 

the Technical Specification Surveillance 

Requirement (SR) 3.8.1.10 for the Diesel 

Generator (DG) endurance test. The load 

ranges and power factors specified for 

the test will be changed for consistency 

with the associated safety analyses. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The proposed change revises the 

acceptance criteria to be applied to an 

existing surveillance test of the facility 

emergency diesel generators (DGs). 

Performing a surveillance test is not an 

accident initiator and does not increase the 

probability of an accident occurring. The 

proposed new acceptance criteria will assure 

that the DGs are capable of carrying the peak 

electrical loading assumed in the various 

existing safety analyses which take credit for 

the operation of the DGs. Establishing 

acceptance criteria that bound existing 

analyses validates the related assumption 

used in those analyses regarding the 

capability of equipment to mitigate accident 

conditions. 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? 

No. The proposed change revises the test 

acceptance criteria for a specific performance 

test conducted on the existing DGs. The 

proposed change does not involve 

installation of new equipment or 

modification of existing equipment, so no 

new equipment failure modes are introduced. 

The proposed revision to the DG surveillance 

test acceptance criteria also is not a change 

to the way that the equipment or facility is 

operated and no new accident initiators are 

created. 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? 

No. The conduct of performance tests on 

safety-related plant equipment is a means of 

assuring that the equipment is capable of 

maintaining the margin of safety established 

in the safety analyses for the facility. The 

proposed change in the DG technical 

specification surveillance test acceptance 

criteria is consistent with values assumed in 

existing safety analyses is consistent with the 

design rating of the DGs. Therefore the 

propose 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. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Mark G. Kowal. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Plant, Van 

Buren County, Michigan 

Date of amendment request: May 5, 

2008. 

Description of amendment request: 

The proposed amendment would 

correct an error in Section A.1 of the 

renewed operating license and remove 

several outdated license conditions 

relating to surveillance requirements. 

Specifically, it would remove the words 

‘‘filed by Entergy Nuclear Palisades, 

LLC (ENP) and Entergy Nuclear 


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Operations, Inc. (ENO)’’ in Section A.1, 

spell-out acronyms used in Section 1.F, 

and delete license conditions 2.C.(4) 

and 2.C.(5), and delete Table 2.C.(5). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed license amendment deletes 

incorrect or outdated information from the 

renewed facility operating license. The 

proposed amendment does not involve 

operation of the required structures, systems 

or components (SSCs) in a manner or 

configuration different from those previously 

recognized or evaluated. 

Modification of renewed facility operating 

license sections 1.A and 1.F and deletion of 

license conditions 2.C.(4), 2.C.(5), and Table 

2.C.(5) is administrative and has no impact 

on plant operation or equipment. 

Therefore, operation of the facility in 

accordance with the proposed amendment 

would not involve a significant increase in 

the probability or consequences of an 

accident previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed license amendment does not 

involve a physical alteration of any SSC or 

change the way any SSC is operated. The 

proposed license amendment does not 

involve operation of any required SSCs in a 

manner or configuration different from those 

previously recognized or evaluated. 

Modification of renewed facility operating 

license sections 1.A and 1.17 and deletion of 

license conditions 2.C.(4), 2.C.(5), and Table 

2.C.(5) is administrative and has no impact 

on plant operation or equipment. 

Therefore, the proposed amendment does 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Modification of renewed facility operating 

license sections 1.A and 1.F and deletion of 

license conditions 2.C.(4), 2.C.(5), and Table 

2.C.(5) is administrative and has no impact 

on plant operation or equipment or on any 

margin of safety. 

Therefore, the proposed amendment would 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. William 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Ave., White Plains, NY 10601. 

NRC Branch Chief: Lois M. James. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Plant, Van 

Buren County, Michigan 

Date of amendment request: May 5, 

2008 

Description of amendment request: 

The proposed amendment would revise 

renewed facility operating license DPR–

20 to remove license condition 2F. This 

license condition describes reporting 

requirements for exceeding the facility 

steady-state reactor core power level 

described in condition 2.C.(1). The 

proposed change is consistent with the 

Nuclear Regulatory Commission (NRC)- 

approved change notice published in 

the Federal Register on November 4, 

2005, announcing the availability of this 

improvement through the consolidated 

line item improvement process. The 

Federal Register Notice included a 

model safety evaluation and model no 

significant hazards consideration 

(NSHC) determination, relating to the 

elimination of the license condition 

involving reporting of violations of 

other requirements (typically in License 

Conditions 2.C) in the operating license 

of some commercial nuclear power 

plants. The licensee affirmed the 

applicability of the model NSHC 

determination in its application dated 

May 5, 2008. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the change involve a significant 

increase in the probability or consequences 

of an accident previously evaluated? 

Response: No. 

The proposed change involves the deletion 

of a reporting requirement. The change does 

not affect plant equipment or operating 

practices and therefore does not significantly 

increase the probability or consequences of 

an accident previously evaluated. 

2. Does the change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

Response: No. 

The proposed change is administrative in 

that it deletes a reporting requirement. The 

change does not add new plant equipment, 

change existing plant equipment, or affect the 

operating practices of the facility. Therefore, 

the 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 deletes a reporting 

requirement. The change does not affect 

plant equipment or operating practices and 

therefore 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. William 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Ave., White Plains, NY 10601. 

NRC Branch Chief: Lois M. James. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and STN 50–

457, Braidwood; Station, Units 1 and 2, 

Will County, Illinois; Docket Nos. STN 

50–454 and STN 50–455, Byron Station, 

Unit Nos. 1 and 2, Ogle County, Illinois 

Date of amendment request: July 29, 

2008. 

Description of amendment request: 

The proposed amendments would 

remove time, cycle, or modificationrelated items from the operating licenses 

(OLs) and technical specifications (TSs) 

at both stations. Additionally, the 

proposed amendments would correct 

typographical errors introduced into the 

TSs at both stations in previous 

amendments. The time, cycle, or 

modification-related items have been 

implemented or superseded, are no 

longer applicable, and no longer need to 

be maintained in their associated OLs or 

TSs. 

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. 

The initial conditions and methodologies 

used in the accident analyses remain 

unchanged. The proposed changes do not 

change or alter the design assumptions for 

the systems or components used to mitigate 

the consequences of an accident. Therefore, 

accident analyses results are not impacted. 

