Federal Register / Vol. 74, No. 45 / Tuesday, March 10, 2009 / Notices 10305 

DEPARTMENT OF LABOR 

Employment and Training 

Administration 

[TA–W–65,339] 

Pentagon Technologies Group, Inc. 

Portland, OR; Notice of Termination of 

Investigation 

Pursuant to Section 221 of the Trade 

Act of 1974, as amended, an 

investigation was initiated on February 

23, 2009 in response to a worker 

petition filed by a company official on 

behalf of workers of Pentagon 

Technologies Group, Inc., Portland, 

Oregon. 

The petitioner has requested that the 

petition be withdrawn. Consequently, 

the investigation has been terminated. 

Signed at Washington, DC, this 24th day of 

February 2009. 

Richard Church, 

Certifying Officer, Division of Trade 

Adjustment Assistance. 

[FR Doc. E9–5050 Filed 3–9–09; 8:45 am] 

BILLING CODE 4510–FN–P 

DEPARTMENT OF LABOR 

Employment and Training 

Administration 

[TA–W–65,299] 

United States Steel Great Lakes Works, 

Ecorse, MI; Notice of Termination of 

Investigation 

Pursuant to Section 221 of the Trade 

Act of 1974, as amended, an 

investigation was initiated on February 

19, 2009 in response to a petition filed 

by the United Steelworkers of America, 

Local 1299 on behalf of workers of 

United States Steel Great Lakes Works, 

Ecorse, Michigan. 

The petitioning group of workers is 

covered by an earlier petition (TA–W– 

64,773) filed on December 19, 2008 that 

is the subject of an ongoing 

investigation for which a determination 

has not yet been issued. Further 

investigation in this case would 

duplicate efforts and serve no purpose; 

therefore the investigation under this 

petition has been terminated. 

Signed in Washington, DC, this 24th day of 

February 2009. 

Linda G. Poole, 

Certifying Officer, Division of Trade 

Adjustment Assistance. 

[FR Doc. E9–5048 Filed 3–9–09; 8:45 am] 

BILLING CODE 4510–FN–P 

DEPARTMENT OF LABOR 

Office of the Assistant Secretary for 

Veterans’ Employment and Training 

The Advisory Committee on Veterans’ 

Employment, Training and Employer 

Outreach (ACVETEO); Notice of Open 

Meeting 

The Advisory Committee on Veterans’ 

Employment, Training and Employer 

Outreach (ACVETEO) was established 

pursuant to Title II of the Veterans’ 

Housing Opportunity and Benefits 

Improvement Act of 2006 (Pub. L. 109– 

233) and Section 9 of the Federal 

Advisory Committee Act (FACA) (Pub. 

L. 92–462, Title 5 U.S.C. app.II). The 

authority of the ACVETEO is codified in 

Title 38 U.S. Code, Section 4110. 

The ACVETEO is responsible for 

assessing employment and training 

needs of veterans; determining the 

extent to which the programs and 

activities of the U.S. Department of 

Labor meet these needs; and assisting to 

conduct outreach to employers seeking 

to hire veterans. The ACVETEO will 

conduct a business meeting on Friday, 

March 20, 2009 from 8:30 a.m. to 3:30 

p.m., at the Omni Hotel, 401 Chestnut 

Street, second floor meeting room, 

Philadelphia, PA. The ACVETEO will 

discuss programs to assist veterans 

seeking employment and to raise 

employer awareness as to the 

advantages of hiring veterans, with 

special emphasis on employer outreach 

and wounded and injured veterans. 

Individuals needing special 

accommodations should notify Margaret 

Hill Watts at (202) 693–4744 by March 

9, 2009. 

Signed in Washington, DC, this 2nd day of 

March 2009. 

John M. McWilliam, 

Deputy Assistant Secretary, Veterans’ 

Employment and Training Service. 

[FR Doc. E9–4915 Filed 3–9–09; 8:45 am] 

BILLING CODE 4510–79–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2009–0100] 

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 February 12, 

2009, to February 25, 2009. The last 

biweekly notice was published on 

February 24, 2009 (74 FR 8281). 

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 

10306 Federal Register / Vol. 74, No. 45 / Tuesday, March 10, 2009 / Notices 

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. 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. 

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. 

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. 

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 on 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 

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 

Federal Register / Vol. 74, No. 45 / Tuesday, March 10, 2009 / Notices 10307 

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, 

excluding government holidays. 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, 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). 

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, the 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, unless an NRC regulation 

or other law requires submission of such 

information. 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. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units 1, 2, and 3, 

Maricopa County, Arizona 

Date of amendment request: January 

15, 2009. 

Description of amendment request: 

The amendments would modify 

Technical Specifications (TSs) 3.3.10, 

3.6.7, and 5.6.6 to delete the 

requirements related to hydrogen 

recombiners and hydrogen monitors. 

The proposed TS changes would 

support implementation of the revisions 

to 10 CFR 50.44, ‘‘Standards for 

Combustible Gas Control System in 

Light-Water-Cooled Power Reactors,’’ 

that became effective on October 16, 

2003. The proposed changes are 

consistent with Revision 1 of the NRCapproved Industry/Technical 

Specification Task Force (TSTF) 

Standard Technical Specification 

Change Traveler, TSTF–447, 

‘‘Elimination of Hydrogen Recombiners 

and Change to Hydrogen and Oxygen 

Monitors.’’ 

The NRC staff issued a notice of 

opportunity for public comments on 

TSTF–447, Revision 1, published in the 

Federal Register on August 2, 2002 (67 

FR 50374), soliciting comments on a 

model safety evaluation (SE) and a 

model no significant hazards 

consideration (NSHC) determination for 

the elimination of requirements for 

hydrogen recombiners, and hydrogen 

and oxygen monitors from TS. Based on 

its evaluation of the public comments 

received, the NRC staff made 

appropriate changes to the models and 

included final versions in a notice of 

availability published in the Federal 

Register on September 25, 2003 (68 FR 

55416), regarding the adoption of TSTF– 

447, Revision 1, as part of the NRC’s 

consolidated line item improvement 

process (CLIIP). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of NSHC adopted 

by the licensee is presented below: 

Criterion 1—The Proposed Change 

Does Not Involve a Significant Increase 

in the Probability or Consequences of an 

Accident Previously Evaluated 

The revised 10 CFR 50.44 no longer 

defines a design-basis loss-of-coolant 

accident (LOCA) hydrogen release, and 

eliminates requirements for hydrogen 

control systems to mitigate such a 

release. The installation of hydrogen 

recombiners and/or vent and purge 

systems required by 10 CFR 50.44(b)(3) 

was intended to address the limited 

quantity and rate of hydrogen 

generation that was postulated from a 

design-basis LOCA. The Commission 

has found that this hydrogen release is 

not risk-significant because the designbasis LOCA hydrogen release does not 

contribute to the conditional probability 

of a large release up to approximately 24 

hours after the onset of core damage. In 

addition, these systems were ineffective 

at mitigating hydrogen releases from 

risk-significant accident sequences that 

could threaten containment integrity. 

