Federal Register / Vol. 74, No. 14 / Friday, January 23, 2009 / Notices 4247 

principal place of business in Lafayette, 

LA. The patent rights in this invention 

have been assigned to the United States 

of America as represented by the 

Administrator of the National 

Aeronautics and Space Administration. 

The prospective exclusive license will 

comply with the terms and conditions 

of 35 U.S.C. 209 and 37 CFR 404.7. 

NASA has not yet made a determination 

to grant the requested license and may 

deny the requested license even if no 

objections are submitted within the 

comment period. 

DATES: The prospective exclusive 

license may be granted unless, within 

fifteen (15) days from the date of this 

published notice, NASA receives 

written objections including evidence 

and argument that establish that the 

grant of the license would not be 

consistent with the requirements of 35 

U.S.C. 209 and 37 CFR 404.7. 

Competing applications completed and 

received by NASA within fifteen (15) 

days of the date of this published notice 

will also be treated as objections to the 

grant of the contemplated exclusive 

license. 

Objections submitted in response to 

this notice will not be made available to 

the public for inspection and, to the 

extent permitted by law, will not be 

released under the Freedom of 

Information Act, 5 U.S.C. 552. 

ADDRESSES: Objections relating to the 

prospective license may be submitted to 

Mr. James J. McGroary, Chief Patent 

Counsel/LS01, Marshall Space Flight 

Center, Huntsville, AL 35812, (256) 

544–0013. 

FOR FURTHER INFORMATION CONTACT: 

Sammy A. Nabors, Technology Transfer 

Program Office/ED03, Marshall Space 

Flight Center, Huntsville, AL 35812, 

(256) 544–5226. Information about other 

NASA inventions available for licensing 

can be found online at http:// 

technology.nasa.gov. 

Dated: January 15, 2009. 

Richard W. Sherman, 

Acting Deputy General Counsel. 

[FR Doc. E9–1328 Filed 1–22–09; 8:45 am] 

BILLING CODE 7510–13–P 

NATIONAL AERONAUTICS AND 

SPACE ADMINISTRATION 

[Notice (09–007)] 

Aerospace Safety Advisory Panel; 

Meeting 

AGENCY: National Aeronautics and 

Space Administration (NASA). 

ACTION: Notice of meeting. 

SUMMARY: In accordance with the 

Federal Advisory Committee Act, Public 

Law 92–463, as amended, the National 

Aeronautics and Space Administration 

announces a forthcoming meeting of the 

Aerospace Safety Advisory Panel. 

DATES: Wednesday, February 18, 2009, 

1 p.m. to 3 p.m. Eastern Standard Time. 

ADDRESSES: NASA Headquarters, 300 E 

Street, SW., Washington, DC 20546, 

Room 9H40. 

FOR FURTHER INFORMATION CONTACT: Ms. 

Kathy Dakon, Aerospace Safety 

Advisory Panel Executive Director, 

National Aeronautics and Space 

Administration, Washington, DC 20546, 

(202) 358–0732. 

SUPPLEMENTARY INFORMATION: The 

Aerospace Safety Advisory Panel will 

hold its first Quarterly Meeting for 2009. 

This discussion is pursuant to carrying 

out its statutory duties for which the 

Panel reviews, identifies, evaluates, and 

advises on those program activities, 

systems, procedures, and management 

activities that can contribute to program 

risk. Priority is given to those programs 

that involve the safety of human flight. 

The agenda will include Human Capital 

Update, Technical Excellence Overview, 

Human Rating Requirements 

Development, Constellation Program 

Implementation of NASA Human Rating 

Requirements, Office of the Chief 

Engineer Briefing on Human Rating, and 

Exploration Systems Mission 

Directorate Overview. The meeting will 

be open to the public up to the seating 

capacity of the room. Seating will be on 

a first-come basis. Please contact Ms. 

Susan Burch on (202) 358–0550 at least 

48 hours in advance to reserve a seat. It 

is imperative that the meeting be held 

on this date to accommodate the 

scheduling priorities of the key 

participants. Attendees will be required 

to sign a register and to comply with 

NASA security requirements, including 

the presentation of a valid picture ID, 

before receiving an access badge. All 

attendees will need to provide the 

following information to receive an 

access badge: Full name; gender; date/ 

place of birth; citizenship; employer/ 

affiliation information (name of 

institution, address, county, phone), and 

title/position. Foreign Nationals will 

need to provide the following additional 

information: Visa/green card 

information (number, type, expiration 

date). To expedite admittance, attendees 

can provide their identifying 

information in advance by contacting 

Ms. Susan Burch via e-mail at 

susan.burch@nasa.gov or by telephone 

at (202) 358–0550. Persons with 

disabilities who require assistance 

should indicate this. 

Photographs will only be permitted 

during the first 10 minutes of the 

meeting. During the first 30 minutes of 

the meeting, members of the public may 

make a 5-minute verbal presentation to 

the Panel on the subject of safety in 

NASA. To do so, please contact Ms. 

Susan Burch on (202) 358–0550 at least 

48 hours in advance. Any member of the 

public is permitted to file a written 

statement with the Panel at the time of 

the meeting. Verbal presentations and 

written comments should be limited to 

the subject of safety in NASA. 

P. Diane Rausch, 

Advisory Committee Management Officer, 

National Aeronautics and Space 

Administration. 

[FR Doc. E9–1337 Filed 1–22–09; 8:45 am] 

BILLING CODE 7510–13–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2009–0004] 

Notice; Applications and Amendments 

to Facility Operating Licenses 

Involving Proposed No Significant 

Hazards Considerations and 

Containing Sensitive Unclassified NonSafeguards Information or Safeguards 

Information and Order Imposing 

Procedures for Access to Sensitive 

Unclassified Non-Safeguards 

Information or Safeguards Information 

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 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 notice includes notices of 

amendments containing sensitive 

unclassified non-safeguards information 

(SUNSI) or safeguards information 

(SGI). 

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 

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4248 Federal Register / Vol. 74, No. 14 / Friday, January 23, 2009 / Notices 

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, TWB– 

05–B01M, 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. Documents may be examined, 

and/or copied for a fee, at the NRC(s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1 F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. 

Within 60 days after the date of 

publication of this notice, any person(s) 

whose interest may be affected by this 

action may file a request for a hearing 

and a petition to intervene with respect 

to issuance of the amendment to the 

subject facility operating license. 

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, or at 

http://www.nrc.gov/reading-rm/doccollections/cfr/part002/part002– 

0309.html. 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.html. 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. 

All documents filed in NRC 

adjudicatory proceedings, including a 

request for hearing, a petition for leave 

to intervene, any motion or other 

document filed in the proceeding prior 

to the submission of a request for 

hearing or petition to intervene, and 

documents filed by interested 

governmental entities participating 

under 10 CFR 2.315(c), 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 all adjudicatory 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 

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Federal Register / Vol. 74, No. 14 / Friday, January 23, 2009 / Notices 4249 

a waiver in accordance with the 

procedures described below. 

To comply with the procedural 

requirements of E-Filing, at least ten 

(10) days prior to the filing deadline, the 

petitioner/requestor must contact the 

Office of the Secretary by e-mail at 

hearing.docket@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 

Viewer(tm) to access the Electronic 

Information Exchange (EIE), a 

component of the E-Filing system. The 

Workplace Forms Viewer(tm) 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 

electronic filing Help Desk, which is 

available between 8 a.m. and 8 p.m., 

Eastern Time, Monday through Friday. 