All changes proposed by EGC in this 

amendment request are administrative in 

nature, and are removing one-time 

requirements that have been satisfied or 

items that are no longer applicable. There are 

no physical changes to the facilities, nor any 

changes to the station operating procedures, 


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limiting conditions for operation, or limiting 

safety system settings. 

Based on the above discussion, the 

proposed 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. 

None of the proposed changes affect the 

design or operation of any system, structure, 

or component in the plant. The safety 

functions of the related structures, systems, 

or components are not changed in any 

manner, nor is the reliability of any structure, 

system, or component reduced by the revised 

surveillance or testing requirements. The 

changes do not affect the manner by which 

the facility is operated and do not change any 

facility design feature, structure, system, or 

component. No new or different type of 

equipment will be installed. Since there is no 

change to the facility or operating 

procedures, and the safety functions and 

reliability of structures, systems, or 

components are not affected, the proposed 

changes do not create the possibility of a new 

or different kind of accident from any 

accident previously evaluated. 

Based on this evaluation, the proposed 

changes do not create the possibility of a new 

or different kind of accident from any 

accident previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes to the Facility 

Operating Licenses and TS are administrative 

in nature and have no impact on the margin 

of safety of any of the TS. There is no impact 

on safety limits or limiting safety system 

settings. The changes do not affect any plant 

safety parameters or setpoints. The Operating 

License Conditions have been satisfied as 

required. There are no changes to the 

conditions themselves. 

Based on this evaluation, the proposed 

change does not involve a significant 

reduction in a margin of safety. 

The Nuclear Regulatory Commission 

(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. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Russell Gibbs. 

Florida Power and Light Company, 

et al. , Docket No. 50–389, St. Lucie 

Plant, Unit No. 2, St. Lucie County, 

Florida 

Date of amendment request: January 

23, 2008. 

Description of amendment request: 

Replace the current Technical 

Specification pressure/temperature (P/ 

T) limit curves with new P/T limit 

curves applicable to 55 effective fullpower years (EFPY). The lowtemperature overpressure protection 

(LTOP) requirements, which are based 

on the P/T limits, will also be applicable 

to 55 EFPY. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

(1) Operation of the facility in accordance 

with the proposed amendment would not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

The proposed changes have been 

determined in accordance with the 

methodologies set forth in the regulations to 

provide an adequate margin of safety to 

ensure that the reactor vessel will withstand 

the effects of normal startup and shutdown 

cyclic loads due to system temperature and 

pressure changes as well as the loads 

associated with reactor trips. The regulations 

of 10 CFR Part 50 Appendix A, Design 

Criterion 14 and Design Criterion 31 remains 

satisfied. The pressure-temperature (P/T) 

limit curves in the Technical Specifications 

are conservatively generated in accordance 

with the fracture toughness requirements of 

the ASME [American Society of Mechanical 

Engineers] Code Section XI, Appendix G. The 

margins of safety against fracture provided by 

the P/T limits using the requirements of 10 

CFR 50 Appendix G are equivalent to those 

recommended in ASME Section XI, 

Appendix G. The Adjusted Reference 

Temperature (ART) values are based on the 

guidance of RG [Regulatory Guide] 1.99 

[Reference 4]. 

The proposed changes will not result in 

physical changes to structures, systems or 

components SSCs or to event initiators or 

precursors. Changing the heatup and 

cooldown curves and the pressure relief 

setpoints to reflect 55 EFPY does not affect 

the ability to control the RCS [reactor coolant 

system] at low temperatures such that the 

integrity of the reactor coolant pressure 

boundary would not be compromised by 

violating the P/T limits. 

The proposed changes will not impact 

assumptions and conditions previously used 

in the radiological consequence evaluations 

nor affect mitigation of these consequences 

due to an accident described in the UFSAR 

[Updated Final Safety Analysis Report]. Also, 

the proposed changes will not impact a plant 

system such that previously analyzed SSCs 

might be more likely to fail. The initiating 

conditions and assumptions for accidents 

described in the UFSAR remain as analyzed. 

Thus, based on the above, reasonable 

assurance is provided that the proposed 

amendment does not significantly increase 

the probability or consequences of accidents 

previously evaluated. 

(2) Operation of the facility in accordance 

with the proposed amendment would not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

The requirements for P/T limit curves and 

LTOP have been in place since the beginning 

of plant operation. The revised curves are 

based on a later edition of Section XI of the 

ASME Code that incorporates current 

industry standards for P/T curves. The 

revised curves also are based on reactor 

vessel irradiation damage predictions using 

RG 1.99 methodology. No new failure modes 

are identified nor are any SSCs required to 

be operated outside of their design bases. 

Consequently, the proposed amendment does 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

(3) Operation of the facility in accordance 

with the proposed amendment would not 

involve a significant reduction in a margin of 

safety. 

The proposed P/T curves continue to 

maintain the safety margins of 10 CFR 50 

Appendix G by defining the limits of 

operation which prevent nonductile failure 

of the reactor pressure vessel. Analyses have 

demonstrated that the fracture toughness 

requirements are satisfied and that 

conservative operating restrictions are 

maintained for the purpose of low 

temperature overpressure protection. The P/ 

T limit curves provide assurance that the 

RCS pressure boundary will behave in a 

ductile manner and that the probability of a 

rapidly propagating fracture is minimized. 

Therefore, operation in accordance with the 

proposed amendment would not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: M.S. Ross, 

Attorney, Florida Power & Light, P.O. 

Box 14000, Juno Beach, Florida 33408–

0420. 

NRC Branch Chief: Thomas H. Boyce. 

Nuclear Management Company, LLC, 

Docket No. 50–263, Monticello Nuclear 

Generating Plant (MNGP), Wright 

County, Minnesota 

Date of amendment request: April 4, 

2008. 