With the elimination of the designbasis LOCA hydrogen release, hydrogen 

monitors are no longer required to 

mitigate design-basis accidents and, 

therefore, the hydrogen monitors do not 

meet the definition of a safety-related 

component as defined in 10 CFR 50.2. 

RG [Regulatory Guide] 1.97 Category 1 

is intended for key variables that most 

directly indicate the accomplishment of 

10308 Federal Register / Vol. 74, No. 45 / Tuesday, March 10, 2009 / Notices 

a safety function for design-basis 

accident events. The hydrogen monitors 

no longer meet the definition of 

Category 1 in RG 1.97. As part of the 

rulemaking to revise 10 CFR 50.44 the 

Commission found that Category 3, as 

defined in RG 1.97, is an appropriate 

categorization for the hydrogen 

monitors because the monitors are 

required to diagnose the course of 

beyond design-basis accidents. 

The regulatory requirements for the 

hydrogen monitors can be relaxed 

without degrading the plant emergency 

response. The emergency response, in 

this sense, refers to the methodologies 

used in ascertaining the condition of the 

reactor core, mitigating the 

consequences of an accident, assessing 

and projecting offsite releases of 

radioactivity, and establishing 

protective action recommendations to 

be communicated to offsite authorities. 

Classification of the hydrogen monitors 

as Category 3 and removal of the 

hydrogen monitors from TS will not 

prevent an accident management 

strategy through the use of the SAMGs 

[severe accident management 

guidelines], the emergency plan (EP), 

the emergency operating procedures 

(EOP), and site survey monitoring that 

support modification of emergency plan 

protective action recommendations 

(PARs). 

Therefore, the elimination of the 

hydrogen recombiner requirements and 

relaxation of the hydrogen monitor 

requirements, including removal of 

these requirements from TS, does not 

involve a significant increase in the 

probability or the consequences of any 

accident previously evaluated. 

Criterion 2—The Proposed Change 

Does Not Create the Possibility of a New 

or Different Kind of Accident from Any 

Previously Evaluated 

The elimination of the hydrogen 

recombiner requirements and relaxation 

of the hydrogen monitor requirements, 

including removal of these requirements 

from TS, will not result in any failure 

mode not previously analyzed. The 

hydrogen recombiner and hydrogen 

monitor equipment was intended to 

mitigate a design-basis hydrogen 

release. The hydrogen recombiner and 

hydrogen monitor equipment are not 

considered accident precursors, nor 

does their existence or elimination have 

any adverse impact on the pre-accident 

state of the reactor core or post accident 

confinement of radionuclides within the 

containment building. 

Therefore, this change does not create 

the possibility of a new or different kind 

of accident from any previously 

evaluated. 

Criterion 3—The Proposed Change 

Does Not Involve a Significant 

Reduction in the Margin of Safety 

The elimination of the hydrogen 

recombiner requirements and relaxation 

of the hydrogen monitor requirements, 

including removal of these requirements 

from TS, in light of existing plant 

equipment, instrumentation, 

procedures, and programs that provide 

effective mitigation of and recovery 

from reactor accidents, results in a 

neutral impact to the margin of safety. 

The installation of hydrogen 

recombiners and/or vent and purge 

systems required by 10 CFR 50.44(b)(3) 

was intended to address the limited 

quantity and rate of hydrogen 

generation that was postulated from a 

design-basis LOCA. The Commission 

has found that this hydrogen release is 

not risk-significant because the designbasis LOCA hydrogen release does not 

contribute to the conditional probability 

of a large release up to approximately 24 

hours after the onset of core damage. 

Category 3 hydrogen monitors are 

adequate to provide rapid assessment of 

current reactor core conditions and the 

direction of degradation while 

effectively responding to the event in 

order to mitigate the consequences of 

the accident. The intent of the 

requirements established as a result of 

the [Three Mile Island], Unit 2 accident, 

can be adequately met without reliance 

on safety-related hydrogen monitors. 

Therefore, this change does not 

involve a significant reduction in the 

margin of safety. Removal of hydrogen 

monitoring from TS will not result in a 

significant reduction in their 

functionality, reliability, and 

availability. 

The NRC staff has reviewed the 

analysis adopted by the licensee 

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 

request for amendments involves NSHC. 

Attorney for licensee: Michael G. 

Green, Senior Regulatory Counsel, 

Pinnacle West Capital Corporation, P.O. 

Box 52034, Mail Station 8695, Phoenix, 

Arizona 85072–2034. 

NRC Branch Chief: Michael T. 

Markley. 

Exelon Generation Company, LLC, 

Docket No. 50–289, Three Mile Island 

Nuclear Station, Unit 1, Dauphin 

County, Pennsylvania 

Date of amendment request: 

September 29, 2008. 

Description of amendment request: 

The proposed changes would revise the 

TMI–1 technical specifications (TSs) to 

reflect design changes resulting from the 

planned control rod drive control 

system (CRDCS) digital upgrade project. 

In addition, the proposed amendment 

would revise the TS to remove all 

references to the axial power shaping 

rods (APSRs) to reflect changes resulting 

from their proposed elimination from 

the TMI–1 reactor. 

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, with U.S. Nuclear Regulatory 

Commission (NRC) staff edits in 

brackets: 

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 modifies 

the Technical Specifications (TSs) to 

incorporate new TS requirements associated 

with the new Digital Control Rod Drive 

Control System (DCRDCS) and an evaluation 

to permanently remove the Axial Power 

Shaping Rods (APSRs) from the reactor core. 

The proposed license amendment will 

continue to ensure reliability and operability 

of the control rod drive Reactor Trip Breakers 

(RTBs) to perform their safety function of 

tripping the reactor. The existing channel 

independence, separation and performance 

requirements of the RTBs and the Reactor 

Protection System (RPS) response time are 

retained for the new configuration. The RTB 

design was reviewed for credible common 

mode failures and no credible common mode 

failures were identified that would prevent 

the breakers from performing the reactor trip 

function. Reliable RTBs and their associated 

support circuitry provide assurance that a 

reactor trip will occur when initiated. The 

planned DCRDCS modification upgrades the 

relay-based Control Rod Drive Control 

System (CRDCS) to a solid state 

programmable DCRDCS using single rod 

power supplies assigned to each of the 61 

Control Rod Drives (CRDs). The new 

components will meet the same design 

requirements (i.e., seismic, environmental, 

quality, separation, single failure criteria) as 

the existing components in the CRDCS/RPS 

interface. The DCRDCS modification will 

improve the reliability of the system by 

resolving age-related degradation issues and 

replacing obsolete equipment. 