The electronic filing Help Desk can be 

contacted by telephone at 1–866–672– 

7640 or by e-mail at 

MSHD.Resource@nrc.gov. 

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 or the presiding officer of 

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). 

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.resource@nrc.gov. 

Entergy Gulf States Louisiana, LLC, and 

Entergy Operations, Inc., Docket No. 50– 

458, River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: 

November 20, 2008. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). The proposed 

amendment revises Technical 

Specification (TS) 5.6.5, ‘‘Core 

Operating Limits Report (COLR),’’ to 

add a reference to an analytical method 

that will be used to determine core 

operating limits. The new reference, 

NEDC–33383P, ‘‘GEXL97 Correlation 

Applicable to ATRIUM–10 Fuel,’’ will 

allow the licensee to use a Global 

Nuclear Fuel method to determine fuel 

assembly critical power of AREVA 

ATRIUM–10 fuel. 

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. 

Core operating limits are established each 

operating cycle in accordance with TS 3.2, 

‘‘Power Distribution’’ and TS 5.6.5, ‘‘Core 

Operating Limits Report (COLR).’’ These core 

operating limits ensure that the fuel design 

limits are not exceeded during any 

conditions of normal operation or in the 

event of any Anticipated Operational 

Occurrence (AOO). The methods used to 

determine the operating limits are those 

previously found acceptable by the NRC and 

listed in TS section 5.6.5.b. 

A change to TS 5.6.5.b is requested to 

include an additional reference to the list of 

analytical methods. RBS [River Bend Station] 

currently operates with a full core of AREVA 

ATRIUM–10 fuel but is scheduled to load 

GE14 fuel during the next refueling outage. 

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4250 Federal Register / Vol. 74, No. 14 / Friday, January 23, 2009 / Notices 

RBS plans to use the analysis methods of the 

new fuel vendor, GNF [Global Nuclear Fuel], 

for the analysis of the mixed core. The 

GEXL97 correlation accurately models 

predicted core behavior and appropriately 

determines the overall critical power 

uncertainty of this method. In addition, the 

GEXL97 application range covers the range of 

expected operation of the ATRIUM–10 fuel 

during normal steady state and transient 

conditions in the RBS reload cores. 

The requested TS changes concern the use 

of analytical methods and do not involve any 

plant modifications or operational changes 

that could affect any postulated accident 

precursors or accident mitigation systems 

and do not introduce any new accident 

initiation mechanisms. The proposed 

changes have no effect on the type or amount 

of radiation released and [have] no effect on 

predicted offsite doses in the event of an 

accident. Thus, the proposed change does not 

affect the probability of an accident 

previously evaluated nor does it increase the 

radiological consequences of any accident 

previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed TS changes will not change 

the design function, reliability, performance, 

or operation of any plant systems, 

components, or structures. It does not create 

the possibility of a new failure mechanism, 

malfunction, or accident initiators not 

considered in the design and licensing bases. 

Plant operation will continue to be within 

the core operating limits that are established 

using NRC approved methods that are 

applicable to the RBS design and the RBS 

fuel. 

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 change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change adds GEXL97 to the 

list of analytical methods in TS 5.6.5.b that 

can be used to determine core operating 

limits. Use of the GEXL97 correlation 

analytical method provides an equivalent 

level of protection as that currently provided. 

The change does not alter any method of 

analysis as described in the NRC approved 

versions of GESTAR–II [NEDE–24011–P-A, 

‘‘General Electric Standard Application for 

Reactor Fuel (GESTAR–II)’’]. The proposed 

change does not modify the safety limits or 

setpoints at which protective actions are 

initiated, and do not change the requirements 

governing operation or availability of safety 

equipment assumed to operate to preserve 

the margin of safety. 

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: Terence A. 

Burke, Associate General Counsel— 

Nuclear Entergy Services, Inc., 1340 

Echelon Parkway, Jackson, Mississippi 

39213. 

NRC Branch Chief: Michael T. 

Markley. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of amendment request: 

December 16, 2008. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). This amendment 

request would revise the Technical 

Specifications (TSs) Section 2.1.2, 

Safety Limit Minimum Critical Power 

Ratio (SLMCPR) for two-loop and 

single-loop operation. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. The proposed changes to Technical 

Specification do not involve a significant 

increase in the probability of an accident 

previously evaluated. 

The proposed Safety Limit MCPR 

(SLMCPR), and its use to determine the 

Operating Cycle 18 thermal limits, have been 

derived using NRC approved methods 

specified in the Reference section of the 

Technical Specification Bases Section for 2.0 

SAFETY LIMITS. These methods do not 

change the method of operating the plant and 

have no effect on the probability of an 

accident initiating event or transient. 

The basis of the SLMCPR is to ensure no 

mechanistic fuel damage is calculated to 

occur if the limit is not violated. The new 

SLMCPR preserves the margin to transition 

boiling, and the probability of fuel damage is 

not increased. 

Therefore, the proposed changes to 

Technical Specifications do not involve an 

increase in the probability or consequences 

of an accident previously evaluated. 

2. The proposed changes to Technical 

Specifications do not create the possibility of 

a new or different kind of accident from any 

accident previously evaluated. 

The proposed changes result only from 

revised methods of analysis for the Cycle 18 

core reload. These methods have been 

reviewed and approved by the NRC, do not 

involve any new or unapproved method for 

operating the facility, and do not involve any 

facility modifications. No new initiating 

events or transients result from these 

changes. 

Therefore, the proposed changes to 

technical specifications do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. The proposed changes to Technical 

Specifications do not involve a significant 

reduction in a margin of safety. 

The margin of safety as defined in the TS 

bases will remain the same. The new 

SLMCPR was derived using NRC approved 

methods which are in accordance with the 

current fuel design and licensing criteria. The 

SLMCPR remains high enough to ensure that 

greater than 99.9% of all fuel rods in the core 

will avoid transition boiling if the limit is not 

violated, thereby preserving the fuel cladding 

integrity. 

Therefore, the proposed changes to 

technical specifications do not involve a 

significant reduction in the margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 400 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Mark G. Kowal. 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, LaSalle 

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Date of amendment request: July 25, 

2008. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). The proposed 

amendments would revise Technical 

Specification 3.3.1.1, ‘‘Reactor 

Protection System (RPS) 

Instrumentation,’’ Surveillance 

Requirement (SR) 3.3.1.1.8 and TS 

3.3.1.3, ‘‘Oscillation Power Range 

Monitor (OPRM) Instrumentation,’’ SR 

3.3.1.3.2 to increase the frequency 

interval between local power range 

monitor calibrations from 1000 effective 

full power hours (EFPH) to 2000 EFPH 

for the LaSalle County Station, Units 1 

and 2 (LSCS). 

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? 

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Federal Register / Vol. 74, No. 14 / Friday, January 23, 2009 / Notices 4251 

Response: No. 

The proposed change is a result of 

increasing the surveillance interval of the 

LPRM [Local Power Range Monitor] 

calibration frequency from 1000 EFPH to 

2000 EFPH. Increasing the frequency interval 

between required LRPM calibrations is 

acceptable due to improvements in the fuel 

analytical bases and therefore, the revised 

surveillance interval continues to ensure that 

the LPRM detector signal is adequately 

calibrated. Extending the LPRM calibration 

surveillance interval will increase the LPRM 

signal uncertainty value used in the LSCS 

SLMCPR [Safety Limit Minimum Critical 

Power Ratio] analysis, however, this increase 

in the LRPM signal uncertainty value is 

acceptable since the increase is bounded by 

the values used by the AREVA analysis. 