Description of amendment request: 

The licensee proposed to change the 

Technical Specifications (TS) to revise 

requirements for unavailable barriers by 

adding Limiting Condition for 

Operation (LCO) 3.0.9. This LCO would 

establish conditions under which 

systems would remain operable when 

required physical barriers are not 

capable of providing their related 

support function. This proposed 

amendment is consistent with the NRC’s 

approved Technical Specification Task 


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Force (TSTF) Improved Standard 

Technical Specifications Change 

Traveler, TSTF–427, Revision 2. A 

notice of availability of this TS 

improvement was published in the 

Federal Register on October 3, 2006 (71 

FR 58444) as part of NRC’s Consolidated 

Line Item Improvement Process (CLIIP). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee provided an analysis of the 

issue of no significant hazards 

consideration by citing the proposed 

NSHC determination published by the 

NRC staff in the Federal Register 

referenced above. That proposed NSHC 

is reproduced below: 

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change allows a delay time 

for entering a supported system technical 

specification (TS) when the inoperability is 

due solely to an unavailable barrier if risk is 

assessed and managed. The postulated 

initiating events which may require a 

functional barrier are limited to those with 

low frequencies of occurrence, and the 

overall TS system safety function would still 

be available for the majority of anticipated 

challenges. Therefore, the probability of an 

accident previously evaluated is not 

significantly increased, if at all. The 

consequences of an accident while relying on 

the allowance provided by proposed LCO 

3.0.9 are no different than the consequences 

of an accident while relying on the TS 

required actions in effect without the 

allowance provided by proposed LCO 3.0.9. 

Therefore, the consequences of an accident 

previously evaluated are not significantly 

affected by this change. The addition of a 

requirement to assess and manage the risk 

introduced by this change will further 

minimize possible concerns. Therefore, this 

change does not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From any Previously 

Evaluated 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed). 

Allowing delay times for entering supported 

system TS when inoperability is due solely 

to an unavailable barrier, if risk is assessed 

and managed, will not introduce new failure 

modes or effects and will not, in the absence 

of other unrelated failures, lead to an 

accident whose consequences exceed the 

consequences of accidents previously 

evaluated. The addition of a requirement to 

assess and manage the risk introduced by this 

change will further minimize possible 

concerns. Thus, this change does not create 

the possibility of a new or different kind of 

accident from an accident previously 

evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety 

The proposed change allows a delay time 

for entering a supported system TS when the 

inoperability is due solely to an unavailable 

barrier, if risk is assessed and managed. The 

postulated initiating events which may 

require a functional barrier are limited to 

those with low frequencies of occurrence, 

and the overall TS system safety function 

would still be available for the majority of 

anticipated challenges. The risk impact of the 

proposed TS changes was assessed following 

the three-tiered approach recommended in 

RG [Regulatory Guide] 1.177. A bounding 

risk assessment was performed to justify the 

proposed TS changes. This application of 

LCO 3.0.9 is predicated upon the licensee’s 

performance of a risk assessment and the 

management of plant risk. The net change to 

the margin of safety is insignificant as 

indicated by the anticipated low levels of 

associated risk (ICCDP [incremental 

conditional core damage probability] and 

ICLERP [incremental conditional large early 

release probability] ) as shown in Table 1 of 

Section 3.1.1 in the Safety Evaluation. 

Therefore, this change does not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

analysis cited by the licensee, and has 

found that the three standards of 10 CFR 

50.92(c) are satisfied. Therefore, the 

NRC staff proposes to determine that the 

proposed amendment involves no 

significant hazards consideration. 

Attorney for licensee: Jonathan Rogoff, 

Esquire, Vice President, Counsel & 

Secretary, Nuclear Management 

Company, LLC, 700 First Street, 

Hudson, WI 54016. 

NRC Acting Branch Chief: Lois M. 

James. 

Nuclear Management Company, LLC, 

Docket No. 50–263, Monticello Nuclear 

Generating Plant (MNGP), Wright 

County, Minnesota 

Date of amendment request: April 22, 

2008. 

Description of amendment request: 

The licensee proposed to change the 

Technical Specifications (TS) to (1) 

revise the surveillance requirement 

frequency in Specification 3.1.3, 

‘‘Control Rod Operability,’’ to require 

control rod notch testing to be 

performed at a 31-day frequency for 

both partially and fully withdrawn 

control rods; and (2) revise Example 

1.4–3 in Section 1.4, ‘‘Frequency,’’ to 

clarify the applicability of the 1.25 

surveillance test interval extension. 

These proposed changes are consistent 

with the NRC’s approved Technical 

Specification Task Force (TSTF) 

Improved Standard Technical 

Specifications (STS) Change Traveler, 

TSTF–475, Revision 1. A notice of 

availability of this TS improvement was 

published in the Federal Register on 

November 13, 2007 (72 FR 63935), as 

part of the NRC’s Consolidated Line 

Item Improvement Process (CLIIP). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 FR 50.91(a), the 

licensee provided an analysis of the 

issue of no significant hazards 

consideration by citing the proposed 

NSHC determination published by the 

NRC staff in the Federal Register notice 

referenced above. That proposed NSHC 

is reproduced below: 

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change generically 

implements TSTF–475, Revision 1, ‘‘Control 

Rod Notch Testing Frequency and SRM 

[Source Range Monitoring] Insert Control 

Rod Action.’’ TSTF–475, Revision 1, 

modifies NUREG–1433 (BWR [Boiling Water 

Reactor]/4) and NUREG–1434 (BWR/6) STS. 

The changes (1) revise TS testing frequency 

for surveillance requirement (SR) 3.1.3.2 in 

TS 3.1.3, ‘‘Control Rod OPERABILITY,’’, and 

(2) revise Example 1.4–3 in Section 1.4 

‘‘Frequency’’ to clarify the applicability of 

the 1.25 surveillance test interval extension. 