Malfunction of the CRD control system (or 

operator error) is an initiator of the startup 

and rod withdrawal accidents. The new 

DCRDCS meets the design requirements of 

the original system including redundancy of 

critical functions, isolation from safety 

related systems, reactivity rate limit, and 

single failure requirements. Electrical ratings, 

heat loading, structural and environmental 

aspects have been verified to be acceptable. 

Therefore, there is no increase in the 

frequency of occurrence or probability of a 

malfunction of equipment important to 

safety. The DCRDCS is not required for 

accident mitigation, post accident response 

Federal Register / Vol. 74, No. 45 / Tuesday, March 10, 2009 / Notices 10309 

or offsite release mitigation. The action of the 

RPS to trip the RTBs, to remove power from 

the control rods, and drop the rods into the 

core, remains independent of the DCRDCS. 

Therefore, there is no increase in the 

consequences or probability of occurrence of 

an accident previously evaluated. 

The modified Diverse Scram System (DSS) 

design utilizes the same power sources as the 

existing DSS, which are independent of 

reactor trip (i.e., RPS) related power sources. 

There is no change to the DSS logic circuitry. 

The DSS sensors and trip setpoint remains 

unchanged. Updated Final Safety Analysis 

(UFSAR) Section 7.1.5.4 indicates that: ‘‘The 

DSS provides an independent method of 

automatically tripping the reactor in the 

event the RPS related reactor trip system 

fails. It is designed in accordance with the 

Anticipated Transient Without Scram 

(ATWS) rule and, as such, its critical features 

are independence and diversity from the 

reactor trip system and emphasis on not 

failing in a tripped state.’’ However, DSS is 

not safety related and is not credited in any 

safety analysis in UFSAR Chapter 14, ‘‘Safety 

Analysis.’’ The assumed DSS response time 

increase from 1.0 second to 2.0 seconds has 

been evaluated and the results of the analysis 

concluded that the original acceptance 

criteria are maintained. Therefore, the 

proposed change to the DSS [is not adverse 

and] does not increase the consequence of an 

ATWS event. 

The proposed license amendment will 

continue to ensure the reliability and 

operation of the reactor core. Analyses have 

shown that the core designs employed at 

TMI–1 are stable with respect to axial 

oscillations and that xenon oscillations 

initiated during power transients are 

naturally damped or can be manually 

suppressed using regulating control rods (i.e., 

Control Rod Group 7 (CRG–7)). Actual 

operating experience at TMI–1 bears out the 

analysis conclusions that adequate axial 

imbalance control can be maintained using 

coordinated movements of CRG–7 [and] 

timed water additions. A review of the TMI– 

1 safety analyses found no mention or credit 

for APSRs in any of the events analyzed for 

TMI–1, and safety analysis assumptions are 

verified to bound key core parameters for 

each reload with explicit accounting for the 

presence of (or lack of) APSRs in the core. 

Therefore, there is no affect of APSRs on 

transient analyses, as APSR positions do not 

change in the event of a reactor trip. 

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. 

The systems affected by implementing the 

proposed changes to the TS are not assumed 

to initiate design basis accidents. Rather, the 

CRDCS/RPS interface (i.e., RTBs) is used to 

mitigate the consequences of an accident that 

has already occurred. The proposed TS 

changes do not affect the mitigating function 

of this system. The failure of any one RTB 

will not inhibit the reactor trip function. The 

modification interfaces with the DSS, which 

mitigates the ATWS event, but the interface 

function remains the same. 

A Failure Modes and Effects Analysis 

(FMEA) was performed on the DCRDCS 

design to determine if adverse effects (i.e., 

loss of reactor control, uncontrolled rod 

withdrawal, reactor trip, or prevention of 

reactor trip) could result from the credible 

failure of a single component. The FMEA 

concluded that no credible single component 

failure would cause a total loss of reactor 

control, an uncontrolled rod withdrawal, a 

reactor trip, or prevent a reactor trip. All 

operation critical to the safe and effective 

performance of the DCRDCS maintained 

sufficient redundancy such that no credible 

single failure could compromise the design 

functionality. 

The APSRs’ original function was to 

control any reactor core tendency towards 

axial oscillations resulting from xenon 

instabilities that could occur for certain early 

reactor core designs (i.e., rodded core 

designs). More recent non-rodded feed-andbleed core designs have been shown to be 

self-dampened with respect to axial xenon 

oscillations such that APSRs have not been 

moved at TMI–1 for axial power control since 

1994, and have been withdrawn from the 

reactor core since Fall 2005 with Core 

Operating Limits Report limits preventing 

insertion, consistent with AREVA reload 

methods. 

Use of [CRG–7] has been shown to 

adequately suppress axial xenon oscillations. 

The proposed changes to the CRDCS and 

APSRs and associated TS changes do not 

introduce any new accident initiators, nor do 

they reduce or adversely affect the 

capabilities of any plant structure, system, or 

component to perform their safety function. 

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 amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed TS changes do not adversely 

impact any plant safety limits, setpoints, 

response times, or design parameters. The 

changes do not negatively affect the fuel, fuel 

cladding, reactor coolant system, or 

containment integrity [under normal, 

transient or accident conditions]. 

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: J. Bradley 

Fewell, Esquire, Associate General 

Counsel, Exelon Generation Company, 

LLC, 4300 Winfield Road, Warrenville, 

IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

Exelon Generation Company, LLC, 

Docket No. 50–289, Three Mile Island 

Nuclear Station, Unit 1, (TMI–1) 

Dauphin County, Pennsylvania 

Date of amendment request: 

November 6, 2008. 

Description of amendment request: 

The proposed amendment would 

modify the TMI–1 Technical 

Specifications (TS), to replace the 

current limits on primary coolant gross 

specific activity with limits on primary 

coolant noble gas activity. The noble gas 

activity would be based on dose 

equivalent Xenon-133 (DEX) and would 

take into account only the noble gas 

activity in the primary coolant. The 

completion time for DEX being out of 

specification would be increased to 

match the action time requirements for 

the dose equivalent Iodine-131 (DEI) 

specification. In addition, the current 

DEI definition would be revised to allow 

the use of additional options for 

determining thyroid dose conversion 

factors. This change was proposed by 

the industry’s Technical Specification 

Task Force (TSTF) and is designated 

TSTF–490. The NRC staff issued a 

notice of opportunity for comment in 

the Federal Register on November 20, 

2006 (71 FR 67170), on possible 

amendments concerning TSTF–490, 

including a model safety evaluation and 

model no significant hazards (NSHC) 

determination, using the consolidated 

line item improvement process (CLIIP). 