This change will not alter the operation of 

process variables, structures, systems, or 

components as described in the LSCS 

Updated Final Safety Analysis Report 

(UFSAR). The proposed change does not alter 

the initiation conditions or operational 

parameters for the system and there is no 

new equipment introduced by the extension 

of the LPRM calibration frequency interval. 

The performance of the Average Power Range 

Monitor (APRM), Rod Block Monitor (RBM) 

and Oscillation Power Range Monitor 

(OPRM) systems are not significantly affected 

by the proposed surveillance interval 

increase. The proposed LPRM calibration 

interval extension will have no significant 

effect on the Reactor Protection System (RPS) 

instrumentation accuracy during power 

maneuvers or transients and will, therefore, 

not significantly affect the performance of the 

RPS. As such, the probability of occurrences 

for a previously evaluated accident is not 

increased. 

The radiological consequences of an 

accident can be affected by the thermal limits 

existing at the time of the postulated 

accident, however, increasing the 

surveillance interval frequency will not 

increase the calculated thermal limits since 

all uncertainties associated with the 

increased interval are currently implemented 

and are currently used to calculate the 

existing Safety Limits. Plant specific 

evaluation of LPRM sensitivity to exposure 

has determined that the extended calibration 

frequency increases the LPRM signal 

uncertainty value used in the LSCS SLMCPR 

analysis, however, the increase is bounded 

by the values currently used in the safety 

analysis. Therefore, the thermal limit 

calculation is not significantly affected by 

LPRM calibration frequency, and thus the 

radiological consequences of any accident 

previously evaluated are not increased. 

Based on the above information, 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. 

The performance of the APRM, RBM, and 

OPRM systems are not significantly affected 

by the proposed LPRM surveillance interval 

increase. The proposed change does not 

affect the control parameters governing unit 

operation or the response of plant equipment 

to transient conditions. For the proposed 

LPRM extended calibration interval 

frequency all uncertainties remain less than 

the uncertainties assumed in the existing 

thermal limit calculations. The proposed 

change does not change or introduce any new 

equipment, modes of system operation or 

failure mechanisms. 

Based on the above information, 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 has no impact on 

equipment design or fundamental operation, 

and there are no changes being made to 

safety limits or safety system allowable 

values that would adversely affect plant 

safety as a result of the proposed LPRM 

surveillance interval increase. The 

performance of the APRM, RBM, and OPRM 

systems are not significantly affected by the 

proposed change. The margin of safety can be 

affected by the thermal limits existing at the 

time of the postulated accident; however, 

uncertainties associated with LPRM chamber 

exposure have no significant effect on the 

calculated thermal limits. Plant specific 

evaluation of LPRM sensitivity to exposure 

has determined that the extended calibration 

frequency increases the LPRM signal 

uncertainty value used in the LSCS SLMCPR 

analysis, however, the increase is bounded 

by the values currently used in the safety 

analysis. The thermal limit calculation is not 

significantly affected since the LPRM 

sensitivity with exposure is well defined. 

LPRM accuracy remains within the total 

nodal power uncertainty assumed in the 

thermal analysis basis, therefore maintaining 

thermal limits and the safety margin. The 

proposed change does not affect safety 

analysis assumptions or initial conditions 

and the margin of safety in the original safety 

analysis are therefore maintained. 

Based on this information, the proposed 

change does not involve a significant 

reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Nuclear, 4300 Winfield Road, 

Warrenville, IL 60555. 

NRC Branch Chief: Russell Gibbs. 

Nebraska Public Power District, Docket 

No. 50–298, Cooper Nuclear Station, 

Nemaha County, Nebraska 

Date of amendment request: October 

13, 2008. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). The proposed 

amendment would revise the licensing 

basis by approving adoption of the 

Alternative Source Term (AST), in 

accordance with 10 CFR 50.67, for use 

in calculating the loss-of-coolant 

accident (LOCA) dose consequences. 

The proposed amendment would revise 

the Technical Specifications (TSs) to (1) 

change the TS definition for DOSE 

EQUIVALENT I–131 to adopt Federal 

Guidance Report (FGR) 11 dose 

conversion factors, (2) require 

operability of the Standby Liquid 

Control (SLC) system in Mode 3, to 

reflect its credit in the LOCA analysis, 

(3) establish a Main Steam (MS) 

Pathway leakage limit that effectively 

increases the previous MS isolation 

valve leakage limit, and (4) change TS 

Section 5.5.12 to reflect a requested 

permanent exemption from the 

requirements of 10 CFR Part 50, 

Appendix J, Option B, Paragraph III.A, 

to allow exclusion of MS Pathway 

leakage from the overall integrated 

leakage rate measured during the 

performance of a Type A test, and from 

the requirements of Appendix J, Option 

B, Paragraph III.B, to allow exclusion of 

the MS Pathway leakage from the 

combined leakage rate of the 

penetrations and valves subject to Type 

B and C tests. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

For the postulated design basis accident 

(DBA) LOCA, the AST is an input to the 

calculations that evaluate the radiological 

consequences of a LOCA. The AST and the 

requested Appendix J exemption do not 

affect the design of the plant or the manner 

in which the plant is normally operated. 

Adoption of the AST and the requested 

Appendix J exemption do not affect the 

initiators of a DBA. Neither the AST nor the 

requested Appendix J exemption [sic] affect 

the response to the DBA LOCA, or the 

pathway of the radiation released from the 

nuclear fuel. Rather, the AST better 

represents the physical characteristics of the 

radiation release. 

Because the initiators of a DBA are not 

affected by adoption of the AST for LOCA 

dose assessment, the probability of an 

accident are not increased by the proposed 

amendment or requested Appendix J 

exemption. 

The AST is an input to calculations used 

to evaluate the radiological consequences of 

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4252 Federal Register / Vol. 74, No. 14 / Friday, January 23, 2009 / Notices 

the LOCA. Use of the AST does not affect the 

plant response to the accident, or the 

pathways to the environment for the 

radiation and activity released from the fuel. 

The LOCA radiological analyses have been 

performed using the AST. Adoption of the 

AST methodology revises the acceptance 

criteria for the accident to the limits specified 

in 10 CFR 50.67. The results of those 

analyses demonstrate that the dose 

consequences are within the acceptance 

criteria presented in 10 CFR 50.67 and in 

NRC RG [Regulatory Guide] 1.183. 

Implementation of the AST for the LOCA 

involves the use of the SLC System to control 

the pH of the suppression pool during 

mitigation of a LOCA. As a result the 

proposed amendment revises the CNS 

[Cooper Nuclear Station] TS for the SLC 

System. These changes do not require any 

physical modification of the plant, nor result 

in any change in normal plant operation. 

This additional use of the SLC system does 

not compromise or adversely affect the 

function of the SLC system as a means of 

shutting down the reactor in addition to the 

control rods. 

Therefore, it is concluded that adoption of 

AST and granting of the Appendix J 

exemption do not involve a significant 

increase in the consequences of an accident 

previously evaluated. Based on the above 

discussion, it is concluded that the proposed 

changes do not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Implementation of the LOCA AST and the 

requested Appendix J exemption do not 

involve a physical alteration of the plant or 

a change in how the plant is normally 

operated. No new or different types of 

equipment will be installed and there are no 

physical modifications to existing equipment 

associated with the proposed changes. The 

proposed changes, effectively increasing the 

allowable MSIV leakage, establishing a 

leakage limit for the MS Pathway, and 

crediting the SLC system for LOCA 

mitigation do not create initiators or 

precursors of a new or different kind of 

accident. New equipment or personnel 

failure modes that might initiate a new type 

of accident are not created as a result of the 

proposed amendment. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

analyzed. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed amendment involves the 

implementation of a new licensing basis for 

the design basis LOCA. Approval of this 

change from the original source term to an 

AST, derived in accordance with the 

guidance of RG 1.183, results in revised 

acceptance criteria for the LOCA analysis. 