The consequences of an accident after 

adopting TSTF–475, Revision 1 are no 

different than the consequences of an 

accident prior to adoption. Therefore, this 

change does not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From any Accident 

Previously Evaluated 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. The proposed change will 

not introduce new failure modes or effects 

and will not, in the absence of other 

unrelated failures, lead to an accident whose 

consequences exceed the consequences of 

accidents previously analyzed. Thus, this 

change does not create the possibility of a 

new or different kind of accident from any 

accident previously evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety 

TSTF–475, Revision 1 will: (1) Revise the 

TS SR 3.1.3.2 frequency in TS 3.1.3, ‘‘Control 

Rod OPERABILITY,’’ and (2) revise Example 

1.4–3 in Section 1.4 ‘‘Frequency’’ to clarify 

the applicability of the 1.25 surveillance test 

interval extension. The GE Nuclear Energy 

Report, ‘‘CRD Notching Surveillance Testing 

for Limerick Generating Station,’’ dated 

November 2006, concludes that extending 

the control rod notch test interval from 

weekly to monthly is not expected to impact 

the reliability of the scram system and that 

the analysis supports the decision to change 

the surveillance frequency. Therefore, the 

proposed changes in TSTF–475, Revision 1 


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52420 Federal Register / Vol. 73, No. 175 / Tuesday, September 9, 2008 / Notices 

are acceptable and do not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

analysis cited by the licensee, and has 

found that the three standards of 10 CFR 

50.92(c) are satisfied. Therefore, the 

NRC staff proposes to determine that the 

proposed amendment involves no 

significant hazards consideration. 

Attorney for licensee: Jonathan Rogoff, 

Esquire, Vice President, Counsel & 

Secretary, Nuclear Management 

Company, LLC, 700 First Street, 

Hudson, WI 54016. 

NRC Acting Branch Chief: Lois M. 

James. 

Nuclear Management Company, LLC, 

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

Island Nuclear Generating Plant, Units 

1 and 2, Goodhue County, Minnesota 

Date of amendment request: June 26, 

2008. 

Description of amendment request: 

The proposed amendments would 

amend the Facility Operating Licenses 

by revising the licensing basis loss of 

coolant accident and main steam line 

break accident radiological dose 

consequences for Prairie Island Nuclear 

Generating Plant, Units 1 and 2, as 

currently described in the Updated 

Safety Analysis Report Section 14.5 and 

Section 14.9. This proposed amendment 

also proposes concomitant amendments 

to Appendix A of the Facility Operating 

Licenses, Technical Specifications (TS) 

3.3.5, ‘‘Containment Ventilation 

Isolation Instrumentation’’, 3.4.17, ‘‘RCS 

[Reactor Coolant System] Specific 

Activity’’, and 3.6.3, ‘‘Containment 

Isolation Valves’’, which are necessary 

to implement the proposed revised 

analyses. 

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. 

This license amendment request proposes 

implementing revised loss of coolant 

accident and main steam line break accident 

dose consequence analyses to address 

modeling nonconservatisms and update the 

analyses for new fuel types and provide 

margin for power uncertainty. These analyses 

assumed that the containment inservice 

purge system penetrations are isolated, thus 

this license amendment request proposes 

Technical Specification revisions which will 

require these penetrations to be blind flanged 

during plant operations; these changes allow 

the Technical Specification requirements for 

containment ventilation isolation 

instrumentation to be removed. This license 

amendment request also proposes associated 

more restrictive limits in the Technical 

Specification for reactor coolant system 

specific activity since the main steam line 

break accident analysis assumed lower 

limits. 

The accident radiological dose 

consequences analyses inputs, methodologies 

and outputs modified by this request are not 

accident initiators and do not affect the 

frequency of occurrence of previously 

analyzed transients. Likewise, the reactor 

coolant system specific activity limits are not 

accident initiators and do not affect the 

frequency of occurrence of previously 

analyzed transients. 

The containment inservice purge system is 

not an accident initiator and therefore 

removal of its Technical Specifications does 

not involve an increase in the probability of 

an accident. The Technical Specification 

changes proposed in this license amendment 

request require the containment inservice 

purge system to be blind flanged during 

Modes 1, 2, 3, and 4, therefore removal of the 

containment ventilation isolation 

instrumentation Technical Specifications and 

other Technical Specification system 

operating requirements does not involve an 

increase in the consequences of an accident 

previously evaluated. 

The loss of coolant accident and main 

steam line break accident radiological dose 

consequences analyses demonstrated the 

results are within the applicable regulatory 

limits and guidance using revised inputs, 

including the proposed lower Technical 

Specification reactor coolant system specific 

activity limits, and methodologies. Thus 

these changes do not involve a significant 

increase in the consequences of an accident. 

Therefore, the proposed 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. 

This license amendment request proposes 

implementing revised loss of coolant 

accident and main steam line break accident 

dose consequence analyses to address 

modeling nonconservatisms and update the 

analyses for new fuel types and provide 

margin for power uncertainty. These analyses 

assumed that the containment inservice 

purge system penetrations are isolated, thus 

this license amendment request proposes 

Technical Specification revisions which will 

require these penetrations to be blind flanged 

during plant operations; these changes allow 

the Technical Specification requirements for 

containment ventilation isolation 

instrumentation to be removed. This license 

amendment request also proposes associated 

more restrictive limits in the Technical 

Specification for reactor coolant system 

specific activity since the main steam line 

break accident analysis assumed lower 

limits. 

This license amendment request does not 

involve physical changes to the plant 

structures, systems or components and there 

is no adverse impact on component or system 

interactions due to the proposed changes. 

The modes of operation of the plant remain 

unchanged and the design functions of the 

safety systems remain in compliance with the 

applicable safety analysis acceptance criteria. 

These changes do not create new failure 

modes or mechanisms and no new accident 

precursors are generated. 

When the containment inservice purge 

system is not being operated, current 

Technical Specifications require the system’s 

penetrations to be blind flanged in Modes 1, 

2, 3, and 4 to provide post-accident 

containment integrity. This license 

amendment proposes to require the system 

penetrations to be blind flanged at all times 

during these Modes and prevent operation of 

the system in these Modes. Since 

containment integrity is provided with the 

penetrations blind flanged and this change 

only extends the time during which the 

system is in this configuration, these changes 

do not create the possibility of a new or 

different kind of accident. 

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 amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

This license amendment request proposes 

implementing revised loss of coolant 

accident and main steam line break accident 

dose consequence analyses to address 

modeling nonconservatisms and update the 

analyses for new fuel types and provide 

margin for power uncertainty. These analyses 

assumed that the containment inservice 

purge system penetrations are isolated, thus 

this license amendment request proposes 

Technical Specification revisions which will 

require these penetrations to be blind flanged 

during plant operations; these changes allow 

the Technical Specification requirements for 

containment ventilation isolation 

instrumentation to be removed. This license 

amendment request also proposes associated 

more restrictive limits in the Technical 

Specification for reactor coolant system 

specific activity since the main steam line 

break accident analysis assumed lower 

limits. 