The NRC staff subsequently issued a 

notice of availability of the models for 

referencing in license amendment 

applications in the Federal Register on 

March 15, 2007 (72 FR 12217). The 

licensee affirmed the applicability of the 

following NSHC determination in its 

application dated November 6, 2008. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below: 

Criterion 1: The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

Reactor coolant specific activity is not an 

initiator for any accident previously 

evaluated. The Completion Time when 

primary coolant gross activity is not within 

limit is not an initiator for any accident 

previously evaluated. The current variable 

limit on primary coolant iodine 

concentration is not an initiator to any 

accident previously evaluated. As a result, 

the proposed change does not significantly 

increase the probability of an accident. The 

proposed change will limit primary coolant 

noble gases to concentrations consistent with 

the accident analyses. The proposed change 

to the Completion Time has no impact on the 

10310 Federal Register / Vol. 74, No. 45 / Tuesday, March 10, 2009 / Notices 

consequences of any design basis accident 

since the consequences of an accident during 

the extended Completion Time are the same 

as the consequences of an accident during 

the Completion Time. As a result, the 

consequences of any accident previously 

evaluated are not significantly increased. 

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 in specific activity 

limits does not alter any physical part of the 

plant nor does it affect any plant operating 

parameter. The change does not create the 

potential for a new or different kind of 

accident from any previously calculated. 

Criterion 3: The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety 

The proposed change revises the limits on 

noble gas radioactivity in the primary 

coolant. The proposed change is consistent 

with the assumptions in the safety analyses 

and will ensure the monitored values protect 

the initial assumptions in the safety analyses. 

Based upon the reasoning presented above, 

the requested change does not involve a 

significant reduction in the margin of safety. 

The NRC staff has reviewed the 

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: J. Bradley 

Fewell, Esquire, Associate General 

Counsel, Exelon Generation Company, 

LLC, 4300 Winfield Road, Warrenville, 

IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

Exelon Generation Company, LLC, 

Docket No. 50–289, Three Mile Island 

Nuclear Station, Unit 1, Dauphin 

County, Pennsylvania 

Date of amendment request: October 

9, 2008. 

Description of amendment request: 

The proposed changes would revise the 

existing Three Mile Island (TMI), Unit 1, 

technical specifications (TSs) relating to 

the steam generator (SG) tube 

surveillance program. The proposed 

changes reflect the planned installation 

of replacement SGs and specifically 

address the new thermally treated Alloy 

690 tubing design of the replacement 

SGs. Removal of sections of the TSs that 

are not applicable to the replacement 

SGs are proposed. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below, with U.S. Nuclear Regulatory 

Commission (NRC) staff edits in 

brackets: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes to the Technical 

Specifications (TSs) for the TMI, Unit 1 

Steam Generator (SG) Program recognize that 

the TMI, Unit 1 SGs are being replaced and 

the standard industry performance criteria 

documented in [Technical Specification Task 

Force (TSTF) Traveler,] TSTF–449[,] for 

Alloy 690-tubed SGs will apply. These 

changes eliminate criteria that were 

established to reflect the condition and 

materials of the current TMI, Unit 1 SGs, and 

add the requirements for inspection of Alloy 

690-tubed SGs from TSTF–449. 

With these proposed TS changes, the 

operational primary-to-secondary leakage 

rate limit established for the original TMI, 

Unit 1 SGs is replaced with the standard 

industry primary-to-secondary leakage rate 

limit. The standard industry limit is that 

limit provided in TSTF–449. The current, 

reduced limit in the TMI, Unit 1 TS was 

implemented in response to upper tubesheet 

tube expansion degradation, and repairs, in 

the original TMI, Unit 1 SGs. A reduced limit 

is not required for the replacement SGs since 

they are fabricated from advanced materials 

and [will not be] subjected to the degradation 

mechanisms that influenced the original 

TMI, Unit 1 SGs. Thus, reverting to the 

standard industry limit is appropriate. The 

slightly higher, industry standard, leak rate 

limit is still low enough to provide assurance 

that the probability of tube ruptures, or of 

rapidly propagating tube leaks, remains 

acceptably low. Thus, the probability of a 

previously evaluated accident is not 

increased. 

The installation of the new SGs, with 

improved materials, will decrease the 

consequences of SG related accidents. The 

removal of accident-induced leakage 

attributable to the current degradation 

mechanisms from TS 6.19.c.1.b [provides a 

reduction in the] accident induced leakage 

limit to 1 gpm per SG. SG accident-induced 

leakage is proportional to dose; a lower 

accident-induced leakage limit will result in 

a lower dose than previously evaluated 

accident consequences. 

The proposed change to replace the 90-day 

report with a report required within 180 days 

is a change to an administrative requirement 

and does not affect the probability or 

consequences of an accident. The 180-day 

period is now industry ‘‘standard’’ practice 

per TSTF–449. 

These changes continue to provide 

reasonable assurance that the SG tubing will 

retain integrity over the full range of 

operating conditions (including startup, 

operation in the power range, hot standby, 

cooldown and all anticipated transients 

included in the design specification). With 

the proposed changes, the SG performance 

criteria (based on tube structural integrity, 

accident-induced leakage, and operational 

leakage) and SG Program are updated to 

reflect the replacement SGs while remaining 

consistent with TSTF–449. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

that was 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 recognize an 

improvement in SG design as a result of SG 

replacement. The replacement SGs contain a 

number of design improvements with respect 

to the plant’s original SGs. However, even 

with the design improvements, the 

replacement SGs are very similar to the 

original SGs and new types of accidents are 

not created. There are no new design 

functions for the Alloy 690 tubing in the 

replacement SGs. The proposed new leakage 

and inspection requirements are the standard 

industry requirements for Alloy 690 tubing. 

Primary-to-secondary leakage monitoring 

equipment is not affected by the proposed 

changes, and primary-to-secondary leakage 

will continue to be monitored to ensure it 

remains within current accident analysis 

assumptions and limits. The proposed 

changes implement the industry ‘‘standard’’ 

TSTF–449 primary-to-secondary leak limits 

for the plant’s Alloy 690-tubed replacement 

SGs. No new types of primary-to-secondary 

leak accidents are created. 