For the LOCA, RG 1.183 sets the Exclusion 

Area Boundary (EAB), Low Population Zone 

(LPZ), and Control Room limit consistent 

with 10 CFR 50.67. The AST LOCA 

radiological analysis has been performed 

using conservative methodologies, as 

specified in RG 1.183. Safety margins have 

been evaluated and confirmed to have not 

been reduced. Analytical conservatism has 

been utilized to ensure that the analysis 

adequately bounds the limiting postulated 

event. The dose consequences of the DBA 

LOCA remain within the acceptance criteria 

presented in 10 CFR 50.67 and RG 1.183. 

The proposed changes continue to ensure 

that the doses at the EAB and LPZ boundary, 

as well as the Control Room, are within the 

corresponding regulatory limits. 

Since the proposed amendment continues 

to ensure the doses at the EAB, LPZ and 

Control Room are within corresponding 

regulatory limits, the proposed license 

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: Mr. John C. 

McClure, Nebraska Public Power 

District, Post Office Box 499, Columbus, 

NE 68602–0499. 

NRC Branch Chief: Michael T. 

Markley. 

Nuclear Management Company, LLC, 

Docket No. 50–263, Monticello Nuclear 

Generating Plant, Wright County, 

Minnesota 

Date of amendment request: 

November 5, 2008. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). The licensee 

proposed to increase the current 

maximum power level authorized by 

section 2.C(1) of the renewed facility 

operating license from 1,775 megawatts 

thermal (Mwt) to 2,004 Mwt, an 

approximately 13 percent increase from 

the current licensed thermal power. The 

current maximum power level of 1,775 

Mwt was approved in 1998, an increase 

of 6.3 percent from the original licensed 

thermal power of 1670 Mwt. Thus, 

when approved, the licensee’s proposed 

amendment would take the maximum 

power level to about 20 percent above 

the original license thermal power. The 

licensee’s application addresses in 

details each of the following major 

technical areas: extended power uprate, 

containment analysis methods change, 

credit for containment overpressure for 

low head emergency core cooling 

system (ECCS) pumps, and reactor 

internal pressure differentials (RIPDs) 

for the steam dryer. 

Basis for proposed no significant 

hazards consideration determination: 

As required by Title 10 of the Code of 

Federal Regulations (10 CFR) Part 

50.91(a), the licensee has provided its 

analysis of the issue of no significant 

hazards consideration (NSHC). The 

licensee’s NSHC analysis, addressing 

each technical area listed above, is 

reproduced below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Extended Power Uprate 

Response: No. 

The probability (frequency of occurrence) 

of [d]esign [b]asis [a]ccidents occurring is not 

affected by the increased power level, 

because Monticello Nuclear Generating Plant 

(MNGP) continues to comply with the 

regulatory and design basis criteria 

established for plant equipment. A 

probabilistic risk assessment demonstrates 

that the calculated core damage frequencies 

do not significantly change due to [e]xtended 

[p]ower [u]prate (EPU). Scram setpoints 

(equipment settings that initiate automatic 

plant shutdowns) are established such that 

there is no significant increase in scram 

frequency due to EPU. No new challenges to 

safety-related equipment result from EPU. 

The changes in consequences of postulated 

accidents, which would occur from 102 

percent of the EPU rated thermal power 

(RTP) compared to those previously 

evaluated, are acceptable. The results of EPU 

accident evaluations do not exceed the 

NRC[-]approved acceptance limits. The 

spectrum of postulated accidents and 

transients has been investigated, and are 

shown to meet the plant’s currently licensed 

regulatory criteria. In the area of fuel and 

core design, for example, the Safety Limit 

Minimum Critical Power Ratio (SLMCPR) 

and other applicable Specified Acceptable 

Fuel Design Limits (SAFDL) are still met. 

Continued compliance with the SLMCPR and 

other SAFDLs will be confirmed on a cycle[- 

]specific basis consistent with the criteria 

accepted by the NRC. 

Challenges to the [r]eactor [c]oolant 

[p]ressure [b]oundary were evaluated at EPU 

conditions (pressure, temperature, flow, and 

radiation) and were found to meet their 

acceptance criteria for allowable stresses and 

overpressure margin. 

Challenges to the containment have been 

evaluated, and the containment and its 

associated cooling systems continue to meet 

the current licensing basis. The increase in 

the calculated post[-] LOCA suppression pool 

temperature above the currently assumed 

peak temperature was evaluated and 

determined to be acceptable. Radiological 

release events (accidents) have been 

evaluated, and have been shown to meet the 

guidelines of 10 CFR 50.67. 

Containment Analysis Methods Change 

Response: No. 

The use of passive heat sinks, variable RHR 

[residual heat removal] heat exchanger 

capability K-value, and mechanistic heat and 

mstockstill on PROD1PC66 with NOTICES

Federal Register / Vol. 74, No. 14 / Friday, January 23, 2009 / Notices 4253 

mass transfer from the suppression pool 

surface to the wetwell airspace after 30 

seconds for the long[-]term design[-] 

basis[-]accident loss[-]of[-]coolant accident 

(DBA–LOCA) containment analysis are not 

relevant to accident initiation, but rather, 

pertain to the method used to accurately 

evaluate postulated accidents. The use of 

these elements does not, in any way, alter 

existing fission product boundaries, and 

provides a conservative prediction of the 

containment response to DBA–LOCAs. 

Therefore, the containment analysis method 

change does not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

Credit for Containment Overpressure for Low 

Head Emergency Core Cooling System (ECCS) 

Pumps 

Response: No. 

These changes update parameters used in 

the MNGP safety analyses and expand the 

range and scope of the analyses. This will 

result in a more realistic analysis of available 

containment overpressure under design[- 

]basis accident conditions. The updated 

analyses affect only the evaluation of 

previously reviewed accidents. No plant 

structure, system, or component (SSC) is 

physically affected by the updated and 

expanded analyses. No method of operation 

of any plant SSC is affected. Therefore, there 

is no significant increase in the probability or 

consequence of a previously evaluated 

accident. 

Reactor Internal Pressure Differentials 

(RIPDs) for the Steam Dryer 

Response: No. 

The revised steam dryer RIPDs are used in 

evaluating loads in reactor vessel internals 

for various conditions (i.e., during normal, 

upset and faulted conditions). The values 

more accurately represent the actual plant 

configuration. No plant structure, system, or 

component (SSC) is physically affected by 

the updated and expanded analyses. No 

method of operation of any plant SSC is 

affected. Therefore, there is no significant 

increase in the probability or consequence of 

a previously evaluated accident. 

The analyses supporting the above 

evaluations were performed at the EPU 

power level of 2,004 Mwt. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Extended Power Uprate 

Response: No. 

Equipment that could be affected by EPU 

has been evaluated. No new operating mode, 

safety-related equipment lineup, accident 

scenario, or equipment failure mode was 

identified. The full spectrum of accident 

considerations has been evaluated and no 

new or different kind of accident has been 

identified. EPU uses developed technology 

and applies it within capabilities of existing 

or modified plant safety[-]related equipment 

in accordance with the regulatory criteria 

(including NRC[-]approved codes, standards 

and methods). No new accidents or event 

precursors have been identified. 