The loss of coolant accident and main 

steam line break accident radiological dose 

consequences analyses have incorporated 

revised inputs, including the proposed lower 

Technical Specification reactor coolant 

system specific activity limits, and utilized 

revised methodologies. The results of these 

revised analyses satisfy the applicable 

regulatory limits and guidance. There is no 

adverse effect on plant safety due to this 

proposed license amendment. 

The containment inservice purge system is 

not credited for mitigation of any accidents 

or any other safety function, thus, removal of 

its associated Technical Specifications does 

not involve reduction in a margin of safety. 

The containment ventilation isolation 

instrumentation system is credited for 

isolation of the containment inservice purge 

system following an accident and the valves 

are assumed to meet containment integrity 


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Federal Register / Vol. 73, No. 175 / Tuesday, September 9, 2008 / Notices 52421 

leakage rate limits. This license amendment 

request proposes to require the containment 

inservice purge system containment 

penetrations to be blind flanged during 

Modes 1, 2, 3, and 4 and the blind flanged 

penetrations will be required to meet 

containment integrity leakage rate limits. 

With these changes, containment integrity is 

maintained in accordance with the current 

Technical Specification requirements, thus, 

this change does not involve reduction in a 

margin of 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 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: Peter M. Glass, 

Assistant General Counsel, Xcel Energy 

Services, Inc., 414 Nicollet Mall, 

Minneapolis, MN 55401. 

NRC Branch Chief: Lois M. James. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of amendment request: July 30, 

2008. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 3.8.3, 

‘‘Onsite Power Distribution Systems,’’ to 

establish a separate TS Action statement 

for inoperable inverters associated with 

the 120 volt alternating current (VAC) 

distribution panels. The intent of the 

proposed amendment is to extend the 

allowed outage time for inoperable 

inverters from 8 hours to 24 hours. 

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. 

The inverters and associated 120 VAC 

distribution panels are not initiators to any 

accident sequence analyzed in the Updated 

Final Safety Analysis Report (UFSAR). 

The proposed change does not increase the 

number of inverters permitted to be 

inoperable at one time. With one or both 

inverters inoperable in a single channel, 

sufficient capacity and capability remain to 

assure required safety functions can be 

performed. The proposed changes do not 

involve any physical change to structures, 

systems, or components (SSCs) and do not 

alter the method of operation or control of 

SSCs. The current assumptions in the safety 

analysis regarding accident initiators and 

mitigation of accidents are unaffected by 

these proposed changes. The likelihood of 

previously analyzed failures remains 

unchanged. 

Therefore, the proposed change does 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. 

No physical changes will be made to the 

plant or how the plant is operated. As such, 

no new or different kind of accident due to 

a credible new failure mechanism, 

malfunction, or accident initiator will be 

created as a result of this proposed change. 

Any alteration in procedures will continue to 

ensure that the plant remains within 

analyzed limits, and no change is required to 

the procedures relied upon to respond to an 

off-normal event as described in the UFSAR. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change would extend the 

allowed outage time for one or two 

inoperable inverters in a single channel. The 

proposed change does not increase the 

number of inverters permitted to be 

inoperable at one time. There is no change 

to any design basis or safety limits. Operation 

in accordance with the proposed TS ensures 

that the 120 VAC instrument distribution 

system is capable of performing its functions 

as described in the UFSAR. 

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: Jeffrie J. Keenan, 

Esquire, Nuclear Business Unit–N21, 

P.O. Box 236, Hancocks Bridge, NJ 

08038. 

NRC Branch Chief: Harold K. 

Chernoff. 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice. 

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: 

July 13, 2007, as supplemented by 

letters dated July 13, September 12, 

November 19, December 13, and 

December 17, 2007; January 10 (4 

letters), January 11 (4 letters), January 

14, and January 18 (5 letters), January 

31, February 25 (2 letters), March 5, 

March 10 (2 letters), March 25, March 

27, April 4, April 24, April 29, May 15, 

May 20, May 21, July 10, and July 16, 

2008. 

Brief description of amendment: The 

amendment increased the Millstone 

Power Station, Unit No. 3 (MPS3) 

maximum steady-state reactor core 

power level from the previous licensed 

thermal power level of 3,411 megawatts 

thermal (MWt) to 3,650 MWt, which is 

an increase of approximately 7 percent. 

The amendment revises the MPS3 

Operating License and Technical 

Specifications necessary to implement 

the increased power level. 

Date of issuance: August 12, 2008. 

Amendment No.: 242. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Facility Operating License No. NPF–

49: Amendment revised the License and 

Technical Specifications. 

Date of individual notice of issuance 

in Federal Register: August 20, 2008 

(73 FR 49222). 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–454 and STN 50–

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

Ogle County, Illinois 

Date of amendment request: June 17, 

2008. 

Brief description of amendment 

request: The proposed amendment 

would revise Technical Specification 

(TS) 5.5.9, Steam Generator (SG) 

Program, and TS 5.6.9, Steam Generator 

Tube Inspection Report. For TS 5.5.9, 

the amendment would incorporate a 

one-cycle interim alternate repair 

criteria in the provisions for SG tube 


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52422 Federal Register / Vol. 73, No. 175 / Tuesday, September 9, 2008 / Notices 

repair criteria during Byron, Unit No. 2, 

refueling outage 14 and the subsequent 

operating cycle. For TS 5.6.9, the 

amendment would revise the current 

reporting requirements. The proposed 

changes only affect Byron, Unit No. 2; 

however, they are docketed for both 

Byron units because the TSs are 

common to both units. 

Date of publication of individual 

notice in Federal Register: August 5, 

2008 (73 FR 45485). 

Expiration date of individual notice: 

September 5, 2008 (public comment), 

October 5, 2008 (hearing requests). 

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 51.22(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. 

Calvert Cliffs Nuclear Power Plant, Inc., 

Docket No. 50–317, Calvert Cliffs 

Nuclear Power Plant, Unit No. 1, Calvert 

County, Maryland 

Date of application for amendment: 

May 10, 2007, as supplemented by 

letters dated January 10 and July 18, 

2008. 