The proposed change to replace the 90-day 

report with a report required within 180 days 

is a change to an administrative requirement 

and does not create a new or different kind 

of accident. The 180-day period is now 

industry ‘‘standard’’ practice per TSTF–449. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

type of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The SG tubes in pressurized water reactors 

[PWRS] are an integral part of the reactor 

coolant pressure boundary and, as such, are 

relied upon to maintain the primary system’s 

pressure and inventory. As part of the reactor 

coolant pressure boundary, the SG tubes are 

unique in that they are also relied upon as 

a heat transfer surface between the primary 

and secondary systems such that residual 

heat can be removed from the primary 

system. The SG tubes also isolate the 

radioactive fission products in the primary 

coolant from the secondary system. In 

summary, the safety function of a SG is 

maintained by ensuring the integrity of its 

tubes. 

SG tube integrity is a function of the 

design, environment, and physical condition 

of the tubing. The proposed changes do not 

affect the operating environment but do 

recognize the improved tube material as a 

result of replacing the SGs. The proposed TS 

changes for inspection, repair, and leakage 

requirements are consistent with industry 

codes and standards for replacement SGs 

with Alloy 690 tubing material. The 

requirements established by the SG Program 

are consistent with those in the applicable 

design codes and standards. The proposed 

changes update the requirements in the 

current TSs to reflect SG replacement. 

The proposed TS changes include a change 

to the current TS limit on primary-toVerDate Nov<24>2008 15:20 Mar 09, 2009 Jkt 217001 PO 00000 Frm 00093 Fmt 4703 Sfmt 4703 E:\FR\FM\10MRN1.SGM 10MRN1
Federal Register / Vol. 74, No. 45 / Tuesday, March 10, 2009 / Notices 10311 

secondary leakage of 144 GPD [gallons per 

day] that was established in the 1980s due to 

SG tube degradation. The basis for this limit 

will no longer be applicable with the 

installation of replacement SGs. The 

proposed limit of 150 gallons per day of 

primary-to-secondary leakage through any 

one SG is ‘‘standard’’ for the U.S. PWR 

industry. This limit is based on operating 

experience with SG tube degradation 

mechanisms that result in leakage and 

provides reasonable assurance that the SG 

tubing will remain capable of fulfilling its 

specific safety function of maintaining 

reactor coolant pressure boundary integrity 

throughout each operating cycle and in the 

unlikely event of a design basis accident. 

Further, if it is not practical to assign the 

leakage to an individual SG, all the primaryto-secondary leakage is conservatively 

assumed to be from one SG. This operational 

leakage rate criterion, in conjunction with the 

implementation of the SG Program, is an 

effective measure for minimizing the 

frequency of SG tube ruptures. [Additionally, 

this TS requirement is significantly less than 

the conditions assumed in the safety 

analysis.] 

The proposed change to replace the 90-day 

report with a report required within 180 days 

is a change to an administrative requirement 

and does not affect the margin of safety. The 

180-day period is now industry ‘‘standard’’ 

practice per TSTF–449. 

For the above reasons, the margin of safety 

is not reduced. 

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: J. Bradley 

Fewell, Esquire, Associate General 

Counsel, Exelon Generation Company, 

LLC, 4300 Winfield Road, Warrenville, 

IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

R.E. Ginna Nuclear Power Plant, LLC, 

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

Power Plant, Wayne County, New York 

Date of amendment request: 

December 4, 2008. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 3.9.3, 

‘‘Containment Penetrations,’’ to permit 

refueling operations with both 

personnel airlock doors open under 

administrative control. Nuclear 

Regulatory Commission (NRC) review 

and approval of a revised non loss-ofcoolant accident (LOCA) gas gap 

fractions and fuel-handling accident 

(FHA) using the revised gap fractions 

and a shorter decay time of 72 hours 

will be necessary to support this license 

amendment. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration. The NRC staff has 

reviewed the licensee’s analysis against 

the standards of 10 CFR 50.92(c). The 

NRC staff’s review is presented below. 

1. Does the proposed change involve 

a significant increase in the probability 

or consequences of an accident 

previously evaluated? 

Response: No. 

There are three separate items 

requiring NRC approval in the licensee’s 

application. The licensee has submitted 

a plant-specific analysis to revise the 

non-LOCA gas gap fractions. Regulatory 

Guide 1.183, ‘‘Alternative Radiological 

Source Terms for Evaluating Design 

Basis Accidents at Nuclear Power 

Reactors,’’ includes Table 3, ‘‘NonLOCA Fraction of Fission Product 

Inventory in Gap.’’ The Ginna licensee 

has determined that a small number of 

fuel rods may exceed the peak power 

and burnup criteria of Table 3 thus 

necessitating the plant-specific analysis. 

The new non-LOCA gap fractions are 

considered a methodology change thus 

requiring NRC review and approval. 

The Ginna FHA currently assumes 

that fuel movement will not occur prior 

to 100 hours following reactor 

shutdown. The licensee has submitted a 

revised FHA that assumes both the new 

gas gap fractions discussed above and 

only 72 hours of decay time prior to fuel 

movement. The revised FHA must also 

be reviewed and approved by NRC. 

The proposed change to TS 3.9.3, 

which would permit refueling 

operations with both personnel airlock 

doors open under administrative 

control, impacts the release pathway for 

the FHA. The proposed TS change 

requires NRC review and approval. 

The proposed changes to the gas gap 

fractions and the FHA represent 

analytical changes and do not increase 

the probability of an accident previously 

evaluated. The change to TS 3.9.3 

introduces a new release pathway for 

the FHA and does not increase the 

probability of an FHA or any other 

accident previously evaluated. 

The change in analyzed decay time 

and the non-LOCA gap fractions result 

in an increase in the estimated dose to 

the control room and off-site receptors 

and, upon approval, will become the 

analyses of record. However, the 

increase in dose is within regulatory 

limits so that the changes do not 

represent a significant increase in the 

consequences of the FHA or any other 

accident previously evaluated. The 

proposed change to TS 3.9.3 introduces 

a new release pathway for the FHA. 

However, control room and offsite dose 

calculations are bounded by the release 

pathway from the equipment hatch. As 

a result, the proposed change to TS 3.9.3 

does not involve a significant increase 

in the consequences of an accident 

previously evaluated. 

Therefore, the probability or 

consequences of an accident previously 

evaluated will not be significantly 

increased. 

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 changes in analyzed 

decay time and the non-LOCA gap 

fractions only impact design inputs to 

the FHA. The proposed change to TS 

3.9.3 only impacts isolation 

requirements during refueling 

operations within the containment. The 

only accident which could result in a 

significant release of radioactivity in the 

plant mode where refueling is possible 

is the FHA. No other initiators or 

accident precursors are created by this 

change. 

Therefore, the proposed changes do 

not create the possibility of a new or 

different kind of accident not previously 

evaluated. 