The MNGP TS require revision to 

implement EPU. The revisions have been 

assessed and it was determined that the 

proposed change will not introduce a 

different accident than that previously 

evaluated. Therefore, the proposed changes 

do not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

Containment Analysis Methods Change 

Response: No. 

The use of passive heat sinks, variable RHR 

heat exchanger capability K-value, and 

mechanistic heat and transfer from the 

suppression pool surface to the wetwell 

airspace after 30 seconds for the long[-]term 

DBA–LOCA containment analysis are not 

relevant to accident initiation, but pertain to 

the method used to evaluate currently 

postulated accidents. The use of these 

analytical tools does not involve any physical 

changes to plant structures or systems, and 

does not create a new initiating event for the 

spectrum of events currently postulated. 

Further, they do not result in the need to 

postulate any new accident scenarios. 

Therefore, the containment analysis method 

change does not create the possibility of a 

new or different kind of accident from any 

accident previously evaluated. 

Credit for Containment Overpressure for Low 

Head ECCS Pumps 

Response: No. 

The proposed change involves the 

updating and expansion in scope of the 

existing design bases analysis with respect to 

the available containment overpressure to 

cover additional events. No new failure mode 

or mechanisms have been created for any 

plant SSC important to safety nor has any 

new limiting single failure been identified as 

a result of the proposed analytical changes. 

Therefore, the change to containment 

overpressure credited for low pressure ECCS 

pumps does not create the possibility of a 

new or different kind of accident from any 

accident previously evaluated. 

Reactor Internal Pressure Differentials for the 

Steam Dryer 

Response: No. 

The revised steam dryer RIPDs are used in 

evaluating loads in reactor vessel internals 

for various conditions (i.e., during normal, 

upset and faulted conditions). The steam 

dryer RIPDs are not relevant to accident 

initiation, but only pertain to the method 

used to evaluate reactor vessel internals 

loads. The revised steam dryer RIPD values 

more accurately represent the actual plant 

configuration. Therefore, the change to steam 

dryer RIPDs does not create the possibility of 

a new or different kind of accident from any 

accident previously evaluated. 

The analyses supporting the above 

evaluations were performed at the EPU 

power level of 2,004 Mwt. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Extended Power Uprate 

Response: No. 

The EPU affects only design and 

operational margins. Challenges to the fuel, 

reactor coolant pressure boundary, and 

containment were evaluated for EPU 

conditions. Fuel integrity is maintained by 

meeting existing design and regulatory limits. 

The calculated loads on affected structures, 

systems and components, including the 

reactor coolant pressure boundary, will 

remain within their design allowables for 

design[-]basis event categories. No NRC 

acceptance criterion is exceeded. Because the 

MNGP configuration and responses to 

transients and postulated accidents do not 

result in exceeding the presently approved 

NRC acceptance limits, the proposed changes 

do not involve a significant reduction in a 

margin of safety. 

Containment Analysis Methods Change 

Response: No. 

The use of passive heat sinks, variable RHR 

heat exchanger capability K-value, and 

mechanistic heat and mass transfer from the 

suppression pool surface to the wetwell 

airspace after 30 seconds for the long[-]term 

DBA–LOCA containment analysis are 

realistic phenomena and provide a 

conservative prediction of the plant response 

to DBA–LOCAs. The increase in pressure and 

temperature are relatively small and are 

within design limits. Therefore, the 

containment analysis methods change does 

not involve a significant reduction in the 

margin of safety. 

Increase in Credit for Containment 

Overpressure for Low Head ECCS Pumps 

Response: No. 

The proposed changes revise containment 

response analytical methods and scope for 

containment pressure to assist in ECCS pump 

net positive suction head (NPSH). The 

changes are still based on conservative but 

more realistic analysis of available 

containment overpressure determined using 

analysis methods that minimize containment 

pressure and maximize suppression pool 

temperature. These changes do not constitute 

a significant reduction in the margin of 

safety. 

Reactor Internal Pressure Differentials for the 

Steam Dryer 

Response: No. 

The revised steam dryer RIPDs are used in 

evaluating loads in reactor vessel internals 

for various conditions (i.e., during normal, 

upset and faulted conditions). The revised 

steam dryer RIPD values more accurately 

represent the actual plant configuration. The 

changes are still conservative but more 

accurately represent the MNGP 

configuration. These changes do not 

constitute a significant reduction in the 

margin of safety. 

The analyses supporting the above 

evaluations were performed at the EPU 

power level of 2,004 Mwt. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

mstockstill on PROD1PC66 with NOTICES

4254 Federal Register / Vol. 74, No. 14 / Friday, January 23, 2009 / Notices 

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 proposed 

amendment involves 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. 

PPL Susquehanna, LLC, Docket No. 50– 

388, Susquehanna Steam Electric 

Station, Unit 2, Luzerne County, 

Pennsylvania 

Date of amendment request: October 

30, 2008. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). This amendment 

request would revise PPL Susquehanna, 

LLC, Unit 2 (PPL) Technical 

Specifications (TSs) Section 2.1.1.2, 

Minimum Critical Power Ratio Safety 

Limits (MCPRSLs) for two-loop and 

single-loop operation. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change to the two-loop and 

single-loop MCPRSLs do not directly or 

indirectly affect any plant system, 

equipment, component, or change the 

processes used to operate the plant. Further, 

the proposed MCPRSLs were generated using 

NRC approved methodology and meet the 

applicable acceptance criteria. Thus, this 

proposed amendment does not involve a 

significant increase in the probability of 

occurrence or consequences of an accident 

previously evaluated. 

Prior to the startup of U2C15, licensing 

analyses are performed (using NRC approved 

methodology referenced in TS Section 

5.6.5.b) to determine changes in the CPR as 

a result of anticipated operational 

occurrences. These results are added to the 

MCPRSL values to generate the MCPROLs in 

the COLR [Core Operating Limits Report]. 

These limits could be different from those 

specified for the previous Unit 2 COLR. The 

COLR operating limits thus assure that the 

MCPRSL will not be exceeded during normal 

operation or AOOs [anticipated operational 

occurrences]. Postulated accidents are also 

analyzed prior to the startup and the results 

shown to be within the NRC approved 

criteria. 

Therefore, this proposed amendment does 

not involve a significant increase in the 

probability of occurrence 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 changes to the two-loop and singleloop MCPRSLs do not directly or indirectly 

affect any plant system, equipment, or 

component and therefore does not affect the 

failure modes of any of these items. Thus, the 

proposed change does not create the 

possibility of a previously unevaluated 

operator error or a new single failure. 

Therefore, this 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. 

Since the proposed changes do not alter 

any plant system, equipment, component, or 

processes used to operate the plant, the 

proposed change will not jeopardize or 

degrade the function or operation of any 

plant system or component governed by TS. 

The proposed two-loop and single-loop 

MCPRSLs do not involve a significant 

reduction in the margin of safety as currently 

defined in the Bases of the applicable TS 

sections, because the proposed MCPRSLs 

preserve the required 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 request involves no 

significant hazards consideration. 

Attorney for licensee: Bryan A. Snapp, 

Esquire, Assoc. General Counsel, PPL 

Services Corporation, 2 North Ninth St., 

GENTW3, Allentown, PA 18101–1179. 

NRC Branch Chief: Mark G. Kowal. 