Brief description of amendment: The 

amendment describes the long-term 

coupon surveillance program for the 

carborundum samples found in the Unit 

No. 1 spent fuel pool (SFP). The 

program verifies that the carborundum 

degradation rates assumed in the 

licensee’s analyses to prove 

subcriticality, as required by Title 10 of 

the Code of Federal Regulations, Section 

50.68, remain valid over the 70-year life 

span of the Unit No. 1 SFP. 

Date of issuance: August 27, 2008. 

Effective date: As of the date of 

issuance to be implemented within 30 

days. 

Amendment No.: 288. 

Renewed Facility Operating License 

No. DPR–53: Amendment revised the 

License and fulfills the requirements 

identified in Appendix C, Additional 

Conditions, to Renewed Facility 

Operating License No. DPR–53 as 

further described in Amendment No. 

267 issued on June 3, 2004. 

Date of initial notice in Federal 

Register: June 19, 2007 (72 FR 33780). 

The letters dated January 10 and July 

18, 2008, 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 August 27, 

2008. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Units 1 and 2, 

Will County, Illinois; Exelon Generation 

Company, LLC, Docket Nos. STN 50–454 

and STN 50–455, Byron Station, Unit 

Nos. 1 and 2, Ogle County, Illinois; 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station, Unit No. 1, DeWitt County, 

Illinois; Exelon Generation Company, 

LLC, Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois; Exelon 

Generation Company, LLC, Docket Nos. 

50–373 and 50–374, LaSalle County 

Station, Units 1 and 2, LaSalle County, 

Illinois; AmerGen Energy Company, 

LLC, et al., Docket No. 50–219, Oyster 

Creek Nuclear Generating Station, 

Ocean County, New Jersey; Exelon 

Generation Company, LLC, and PSEG 

Nuclear LLC, Docket Nos. 50–277 and 

50–278, Peach Bottom Atomic Power 

Station, Units 2 and 3, York and 

Lancaster Counties, Pennsylvania; 

Exelon Generation Company, LLC, 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Units 1 

and 2, Rock Island County, Illinois; 

AmerGen Energy Company, LLC, Docket 

No. 50–289, Three Mile Island Nuclear 

Station, Unit 1, Dauphin County, 

Pennsylvania 

Date of application for amendment: 

July 19, 2007, as supplemented on July 

7, 2008. 

Brief description of amendment: The 

amendments will update the 

requirements in the Technical 

Specifications (TS) 5.3.1 ‘‘Facility Staff 

Qualifications,’’ or TS 6.3.1, ‘‘Unit Staff 

Qualifications,’’ that have been outdated 

based on licensed operator training 

programs accredited by the National 

Academy for Nuclear Training Academy 

Document, ACAD 00–003, Revision 1, 

dated April 2004, and the revised Title 

10 of the Code of Federal Regulations, 

Part 55, ‘‘Operators’ Licenses.’’

Date of issuance: July 25, 2008. 

Effective Date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 152, 152, 156, 156, 

180, 228, 220, 189, 176, 267, 267, 271, 

240, 235, 265 

Facility Operating License Nos. NPF–

72, NPF–77, NPF–37 and NPF–66, NPF–

62, DPR–19, DPR–25, NPF–11, NPF–18, 

DPR–16, DPR–55, DPR–56, DPR–29, 

DPR–30 and DPR–50: The amendments 

revised the Technical Specifications and 

License. 

Date of initial notice in Federal 

Register: December 4, 2007 (72 FR 

68214). The supplemental letter 

contained clarifying information, did 


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not change the initial no significant 

hazards consideration determination, 

and did not expand the scope of the 

original Federal Register notice. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated July 25, 2008. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket No. 50–282, Prairie Island 

Nuclear Generating Plant, Unit 1, 

Goodhue County, Minnesota 

Date of application for amendment: 

August 16, 2007, as supplemented by 

letter dated June 13, 2008. 

Brief description of amendment: The 

amendment revises the Technical 

Specifications (TSs) for Prairie Island 

Nuclear Generating Plant, Unit 1. The 

amendment revises TS 3.8.1 ‘‘AC 

Sources—Operating’’ to require monthly 

testing of the Unit 1 emergency diesel 

generators at or above 2500 kilowatts. 

Date of issuance: August 15, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment No.: 187. 

Facility Operating License No. DPR–

42: Amendment revises the TSs. 

Date of initial notice in Federal 

Register: January 28, 2008 (73 FR 5226). 

The supplemental letter contained 

clarifying information and did not 

change the initial no significant hazards 

consideration determination, and did 

not expand the scope of the original 

Federal Register notice. 

The Commission’s related evaluation 

of the amendment is contained in Safety 

Evaluation dated August 15, 2008. 

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: 

August 16, 2007, as supplemented by 

letter dated June 16, 2008. 

Brief description of amendment: The 

amendment revises the Technical 

Specification (TS) requirements related 

to control room envelope habitability in 

TS 3.7.9, ‘‘Control Room Emergency Air 

Treatment System (CREATS),’’ and TS 

Section 5.5, ‘‘Programs and Manuals.’’

The changes are consistent with the 

Nuclear Regulatory Commission 

approved Industry/Technical 

Specification Task Force Traveler No. 

448, Revision 3. The availability of this 

TS improvement was published in the 

Federal Register on January 17, 2007 

(72 FR 2022), as part of the consolidated 

line item improvement process. 

Date of issuance: August 27, 2008. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment No.: 105. 

Renewed Facility Operating License 

No. DPR–18: Amendment revised the 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: October 23, 2007 (72 FR 

60035). 

The June 16, 2008, supplemental 

letter 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 August 27, 

2008. 

No significant hazards consideration 

comments received: No. 

Southern California Edison Company, et 

al., Docket No. 50–362, San Onofre 

Nuclear Generating Station, Unit 3, San 

Diego County, California 

Date of application for amendments: 

September 24, 2007, as supplemented 

by letters dated February 22 and March 

27, 2008. 