3. Does the proposed change involve 

a significant reduction in a margin of 

safety? 

Response: No. 

The change in analyzed decay time 

and the non-LOCA gap fractions result 

in an increase in estimated dose to the 

control room and off site receptors. 

However, the dose remains within 

regulatory guidelines and limits with 

adequate margin. The proposed change 

to TS 3.9.3 introduces a new release 

pathway for the FHA which is bounded 

by the release pathway through the 

equipment hatch. 

Therefore, the proposed change does 

not involve a significant reduction in 

the margin of safety. 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: Carey Fleming, 

Sr. Counsel—Nuclear Generation, 

Constellation Group, LLC, 750 East Pratt 

Street, 17th Floor, Baltimore, MD 21202. 

NRC Branch Chief: Mark G. Kowal. 

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 

10312 Federal Register / Vol. 74, No. 45 / Tuesday, March 10, 2009 / Notices 

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 email to 

pdr@nrc.gov. 

AmerGen Energy Company, LLC, Docket 

No. 50–219, Oyster Creek Nuclear 

Generating Station, Ocean County, New 

Jersey 

Date of amendment request: 

November 13, 2007, as supplemented by 

letter dated February 18, 2009. 

Description of amendment request: 

The amendment deletes Technical 

Specification (TS) Section 6.5 and its 

associated subsections relating to the 

Review and Audit function, as well as 

correcting several administrative items. 

Additionally, the amendment 

implements changes to correct minor 

errors in TS Tables 3.1.1, 4.1.1, and 

4.1.2. 

Date of issuance: February 24, 2009. 

Effective date: As of its date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 273. 

Facility Operating License No. DPR– 

16: The amendment revised the License 

and Technical Specifications. 

Date of initial notice in Federal 

Register: April 8, 2008 (73 FR 19108). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated February 24, 2009. 

No significant hazards consideration 

comments received: No. 

AmerGen Energy Company, LLC, Docket 

No. 50–289, Three Mile Island Nuclear 

Station, Unit 1 (TMI–1), Dauphin 

County, Pennsylvania 

Date of application for amendment: 

November 13, 2007, supplemented by 

letters dated September 29, 2008, and 

February 18, 2009. 

Brief description of amendment: The 

amendment deletes Technical 

Specification (TS) Section 6.5 and its 

associated subsections relating to the 

Review and Audit function, as well as 

correcting several administrative items. 

The administrative items involve: 

correcting typographical errors, 

providing improved TS figure legibility, 

updating the description of the installed 

spent fuel pool storage locations, 

removing references to deleted TS 

sections, and correcting an error in the 

labeling of outfalls on the TMI site 

drawing. 

Date of issuance: February 24, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No.: 269. 

Facility Operating License No. DPR– 

50. Amendment revised the license and 

the technical specifications. 

Date of initial notice in Federal 

Register: April 8, 2008 (73 FR 19109). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated February 24, 2009. 

No significant hazards consideration 

comments received: No. 

Duke Energy Carolinas, LLC, et al., 

Docket No. 50–414, Catawba Nuclear 

Station, Unit 2, York County, South 

Carolina 

Date of application for amendments: 

January 20, 2009. 

Brief description of amendments: The 

amendment revised Technical 

Specification Surveillance Requirement 

(SR) 3.3.1.4 frequency. SR 3.3.1.4 is a 

Trip Actuating Device Operational Test 

of the reactor trip breakers and reactor 

trip bypass breakers. 

Date of issuance: February 13, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment No.: 242. 

Facility Operating License No. NPF– 

52: The amendment revised the license 

and the technical specifications. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): The notice 

provided an opportunity to submit 

comments on the Commission’s 

proposed NSHC determination by 

February 28, 2009. No comments have 

been received to date. However, the 

notice also provided an opportunity to 

request a hearing by March 30, 2009, but 

indicated that if the Commission make 

a final NSHC determination, any such 

hearing would take place after issuance 

of the amendment. 

Date of initial notice in Federal 

Register: January 28, 2009 (74 FR 

4986). The supplement dated February 

5, 2009, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated February 13, 

2009. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Plant, Van 

Buren County, Michigan 

Date of application for amendment: 

November 25, 2008. 

Brief description of amendment: The 

amendment would revise Appendix A 

of Technical Specifications (TSs), as 

they apply to the spent fuel pool storage 

requirements in TS Section 3.7.16 and 

the criticality requirements for the 

Region I spent fuel pool and north tilt 

pit fuel storage racks, in TS Section 

4.3.1.1. 

Date of issuance: February 6, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 236. 

Facility Operating License No. DPR– 

20: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: January 2, 2009 (74 FR 123). 

Federal Register / Vol. 74, No. 45 / Tuesday, March 10, 2009 / Notices 10313 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated February 6, 2009. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Plant, Van 

Buren County, Michigan 

Date of application for amendment: 

July 21, 2008. 

Brief description of amendment: The 

amendment supports a proposed change 

to the in-service inspection program that 

is based on topical report WCAP– 

16168–NP–A, Revision 2, ‘‘RiskInformed Extension of the Reactor 

Vessel In-Service Inspection Interval.’’ 

In the referenced safety evaluation of 

the topical report, the NRC required 

licensees to amend their licenses to 

require that the information and 

analyses requested in Section (e) of the 

final 10 CFR 50.61a (or the proposed 10 

CFR 50.61a, given in 72 FR 56275 prior 

to issuance of the final 10 CFR 50.61a) 

be submitted for NRC staff review and 

approval within one year of completing 

the required reactor vessel weld 

inspection. Entergy Nuclear Operations, 

Inc., added a new license condition to 

provide this information. 

Date of issuance: February 11, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 237. 

Facility Operating License No. DPR– 

20: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: November 4, 2008 (73 FR 

65690). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

February 11, 2009. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., System Energy 

Resources, Inc., South Mississippi 

Electric Power Association, and Entergy 

Mississippi, Inc., Docket No. 50–416, 

Grand Gulf Nuclear Station, Unit 1, 

Claiborne County, Mississippi 

Date of application for amendment: 

March 1, 2007, as supplemented by 

letters dated September 5 and 

September 21, 2007, February 14, 2008, 

and January 19 and February 20, 2009. 

Brief description of amendment: The 

changes revised the allowable values in 

the Grand Gulf Nuclear Station, Unit 1, 

Technical Specification Tables 3.3.5.1– 

1 and 3.3.5.2–1 for the Condensate 

Storage Tank (CST) low level setpoints 

for the High Pressure Core Spray and 

Reactor Core Isolation Cooling suction 

swap from the CST to the Suppression 

Pool. 

Date of issuance: February 25, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No: 181. 