Order Imposing Procedures for Access 

to Sensitive Unclassified NonSafeguards Information (SUNSI) and 

Safeguards Information (SGI) for 

Contention Preparation 

Entergy Gulf States Louisiana, LLC, and 

Entergy Operations, Inc., Docket No. 50– 

458, River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, LaSalle 

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Nebraska Public Power District, Docket 

No. 50–298, Cooper Nuclear Station, 

Nemaha County, Nebraska 

Nuclear Management Company, LLC, 

Docket No. 50–263, Monticello Nuclear 

Generating Plant, Wright County, 

Minnesota 

PPL Susquehanna, LLC, Docket No. 50– 

388, Susquehanna Steam Electric 

Station, Unit 2, Luzerne County, 

Pennsylvania 

1. This order contains instructions 

regarding how potential parties to the 

proceedings listed above may request 

access to documents containing 

sensitive unclassified information 

(SUNSI and SGI). 

2. Within ten (10) days after 

publication of this notice of opportunity 

for hearing, any potential party as 

defined in 10 CFR 2.4 who believes 

access to SUNSI or SGI is necessary for 

a response to the notice may request 

access to SUNSI or SGI. A ‘‘potential 

party’’ is any person who intends or 

may intend to participate as a party by 

demonstrating standing and the filing of 

an admissible contention under 10 CFR 

2.309. Requests submitted later than ten 

(10) days will not be considered absent 

a showing of good cause for the late 

filing, addressing why the request could 

not have been filed earlier. 

3. The requester shall submit a letter 

requesting permission to access SUNSI 

and/or SGI to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

Washington, DC 20555–0001, Attention: 

Rulemakings and Adjudications Staff, 

and provide a copy to the Associate 

General Counsel for Hearings, 

Enforcement and Administration, Office 

of the General Counsel, Washington, DC 

20555–0001. The expedited delivery or 

courier mail address for both offices is 

U.S. Nuclear Regulatory Commission, 

11555 Rockville Pike, Rockville, MD 

20852. The e-mail address for the Office 

of the Secretary and the Office of the 

mstockstill on PROD1PC66 with NOTICES

Federal Register / Vol. 74, No. 14 / Friday, January 23, 2009 / Notices 4255 

1See footnote 6. While a request for hearing or 

petition to intervene in this proceeding must 

comply with the filing requirements of the NRC’s 

‘‘E-Filing Rule,’’ the initial request to access SUNSI 

and/or SGI under these procedures should be 

submitted as described in this paragraph. 

2The requester will be asked to provide his or her 

full name, Social Security number, date and place 

of birth, telephone number, and e-mail address. 

After providing this information, the requester 

usually should be able to obtain access to the online 

form within one business day. 

3Broad SGI requests under these procedures are 

thus highly unlikely to meet the standard for need 

to know; furthermore, staff redaction of information 

from requested documents before their release may 

be appropriate to comport with this requirement. 

These procedures do not authorize unrestricted 

disclosure or less scrutiny of a requester’s need to 

know than ordinarily would be applied in 

connection with an already-admitted contention. 

4 If a presiding officer has not yet been 

designated, the Chief Administrative Judge will 

issue such orders, or will appoint a presiding officer 

to do so. 

5Parties/persons other than the requester and the 

NRC staff will be notified by the NRC staff of a 

favorable access determination (and may participate 

in the development of such a motion and protective 

order) if it concerns SUNSI and if the party/person’s 

interest independent of the proceeding would be 

harmed by the release of the information (e.g., as 

with proprietary information). 

General Counsel are 

hearing.docket@nrc.gov and 

ogcmailcenter.resource@nrc.gov, 

respectively.1 The request must include 

the following information: 

a. A description of the licensing 

action with a citation to this Federal 

Register notice of opportunity for 

hearing; 

b. The name and address of the 

potential party and a description of the 

potential party’s particularized interest 

that could be harmed by the action 

identified in (a); 

c. If the request is for SUNSI, the 

identity of the individual requesting 

access to SUNSI and the requester’s 

need for the information in order to 

meaningfully participate in this 

adjudicatory proceeding, particularly 

why publicly available versions of the 

application would not be sufficient to 

provide the basis and specificity for a 

proffered contention; 

d. If the request is for SGI, the identity 

of the individual requesting access to 

SGI and the identity of any expert, 

consultant or assistant who will aid the 

requester in evaluating the SGI, and 

information that shows: 

(i) Why the information is 

indispensable to meaningful 

participation in this licensing 

proceeding; and 

(ii) The technical competence 

(demonstrable knowledge, skill, 

experience, training or education) of the 

requester to understand and use (or 

evaluate) the requested information to 

provide the basis and specificity for a 

proffered contention. The technical 

competence of a potential party or its 

counsel may be shown by reliance on a 

qualified expert, consultant or assistant 

who demonstrates technical competence 

as well as trustworthiness and 

reliability, and who agrees to sign a nondisclosure affidavit and be bound by the 

terms of a protective order; and 

e. If the request is for SGI, Form SF– 

85, ‘‘Questionnaire for Non-Sensitive 

Positions,’’ Form FD–258 (fingerprint 

card), and a credit check release form 

completed by the individual who seeks 

access to SGI and each individual who 

will aid the requester in evaluating the 

SGI. For security reasons, Form SF–85 

can only be submitted electronically, 

through a restricted-access database. To 

obtain online access to the form, the 

requester should contact the NRC’s 

Office of Administration at 301–492– 

3524.2 The other completed forms must 

be signed in original ink, accompanied 

by a check or money order payable in 

the amount of $191.00 to the U.S. 

Nuclear Regulatory Commission for 

each individual, and mailed to the: 

Office of Administration, Security 

Processing Unit, Mail Stop TWB–05– 

B32M, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0012. 

These forms will be used to initiate 

the background check, which includes 

fingerprinting as part of a criminal 

history records check. Note: copies of 

these forms do not need to be included 

with the request letter to the Office of 

the Secretary, but the request letter 

should state that the forms and fees 

have been submitted as described above. 

4. To avoid delays in processing 

requests for access to SGI, all forms 

should be reviewed for completeness 

and accuracy (including legibility) 

before submitting them to the NRC. 

Incomplete packages will be returned to 

the sender and will not be processed. 

5. Based on an evaluation of the 

information submitted under items 2 

and 3.a through 3.d, above, the NRC 

staff will determine within ten days of 

receipt of the written access request 

whether (1) there is a reasonable basis 

to believe the petitioner is likely to 

establish standing to participate in this 

NRC proceeding, and (2) there is a 

legitimate need for access to SUNSI or 

need to know the SGI requested. For 

SGI, the need to know determination is 

made based on whether the information 

requested is necessary (i.e., 

indispensable) for the proposed 

recipient to proffer and litigate a 

specific contention in this NRC 

proceeding 3 and whether the proposed 

recipient has the technical competence 

(demonstrable knowledge, skill, 

training, education, or experience) to 

evaluate and use the specific SGI 

requested in this proceeding. 

6. If standing and need to know SGI 

are shown, the NRC staff will further 

determine based upon completion of the 

background check whether the proposed 

recipient is trustworthy and reliable. 

The NRC staff will conduct (as 

necessary) an inspection to confirm that 

the recipient’s information protection 

systems are sufficient to protect SGI 

from inadvertent release or disclosure. 

Recipients may opt to view SGI at the 

NRC’s facility rather than establish their 

own SGI protection program to meet SGI 

protection requirements. 