Brief description of amendments: 

Approves the revision to the SONGS 3 

Technical Specification 5.5.2.15, 

‘‘Containment Leakage Rate Testing 

Program,’’ of a one-time extension from 

the currently approved 15-year interval 

since the last Integrated Leak Rate Test 

to a 16-year interval. 

Date of issuance: August 15, 2008. 

Effective date: to be implemented 

within 60 days of issuance. 

Amendment No.: Unit 3–210. 

Facility Operating License No. NPF–

15: The amendments revised the 

Facility Operating Licenses and 

Technical Specifications. 

Date of initial notice in Federal 

Register: October 23, 2007 (72 FR 

60036). The supplements dated 

February 22 and March 27, 2008, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the U.S. 

Nuclear Regulatory Commission staff 

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 August 15, 

2008. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket No. 

50–259, Browns Ferry Nuclear Plant, 

Unit 1, Limestone County, Alabama 

Date of application for amendment: 

March 26, 2008. 

Brief description of amendment: The 

proposed amendment would revise the 

Updated Final Safety Analysis Report 

(UFSAR) to reflect approval to use the 

Boiling Water Reactor Vessel and 

Internals Project reactor pressure vessel 

integrated surveillance program as the 

basis for demonstrating the compliance 

with the requirements of Appendix H to 

Title 10 of the Code of Federal 

Regulations Part 50, ‘‘Reactor Vessel 

Material Surveillance Program 

Requirements.’’

Date of issuance: August 14, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 273. 

Renewed Facility Operating License 

No. DPR–33: Amendment revised the 

UFSAR. 

Date of initial notice in Federal 

Register: June 3, 2008 (73 FR 31723). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated August 14, 

2008. 

No significant hazards consideration 

comments received: No. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of application for amendment: 

August 20, 2007, as supplemented by 

letter dated March 12, 2008. 

Brief description of amendment: The 

amendment revised Technical 

Specification 3.8.3, ‘‘Diesel Fuel Oil, 

Lube Oil, and Starting Air,’’ and its 

associated Surveillance Requirement 

3.8.3.1 to increase the current minimum 

emergency diesel generator (EDG) fuel 

oil inventory required to be maintained 

onsite. The increase in minimum EDG 

fuel oil would provide conservative 

margin against potential vortex effects 

that could occur during fuel oil transfer 

pump operation. 

Date of issuance: August 27, 2008. 

Effective date: As of its date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 185. 

Facility Operating License No. NPF–

30: The amendment revised the 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: September 11, 2007 (72 FR 

51866). The supplemental letter dated 

March 12, 2008, provided additional 


jlentini on PROD1PC65 with NOTICES
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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 as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated August 27, 

2008. 

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, person(s) 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 via electronic 

submission through the NRC E-Filing 

system 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 person(s) 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/readingrm/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 

e-mail 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 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 


jlentini on PROD1PC65 with NOTICES
Federal Register / Vol. 73, No. 175 / Tuesday, September 9, 2008 / Notices 52425 

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. 

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 hearing or a petition for 

leave to intervene must be filed in 

accordance with the NRC E-Filing rule, 

which the NRC promulgated in August 

28, 2007 (72 FR 49139). The E-Filing 

process requires participants to submit 

and serve documents over the internet 

or in some cases to mail copies on 

electronic storage media. Participants 

may not submit paper copies of their 

filings unless they seek a waiver in 

accordance with the procedures 

described below. 

To comply with the procedural 

requirements of E-Filing, at least five (5) 

days prior to the filing deadline, the 

petitioner/requestor must contact the 

Office of the Secretary by e-mail at 

HEARINGDOCKET@NRC.GOV, or by 

calling (301) 415–1677, to request (1) a 

digital ID certificate, which allows the 

participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and/or (2) creation of an 

electronic docket for the proceeding 

(even in instances in which the 

petitioner/requestor (or its counsel or 

representative) already holds an NRCissued digital ID certificate). Each 

petitioner/requestor will need to 

download the Workplace Forms 

ViewerTM to access the Electronic 

Information Exchange (EIE), a 

component of the E-Filing system. The 

Workplace Forms ViewerTM is free and 

is available at http://www.nrc.gov/sitehelp/e-submittals/install-viewer.html. 

Information about applying for a digital 

ID certificate is available on NRC’s 

public Web site at http://www.nrc.gov/ 

site-help/e-submittals/applycertificates.html. 

Once a petitioner/requestor has 

obtained a digital ID certificate, had a 

docket created, and downloaded the EIE 

viewer, it can then submit a request for 

hearing or petition for leave to 

intervene. Submissions should be in 

Portable Document Format (PDF) in 

accordance with NRC guidance 

available on the NRC public Web site at 

http://www.nrc.gov/site-help/esubmittals.html. A filing is considered 

complete at the time the filer submits its 

documents through EIE. To be timely, 

an electronic filing must be submitted to 

the EIE system no later than 11:59 p.m. 

Eastern Time on the due date. Upon 

receipt of a transmission, the E-Filing 

system time-stamps the document and 

sends the submitter an e-mail notice 

confirming receipt of the document. The 

EIE system also distributes an e-mail 

notice that provides access to the 

document to the NRC Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/ 

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically may 

seek assistance through the ‘‘Contact 

Us’’ link located on the NRC Web site 

at http://www.nrc.gov/site-help/esubmittals.html or by calling the NRC 

technical help line, which is available 

between 8:30 a.m. and 4:15 p.m., 

Eastern Time, Monday through Friday. 

The help line number is (800) 397–4209 

or locally, (301) 415–4737. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file a 

motion, in accordance with 10 CFR 

2.302(g), with their initial paper filing 

requesting authorization to continue to 

submit documents in paper format. 

Such filings must be submitted 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; or 

(2) courier, express mail, or expedited 

delivery service to the Office of the 

Secretary, Sixteenth Floor, One White 

Flint North, 11555 Rockville, Pike, 

Rockville, Maryland, 20852, Attention: 

Rulemaking and Adjudications Staff. 

Participants filing a document in this 

manner are responsible for serving the 

document on all other participants. 

Filing is considered complete by firstclass mail as of the time of deposit in 

the mail, or by courier, express mail, or 

expedited delivery service upon 


jlentini on PROD1PC65 with NOTICES
52426 Federal Register / Vol. 73, No. 175 / Tuesday, September 9, 2008 / Notices 

depositing the document with the 

provider of the service. 