Facility Operating License No. NPF– 

29: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: May 8, 2007 (72 FR 26176). 

The supplements dated September 5 

and September 21, 2007, February 14, 

2008, and January 19 and February 20, 

2009, 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 February 25, 

2009. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50– 

382, Waterford Steam Electric Station, 

Unit 3 (Waterford 3), St. Charles Parish, 

Louisiana 

Date of amendment request: August 

16, 2007, as supplemented by letter 

dated January 8, 2009. 

Brief description of amendment: The 

amendment added a new license 

condition on the control room envelope 

(CRE) habitability program; revised the 

TS requirements related to the CRE 

habitability in TS 3.7.6, ‘‘Control Room 

Emergency Air Filtration System— 

Operating,’’ TS 3.7.6.2, ‘‘Control Room 

Emergency Air Filtration System— 

Shutdown,’’ and TS 3.7.6.5, ‘‘Control 

Room Isolation and Pressurization’’; and 

established a CRE habitability program 

in TS Section 6.5, ‘‘Administrative 

Controls—Programs.’’ These changes are 

consistent with the NRC-approved 

Industry/TS Task Force (TSTF) Traveler 

TSTF–448, Revision 3, ‘‘Control Room 

Habitability.’’ 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: February 20, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 120 

days from the date of issuance. 

Amendment No.: 218. 

Facility Operating License No. NPF– 

38: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: September 25, 2007 (72 FR 

54473). 

The supplemental letter dated January 

8, 2009, 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 February 20, 

2009. 

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 1, 

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. 

Exelon Generation Company, LLC, 

Docket No. 50–352 and No. 50–353, 

Limerick Generating Station, Unit 1 and 

2, Montgomery County, Pennsylvania. 

Exelon Generation 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. 

Exelon Generation Company, LLC, 

Docket No. 50–289, Three Mile Island 

Nuclear Station, Unit 1 (TMI–1), 

Dauphin County, Pennsylvania. 

Date of application for amendments: 

February 28, 2008. 

Brief description of amendments: The 

amendment incorporates Technical 

Specification Task Force Change 

10314 Federal Register / Vol. 74, No. 45 / Tuesday, March 10, 2009 / Notices 

Traveler No. 308, Rev. 1, 

‘‘Determination of Cumulative and 

Projected Dose Contributions in the 

Radioactive Effluent Controls Program 

(RECP),’’ which clarified the existing 

wording in the RECP technical 

specification to reflect the intent of 

Generic Letter 89–01, ‘‘Implementation 

of Programmatic and Procedural 

Controls for radiological Effluent 

Technical Specifications (RETS) in the 

Administrative Controls Section of the 

Technical Specifications and the 

Relocation of the Procedural Details of 

RETS to the Offsite Dose Calculation 

Manual or to the Process Control 

Program,’’ regarding the periodicity of 

dose projections for the calendar quarter 

and year. 

Date of issuance: February 23, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 156, 156, 161, 161, 

184, 43, 230, 223, 190, 177, 197, 158, 

272, 270, 274, 242, 237 and 268. 

Facility Operating License Nos. NPF– 

72, NPF–77, NPF–37, NPF–66, NPF–62, 

DPR–2, DPR–19, DPR–25, NPF–11, NPF– 

18, NPF–39, NPF–85, DPR–16, DPR–44, 

DPR–56, DPR–29, DPR–30, and DPR–50: 

The amendments revised the Technical 

Specifications/Licenses. 

Date of initial notice in Federal 

Register: May 20, 2008 (73 FR 29162). 

The Commission’s related evaluation of 

the amendments is contained in a Safety 

Evaluation dated February 23, 2009. 

No significant hazards consideration 

comments received: No. 

FPL Energy Seabrook, LLC, Docket No. 

50–443, Seabrook Station, Unit No. 1, 

Rockingham County, New Hampshire 

Date of amendment request: February 

8, 2008. 

Description of amendment request: 

This amendment changes the Technical 

Specifications to delete Surveillance 

Requirement 4.6.3.1, which specifies 

post-maintenance testing requirements 

for containment isolation valves. 

Date of issuance: February 23, 2009. 

Effective date: As of its date of 

issuance, and shall be implemented 

within 90 days. 

Amendment No.: 120. 

Facility Operating License No. NPF– 

86: The amendment revised the License 

and Technical Specifications. 

Date of initial notice in Federal 

Register: August 26, 2008 (73 FR 

50361). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

February 23, 2009. 

No significant hazards consideration 

comments received: No comments were 

received. However, a hearing was 

requested which included contentions 

challenging the NRC staff’s proposed no 

significant hazards consideration 

determination. On October 14, 2008, the 

request for hearing was denied by the 

Atomic Safety and Licensing Board. In 

accordance with 10 CFR 50.91(a)(3), the 

NRC staff made a final determination of 

no significant hazards consideration 

which is included in the Safety 

Evaluation. 

Nine Mile Point Nuclear Station, LLC, 

Docket No. 50–220, Nine Mile Point 

Nuclear Station, Unit No. 1 (NMP1), 

Oswego County, New York 

Date of application for amendment: 

February 25, 2008. 

Brief description of amendments: The 

amendment revises NMP1 Technical 

Specification (TS) Section 3/4.4.4, 

‘‘Emergency Ventilation System,’’ to 

remove the operability and surveillance 

requirements for the 10,000 watt heater 

located in the common supply inlet air 

duct for the Reactor Building Emergency 

Ventilation System. The amendment 

also revises TS 3/4.4.5, ‘‘Control Room 

Air Treatment System,’’ to reduce the 

10-hour duration monthly system 

operational surveillance test 

requirement to a 15-minute run 

surveillance test requirement. 

Date of issuance: February 17, 2009. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment No.: 201. 

Renewed Facility Operating License 

No. DPR–063: The amendment revises 

the License and TSs. 

Date of initial notice in Federal 

Register: April 8, 2008 (73 FR 19110). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated February 17, 2009. 

No significant hazards consideration 

comments received: No. 

Nine Mile Point Nuclear Station, LLC, 

Docket No. 50–410, Nine Mile Point 

Nuclear Station, Unit No. 2 (NMP2), 

Oswego County, New York 

Date of application for amendment: 

August 14, 2008. 

Brief description of amendment: The 

amendment revises the NMP1 Technical 

Specification (TS) Surveillance 

Requirement frequency in TS 3.1.3, 

‘‘Control Rod Operability,’’ and 

Example 1.4–3 in TS Section 1.4, 

‘‘Frequency,’’ to clarify the applicability 

of the 1.25 test interval extension. The 

proposed changes are consistent with 

the Nuclear Regulatory Commission 

(NRC)-approved Revision 1 to TS Task 

Force (TSTF) Change Traveler, TSTF– 

475, ‘‘Control Rod Notch Testing 

Frequency and SRM Insert Control Rod 

Action,’’ and NUREG–1433, ‘‘Standard 

Technical Specifications General 

Electric Plants, BWR/4,’’ Revision 3.0. A 

notice of availability for this TS 

improvement using the consolidated 

line item improvement process was 

published in the Federal Register on 

November 13, 2007 (72 FR 63935). 