7. A request for access to SUNSI or 

SGI will be granted if: 

a. The request has demonstrated that 

there is a reasonable basis to believe that 

a potential party is likely to establish 

standing to intervene or to otherwise 

participate as a party in this proceeding; 

b. The proposed recipient of the 

information has demonstrated a need for 

SUNSI or a need to know for SGI, and 

that the proposed recipient of SGI is 

trustworthy and reliable; 

c. The proposed recipient of the 

information has executed a NonDisclosure Agreement or Affidavit and 

agrees to be bound by the terms of a 

Protective Order setting forth terms and 

conditions to prevent the unauthorized 

or inadvertent disclosure of SUNSI and/ 

or SGI; and 

d. The presiding officer has issued a 

protective order concerning the 

information or documents requested.4

Any protective order issued shall 

provide that the petitioner must file 

SUNSI or SGI contentions 25 days after 

receipt of (or access to) that information. 

However, if more than 25 days remain 

between the petitioner’s receipt of (or 

access to) the information and the 

deadline for filing all other contentions 

(as established in the notice of hearing 

or opportunity for hearing), the 

petitioner may file its SUNSI or SGI 

contentions by that later deadline. 

8. If the request for access to SUNSI 

or SGI is granted, the terms and 

conditions for access to sensitive 

unclassified information will be set 

forth in a draft protective order and 

affidavit of non-disclosure appended to 

a joint motion by the NRC staff, any 

other affected parties to this 

proceeding,5 and the petitioner(s). If the 

diligent efforts by the relevant parties or 

petitioner(s) fail to result in an 

agreement on the terms and conditions 

for a draft protective order or nondisclosure affidavit, the relevant parties 

mstockstill on PROD1PC66 with NOTICES

4256 Federal Register / Vol. 74, No. 14 / Friday, January 23, 2009 / Notices 

6As of October 15, 2007, the NRC’s final ‘‘EFiling Rule’’ became effective. See Use of Electronic 

Submissions in Agency Hearings (72 FR 49139; 

Aug. 28, 2007). Requesters should note that the 

filing requirements of that rule apply to appeals of 

NRS staff determinations (because they must be 

served on a presiding officer or the Commission, as 

applicable), but not to the initial SUNSI/SGI 

requests submitted to the NRC staff under these 

procedures. 

to the proceeding or the petitioner(s) 

should notify the presiding officer 

within ten (10) days, describing the 

obstacles to the agreement. 

9. If the request for access to SUNSI 

is denied by the NRC staff or a request 

for access to SGI is denied by NRC staff 

either after a determination on standing 

and need to know or, later, after a 

determination on trustworthiness and 

reliability, the NRC staff shall briefly 

state the reasons for the denial. Before 

the Office of Administration makes an 

adverse determination regarding access, 

the proposed recipient must be 

provided an opportunity to correct or 

explain information. The requester may 

challenge the NRC staff’s adverse 

determination with respect to access to 

SUNSI or with respect to standing or 

need to know for SGI by filing a 

challenge within ten (10) days of receipt 

of that determination with (a) the 

presiding officer designated in this 

proceeding; (b) if no presiding officer 

has been appointed, the Chief 

Administrative Judge, or if he or she is 

unavailable, another administrative 

judge, or an administrative law judge 

with jurisdiction pursuant to 10 CFR 

2.318(a); or (c) if another officer has 

been designated to rule on information 

access issues, with that officer. In the 

same manner, an SGI requester may 

challenge an adverse determination on 

trustworthiness and reliability by filing 

a challenge within fifteen (15) days of 

receipt of that determination. 

In the same manner, a party other 

than the requester may challenge an 

NRC staff determination granting access 

to SUNSI whose release would harm 

that party’s interest independent of the 

proceeding. Such a challenge must be 

filed within ten (10) days of the 

notification by the NRC staff of its grant 

of such a request. 

If challenges to the NRC staff 

determinations are filed, these 

procedures give way to the normal 

process for litigating disputes 

concerning access to information. The 

availability of interlocutory review by 

the Commission of orders ruling on 

such NRC staff determinations (whether 

granting or denying access) is governed 

by 10 CFR 2.311.6

10. The Commission expects that the 

NRC staff and presiding officers (and 

any other reviewing officers) will 

consider and resolve requests for access 

to SUNSI and/or SGI, and motions for 

protective orders, in a timely fashion in 

order to minimize any unnecessary 

delays in identifying those petitioners 

who have standing and who have 

propounded contentions meeting the 

specificity and basis requirements in 10 

CFR Part 2. Attachment 1 to this Order 

summarizes the general target schedule 

for processing and resolving requests 

under these procedures. 

Dated at Rockville, Maryland, this 13th day 

of January 2009. 

For the Nuclear Regulatory Commission. 

Annette L. Vietti-Cook, 

Secretary of the Commission. 

ATTACHMENT 1—GENERAL TARGET SCHEDULE FOR PROCESSING AND RESOLVING REQUESTS FOR ACCESS TO SENSITIVE

UNCLASSIFIED NON-SAFEGUARDS INFORMATION (SUNSI) AND SAFEGUARDS INFORMATION (SGI) IN THIS PROCEEDING

Day Event/activity 

0 ............... Publication of Federal Register notice/other notice of proposed action and opportunity for hearing, including order with instructions 

for access requests. 

10 ............. Deadline for submitting requests for access to SUNSI and/or SGI with information: supporting the standing of a potential party identified by name and address; describing the need for the information in order for the potential party to participate meaningfully in 

an adjudicatory proceeding; demonstrating that access should be granted (e.g., showing technical competence for access to 

SGI); and, for SGI, including application fee for fingerprint/background check. 

60 ............. Deadline for submitting petition for intervention containing: (i) Demonstration of standing; (ii) all contentions whose formulation does 

not require access to SUNSI and/or SGI (+25 Answers to petition for intervention; +7 petitioner/requestor reply). 

20 ............. NRC staff informs the requester of the staff’s determination whether the request for access provides a reasonable basis to believe 

standing can be established and shows (1) need for SUNSI or (2) need to know for SGI. (For SUNSI, NRC staff also informs any 

party to the proceeding whose interest independent of the proceeding would be harmed by the release of the information.) If 

NRC staff makes the finding of need for SUNSI and likelihood of standing, NRC staff begins document processing (preparation 

of redactions or review of redacted documents). If NRC staff makes the finding of need to know for SGI and likelihood of standing, NRC staff begins background check (including fingerprinting for a criminal history records check), information processing 

(preparation of redactions or review of redacted documents), and readiness inspections. 

25 ............. If NRC staff finds no ‘‘need,’’ ‘‘need to know,’’ or likelihood of standing, the deadline for petitioner/requester to file a motion seeking 

a ruling to reverse the NRC staff’s denial of access; NRC staff files copy of access determination with the presiding officer (or 

Chief Administrative Judge or other designated officer, as appropriate). If NRC staff finds ‘‘need’’ for SUNSI, the deadline for any 

party to the proceeding whose interest independent of the proceeding would be harmed by the release of the information to file a 

motion seeking a ruling to reverse the NRC staff’s grant of access. 

30 ............. Deadline for NRC staff reply to motions to reverse NRC staff determination(s). 

40 ............. (Receipt +30) If NRC staff finds standing and need for SUNSI, deadline for NRC staff to complete information processing and file 

motion for Protective Order and draft Non-Disclosure Affidavit. Deadline for applicant/licensee to file Non-Disclosure Agreement 

for SUNSI. 

190 ........... (Receipt +180) If NRC staff finds standing, need to know for SGI, and trustworthiness and reliability, deadline for NRC staff to file 

motion for Protective Order and draft Non-disclosure Affidavit (or to make a determination that the proposed recipient of SGI is 

not trustworthy or reliable). Note: Before the Office of Administration makes an adverse determination regarding access, the proposed recipient must be provided an opportunity to correct or explain information. 