Non-timely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission, the presiding officer, or 

the Atomic Safety and Licensing Board 

that the petition and/or request should 

be granted and/or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

CFR 2.309(c)(1)(i)–(viii). To be timely, 

filings must be submitted no later than 

11:59 p.m. Eastern Time on the due 

date. 

Documents submitted in adjudicatory 

proceedings will appear in NRC’s 

electronic hearing docket which is 

available to the public at http:// 

ehd.nrc.gov/EHD_Proceeding/home.asp, 

unless excluded pursuant to an order of 

the Commission, an Atomic Safety and 

Licensing Board, or a Presiding Officer. 

Participants are requested not to include 

personal privacy information, such as 

social security numbers, home 

addresses, or home phone numbers in 

their filings. With respect to copyrighted 

works, except for limited excerpts that 

serve the purpose of the adjudicatory 

filings and would constitute a Fair Use 

application, participants are requested 

not to include copyrighted materials in 

their submission. 

Exelon Generation Company, LLC, 

Docket No. 50–249, Dresden Nuclear 

Power Station, Unit 3, Grundy County, 

Illinois 

Date of amendment request: August 

18, 2008. 

Description of amendment request: 

The amendment revises Technical 

Specification 3.4.5, ‘‘RCS Leakage 

Detection Instrumentation,’’ to support 

implementation of an alternative 

method of verifying that unidentified 

leakage in the drywell is within limits. 

Date of issuance: August 22, 2008. 

Effective date: As of the date of 

issuance and shall be implemented by 

12:00 pm CDT on August 24, 2008. 

Amendment No.: 221. 

Facility Operating License No. DPR–

25: Amendment revises the technical 

specifications and the operating license. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): 

No. On August 17, 2008, the staff 

issued a Notice of Enforcement 

Discretion, which was effective 

immediately and remained in effect 

until this amendment was issued. 

The Commission’s related evaluation 

of the amendment, finding of emergency 

circumstances, state consultation, and 

final NSHC determination are contained 

in a safety evaluation dated August 22, 

2008. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation. 

NRC Branch Chief: Russell Gibbs. 

Dated at Rockville, Maryland, this 29th day 

of August 2008. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E8–20567 Filed 9–8–08; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket Nos. 50–282–LR, 50–306–LR; 

ASLBP No. 08–871–01–LR–BD01] 

Nuclear Management Company, LLC; 

Establishment of Atomic Safety and 

Licensing Board 

Pursuant to delegation by the 

Commission dated December 29, 1972, 

published in the Federal Register, 37 FR 

28,710 (1972), and the Commission’s 

regulations, see 10 CFR 104, 2.300, 

2.303, 2.309, 2.311, 2.318, and 2.321, 

notice is hereby given that an Atomic 

Safety and Licensing Board (Board) is 

being established to preside over the 

following proceeding: 

Nuclear Management Company, LLC 

(Prairie Island Nuclear Generating 

Plant, Units 1 and 2) 

This proceeding involves an 

application for renewal of the licenses 

that authorize Nuclear Management 

Company, LLC to operate Prairie Island 

Nuclear Generating Plant, Units 1 and 2 

for a twenty-year period beyond their 

current expiration dates of, respectively, 

August 9, 2013 and October 29, 2014. In 

response to a June 17, 2008 Notice of 

Acceptance for Docketing of the 

Application and Notice of Opportunity 

for Hearing (73 FR 34,335), a petition to 

intervene has been submitted by Philip 

R. Mahowald on behalf of the Prairie 

Island Indian Community. 

The Board is comprised of the 

following administrative judges: 

William J. Froehlich, Chairman, Atomic 

Safety and Licensing Board Panel, 

U.S. Nuclear Regulatory Commission, 

Washington, DC 20555–0001 

Gary S. Arnold, Atomic Safety and 

Licensing Board Panel, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001 

Thomas J. Hirons, Atomic Safety and 

Licensing Board Panel, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001 

All correspondence, documents, and 

other materials shall be filed in 

accordance with the NRC E-Filing rule, 

which the NRC promulgated in August 

2007 (72 FR 49,139). 

Issued at Rockville, Maryland, this 3rd day 

of September 2008. 

E. Roy Hawkens, 

Chief Administrative Judge, Atomic Safety 

and Licensing Board Panel. 

[FR Doc. E8–20849 Filed 9–8–08; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket No. 50–243; EA–08–251] 

In the Matter of: Oregon State 

University (Oregon State University 

TRIGA Reactor); Order Modifying 

Facility Operating License No. R–106 

I 

Oregon State University (the licensee) 

is the holder of Facility Operating 

License No. R–106 (the license), issued 

by the U.S. Nuclear Regulatory 

Commission (NRC). The NRC plans to 

renew the license on September 10, 

2008. The license authorizes operation 

of the Oregon State University TRIGA 

Reactor (the facility) at a power level up 

to 1,100 kilowatts thermal and in the 

pulse mode, with reactivity insertions 

not to exceed $2.55, and to receive, 

possess, and use special nuclear 

material associated with facility 

operation. The facility is a research 

reactor located on the campus of Oregon 

State University, in the city of Corvallis, 

Benton County, Oregon. The mailing 

address is Radiation Center, Oregon 

State University, 100 Radiation Center, 

Corvallis, Oregon 97331–5903. 

II 

Title 10 of the Code of Federal 

Regulations (10 CFR) Section 50.64, 

limits the use of high-enriched uranium 

(HEU) fuel in domestic non-power 

reactors (research and test reactors) (see 

51 FR 6514). The regulation, which 

became effective on March 27, 1986, 

requires that if Federal Government 

funding for conversion-related costs is 

available, each licensee of a non-power 

reactor authorized to use HEU fuel shall 

replace it with low-enriched uranium 

(LEU) fuel acceptable to the 

Commission unless the Commission has 

determined that the reactor has a unique 

purpose. The Commission’s stated 

purpose for these requirements was to 

reduce, to the maximum extent possible, 

the use of HEU fuel in order to reduce 

the risk of theft and diversion of HEU 

fuel used in non-power reactors. 


jlentini on PROD1PC65 with NOTICES