Date of issuance: February 23, 2009. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment No.: 130. 

Renewed Facility Operating License 

No. NPF–69: Amendment revises the 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: October 21, 2008 (73 FR 

62567). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

February 23, 2009. 

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: 

February 8, 2008, as supplemented by 

letter dated April 25, 2008, and email 

dated January 7, 2009. 

Brief description of amendment: The 

amendment revises Technical 

Specification 5.6.6, ‘‘Reactor Coolant 

System (RCS) Pressure and Temperature 

Limits Report (PTLR),’’ to include a new 

methodology for establishing reactor 

pressure vessel pressure-temperature 

limits in the Ginna PTLR. The new 

PTLR methodology is documented in 

WCAP–14040–A, Revision 4, 

‘‘Methodology Used to Develop Cold 

Overpressure Mitigating System 

Setpoints and RCS Heatup and 

Cooldown Limit Curves.’’ 

Date of issuance: February 23, 2009. 

Effective date: As of the date of 

issuance to be implemented within 90 

days. 

Amendment No.: 106. 

Renewed Facility Operating License 

No. DPR–18: Amendment revised the 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: April 8, 2008 (73 FR 19111). 

The supplemental letter dated April 25, 

2008, and email dated January 7, 2009, 

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 

Federal Register / Vol. 74, No. 45 / Tuesday, March 10, 2009 / Notices 10315 

Safety Evaluation dated February 23, 

2009. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–348 and 50–364, 

Joseph M. Farley Nuclear Plant, Units 1 

and 2, Houston County, Alabama 

Date of amendment request: October 

8, 2008. 

Brief description of amendment 

request: The amendments revise the TS 

for the diesel fuel oil testing program. 

The proposed changes are based on 

NRC-approved Technical Specifications 

Task Force (TSTF) Traveler TSTF–374, 

revision 0. Prior notice of such a 

proposed change using the Consolidated 

Line Item Improvement Process was 

provided in the Federal Register on 

April 21, 2006 (71 FR 20735). 

Date of issuance: February 20, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment Nos.: 181 and 174. 

Facility Operating License Nos. NPF– 

2 and NPF–8: Amendments revised the 

licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: December 16, 2008 (73 FR 

76413) The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

February 20, 2009. 

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: 

December 1, 2008. 

Brief description of amendment: On 

October 31, 2008, the NRC approved 

Amendment No. 186 to allow a one-time 

extension to the Completion Times for 

both essential service water (ESW) 

trains and the emergency diesel 

generators from 72 hours to 14 days. 

Amendment No. 186 was effective on 

the date of issuance and approved 

implementation by December 31, 2008, 

to permit replacement of ESW piping. 

The licensee completed the replacement 

of ESW Train A piping, but deferred the 

replacement of ESW Train B piping to 

early 2009. Amendment No. 191 

authorized implementation of the ESW 

Train B piping prior to April 30, 2009. 

Date of issuance: February 24, 2009. 

Effective date: As of its date of 

issuance, and shall be implemented 

prior to April 30, 2009. 

Amendment No.: 191. 

Facility Operating License No. NPF– 

30: The amendment revised the 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: December 23, 2008 (73 FR 

78858). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 24, 

2009. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 26th day 

of February 2009. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E9–4898 Filed 3–9–09; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Sunshine Federal Register Notice 

AGENCY HOLDING THE MEETINGS: Nuclear 

Regulatory Commission. 

DATES: Weeks of March 9, 16, 23, 30, 

April 6, 13, 2009. 

PLACE: Commissioners’ Conference 

Room, 11555 Rockville Pike, Rockville, 

Maryland. 

STATUS: Public and closed. 

Week of March 9, 2009—Tentative 

There are no meetings scheduled for 

the week of March 9, 2009. 

Week of March 16, 2009—Tentative 

Monday, March 16, 2009 

9:30 a.m. 

Briefing on State of Nuclear Materials 

and Waste Programs (Public 

Meeting) (Contact: Tammy 

Bloomer, 301–415–1725). 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov. 

Tuesday, March 17, 2009 

1:30 p.m. 

Briefing on State of Nuclear Reactor 

Safety Programs (Public Meeting) 

(Contact: Tammy Bloomer, 301– 

415–1725). 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov. 

Friday, March 20, 2009 

9:30 a.m. 

Briefing on the Nuclear Education 

Program (Public Meeting) (Contact: 

John Gutteridge, 301–492–2313). 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov. 

Week of March 23, 2009—Tentative 

There are no meetings scheduled for 

the week of March 23, 2009. 

Week of March 30, 2009—Tentative 

There are no meetings scheduled for 

the week of March 30, 2009. 

Week of April 6, 2009—Tentative 

There are no meetings scheduled for 

the week of April 6, 2009. 

Week of April 13, 2009—Tentative 

Wednesday, April 15, 2009 

9:30 a.m. 

Briefing on NRC Corporate Support 

(Public Meeting) (Contact: Karen 

Olive, 301–415–2276). 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov. 

Thursday, April 16, 2009 

1:30 p.m. 

Briefing on Human Capital and EEO 

(Public Meeting) (Contact: Kristin 

Davis, 301–492–2266). 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov. 

* The schedule for Commission 

meetings is subject to change on short 

notice. To verify the status of meetings, 

call (recording)—(301) 415–1292. 

Contact person for more information: 

Rochelle Bavol, (301) 415–1651. 

The NRC Commission Meeting 

Schedule can be found on the Internet 

at: http://www.nrc.gov/about-nrc/policymaking/schedule.html. 

The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g. 

braille, large print), please notify the 

NRC’s Disability Program Coordinator, 

Rohn Brown, at 301–492–2279, TDD: 

301–415–2100, or by e-mail at 

rohn.brown@nrc.gov. Determinations on 

requests for reasonable accommodation 

will be made on a case-by-case basis. 

This notice is distributed by mail to 

several hundred subscribers; if you no 

longer wish to receive it, or would like 

to be added to the distribution, please 

contact the Office of the Secretary, 

Washington, DC 20555 (301–415–1969). 

In addition, distribution of this meeting 

notice over the Internet system is 

available. If you are interested in 

receiving this Commission meeting 

schedule electronically, please send an 

electronic message to 

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