205 ........... Deadline for petitioner to seek reversal of a final adverse NRC staff determination either before the presiding officer or another designated officer. 

A .............. If access granted: Issuance of presiding officer or other designated officer decision on motion for protective order for access to sensitive information (including schedule for providing access and submission of contentions) or decision reversing a final adverse 

determination by the NRC staff. 

A + 3 ........ Deadline for filing executed Non-Disclosure Affidavits. Access provided to SUNSI and/or SGI consistent with decision issuing the 

protective order. 

mstockstill on PROD1PC66 with NOTICES

Federal Register / Vol. 74, No. 14 / Friday, January 23, 2009 / Notices 4257 

ATTACHMENT 1—GENERAL TARGET SCHEDULE FOR PROCESSING AND RESOLVING REQUESTS FOR ACCESS TO SENSITIVE

UNCLASSIFIED NON-SAFEGUARDS INFORMATION (SUNSI) AND SAFEGUARDS INFORMATION (SGI) IN THIS PROCEEDING—Continued 

Day Event/activity 

A + 28 ...... Deadline for submission of contentions whose development depends upon access to SUNSI and/or SGI. However, if more than 25 

days remain between the petitioner’s receipt of (or access to) the information and the deadline for filing all other contentions (as 

established in the notice of hearing or opportunity for hearing), the petitioner may file its SUNSI or SGI contentions by that later 

deadline. 

A + 53 ...... (Contention receipt +25) Answers to contentions whose development depends upon access to SUNSI and/or SGI. 

A + 60 ...... (Answer receipt +7) Petitioner/Intervenor reply to answers. 

B .............. Decision on contention admission. 

[FR Doc. E9–1152 Filed 1–22–09; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket No. 52–037; NRC–2008–0556] 

Union Electric Company d/b/a Ameren 

UE Callaway Plant Unit 2 Combined 

License Application; Notice of Intent 

To Prepare an Environmental Impact 

Statement and Conduct Scoping 

Process 

Union Electric Company d/b/a 

Ameren UE (AmerenUE) has submitted 

an application for a combined license 

(COL) to build and operate Unit 2 at its 

Callaway Plant site, located on 

approximately 2,800 acres 10 miles 

southeast of the city of Fulton in 

Callaway County, Missouri, and 80 

miles west of the St. Louis metropolitan 

area. AmerenUE submitted the 

application for the COL to the U.S. 

Nuclear Regulatory Commission (NRC) 

by letter dated July 24, 2008, pursuant 

to Title 10 of the Code of Federal 

Regulations (10 CFR) Part 52. A notice 

of receipt and availability of the 

application, including the 

environmental report (ER), was 

published in the Federal Register on 

October 9, 2008 (73 FR 59677). A notice 

of acceptance for docketing of the 

application for the COL was published 

in the Federal Register on December 18, 

2008 (73 FR 77078). A notice of hearing 

and opportunity to petition for leave to 

intervene in the proceeding of the 

application will be published in a future 

Federal Register. The purpose of this 

notice is to inform the public that the 

NRC staff will be preparing an 

environmental impact statement (EIS) as 

part of the review of the application for 

the COL, and to provide the public with 

an opportunity to participate in the 

environmental scoping process as 

defined in 10 CFR 51.29. The U.S. Army 

Corps of Engineers (Corp), Kansas City 

District, has requested to participate in 

the preparation of the EIS as a 

cooperating agency; the NRC has 

accepted their request. The agencies 

will cooperate according to the process 

set forth in the MOU signed by the NRC 

and the Corps, and was published in the 

Federal Register on September 25, 2008 

(73 FR 55546). 

In addition, as outlined in 36 CFR 

800.8(c), ‘‘Coordination with the 

National Environmental Policy Act,’’ the 

NRC staff plans to coordinate 

compliance with Section 106 of the 

National Historic Preservation Act 

(NHPA) with steps taken to meet the 

requirements of the National 

Environmental Policy Act of 1969, as 

amended (NEPA). Pursuant to 36 CFR 

800.8(c), the NRC staff intends to use 

the process and documentation for the 

preparation of the EIS on the proposed 

action to comply with Section 106 of the 

NHPA in lieu of the procedures set forth 

in 36 CFR 800.3 through 800.6. 

In accordance with 10 CFR 51.45 and 

51.50, AmerenUE submitted the ER as 

part of the application. The ER was 

prepared pursuant to 10 CFR Parts 51 

and 52 and is available for public 

inspection at the NRC Public Document 

Room (PDR) located at One White Flint 

North, 11555 Rockville Pike (first floor), 

Rockville, Maryland 20852 or from the 

Publicly Available Records (PAR) 

component of NRC’s Agency-wide 

Documents Access and Management 

System (ADAMS). ADAMS is accessible 

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

adams.html, which provides access 

through the NRC’s Electronic Reading 

Room (ERR) link. The accession number 

in ADAMS for the environmental report 

included in the application is 

ML082520869. Persons who do not have 

access to ADAMS or who encounter 

problems in accessing the documents 

located in ADAMS should contact the 

NRC’s PDR Reference staff at 1–800– 

397–4209/301–415–4737 or by e-mail to 

pdr@nrc.gov. The application may also 

be viewed on the Internet at http:// 

www.nrc.gov/reactors/new-reactors/col/ 

callaway.html. In addition, the Callaway 

County Public Library, 710 Court Street, 

Fulton, MO 65251; and Ellis Library in 

University of Missouri, 106–B Ellis 

Library, Columbia, MO 65201–5149 

have agreed to make the ER available for 

public inspection. The following key 

reference documents related to the 

application and the NRC staff’s review 

processes are available through the 

NRC’s Web site at http://www.nrc.gov: 

a. 10 CFR Part 51, Environmental 

Protection Regulations for Domestic 

Licensing and Related Regulatory 

Functions; 

b. 10 CFR Part 52, Licenses, 

Certifications, and Approvals for 

Nuclear Power Plants; 

c. 10 CFR Part 100, Reactor Site 

Criteria; 

d. NUREG–1555, Standard Review 

Plans for Environmental Reviews for 

Nuclear Power Plants; 

e. NUREG/BR–0298, Brochure on 

Nuclear Power Plant Licensing Process; 

f. Regulatory Guide 4.2, Preparation of 

Environmental Reports for Nuclear 

Power Stations; 

g. Regulatory Guide 4.7, General Site 

Suitability Criteria for Nuclear Power 

Stations; 

h. Fact Sheet on Nuclear Power Plant 

Licensing Process; 

i. Regulatory 1.206, Combined License 

Applications for Nuclear Power Plants; 

and 

j. Nuclear Regulatory Commission 

Policy Statement on the Treatment of 

Environmental Justice Matters in NRC 

Regulatory and Licensing Actions. 

The regulations, NUREG-series 

documents, regulatory guides, and the 

fact sheet can be found under Document 

Collections in the ERR on the NRC Web 

page. The environmental justice policy 

Statement can be found in the Federal 

Register, 69 FR 52040 August 24, 2004. 

This notice advises the public that the 

NRC intends to gather the information 

necessary to prepare an EIS in support 

of the review of the application for COL 

at the Callaway Plant Unit 2 site. 

Possible alternatives to the proposed 

action (issuance of the COL for the 

Callaway Plant Unit 2 site) include no 

action, reasonable alternative energy 

sources, and alternate sites. As set forth 

mstockstill on PROD1PC66 with NOTICES