4764 Federal Register / Vol. 74, No. 16 / Tuesday, January 27, 2009 / Notices 

IV. Request for Comments 

Comments are invited on: (1) Whether 

the proposed collection of information 

is necessary for the proper performance 

of the functions of NASA, including 

whether the information collected has 

practical utility; (2) the accuracy of 

NASA’s estimate of the burden 

(including hours and cost) of the 

proposed collection of information; (3) 

ways to enhance the quality, utility, and 

clarity of the information to be 

collected; and (4) ways to minimize the 

burden of the collection of information 

on respondents, including automated 

collection techniques or the use of other 

forms of information technology. 

Comments submitted in response to 

this notice will be summarized and 

included in the request for OMB 

approval of this information collection. 

They will also become a matter of 

public record. 

Walter Kit, 

NASA Clearance Officer. 

[FR Doc. E9–1709 Filed 1–26–09; 8:45 am] 

BILLING CODE 7510–13–P 

NATIONAL SCIENCE FOUNDATION 

Physics Proposal Review Panel; Notice 

of Meeting 

In accordance with the Federal 

Advisory Committee Act (Pub. L. 92– 

463, as amended), the National Science 

Foundation announces the following 

meeting. 

Name: LIGO Annual Review Policy on 

access to LIGO Data for Physics (1208). 

Date and Time: Tuesday, February 17, 

2009; 8:30 a.m.–5 p.m. 

Wednesday, February 18, 2009; 8:30 a.m.– 

3 p.m. 

Place: National Science Foundation Rm. 

II–535 and Room 130. 

Type of Meeting: Open. 

Contact Person: Dr. Beverly Berger, 

Program Director for Gravitational Physics, 

National Science Foundation, 4201 Wilson 

Blvd., Arlington, VA 22230. Telephone: (703) 

292–7372. 

Purpose of Meeting: To provide advice and 

recommendations concerning NSF support of 

the LIGO project. 

Agenda: To review and evaluate LIGO’s 

practices and proposed policies regarding the 

availability of data. 

Dated: January 21, 2009, 

Susanne Bolton, 

Committee Management Officer. 

[FR Doc. E9–1658 Filed 1–26–09; 8:45 am] 

BILLING CODE 7555–01–P 

NATIONAL TRANSPORTATION 

SAFETY BOARD 

Sunshine Act Meeting; Agenda 

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

January 28, 2009. 

PLACE: NTSB Conference Center, 429 

L’Enfant Plaza SW., Washington, DC 

20594. 

STATUS: The two items are open to the 

public. 

MATTERS TO BE CONSIDERED: 

8077 Aviation Accident Report— 

Midair Collision of Electronic News 

Gathering (ENG) Helicopters, KTVK– 

TV, Eurocopter AS350B2, N613TV, and 

U.S. Helicopters, Inc., Eurocopter 

AS350B2, N215TV, Phoenix, Arizona, 

July 27, 2007. 

7943A Aircraft Accident (Summary) 

Report—In-ifight Fire, Emergency 

Descent and Crash in a Residential Area, 

Cessna 310R, N501N, Sanford, Florida, 

July 10, 2007. 

NEWS MEDIA CONTACT: Telephone: (202) 

314–6100. 

Individuals requesting specific 

accommodations should contact 

Rochelle Hall at (202) 314–6305 by 

Friday, January 23, 2008. 

The public may view the meeting via 

a live or archived webcast by accessing 

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

NTSB home page at www.ntsb.gov. 

FOR FURTHER INFORMATION CONTACT: 

Vicky D’Onofrio, (202) 314–6410. 

Dated: January 12, 2009. 

Vicky D’Onofrio, 

Federal Register Liaison Officer. 

[FR Doc. E9–1619 Filed 1–26–09; 8:45 am] 

BILLING CODE 7533–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2009–0016] 

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 December 

31, 2008 to January 13, 2009. The last 

biweekly notice was published on 

January 13, 2009 (74 FR 1712). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

10 CFR 50.92, this means that operation 

of the facility in accordance with the 

proposed amendment would not (1) 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated; or (2) 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated; or (3) 

involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example, 

in derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

mstockstill on PROD1PC66 with NOTICES6

Federal Register / Vol. 74, No. 16 / Tuesday, January 27, 2009 / Notices 4765 

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 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 

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 

Viewer TM to access the Electronic 

Information Exchange (EIE), a 

component of the E-Filing system. The 

Workplace Forms ViewerTM is free and 

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

Information about applying for a digital 

ID certificate is available on NRC’s 

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

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

Once a petitioner/requestor has 

obtained a digital ID certificate, had a 

docket created, and downloaded the EIE 

viewer, it can then submit a request for 

hearing or petition for leave to 

intervene. Submissions should be in 

Portable Document Format (PDF) in 

accordance with NRC guidance 

available on the NRC public Web site at 

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

complete at the time the filer submits its 

documents through EIE. To be timely, 

an electronic filing must be submitted to 

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

Eastern Time on the due date. Upon 

receipt of a transmission, the E-Filing 

system time-stamps the document and 

sends the submitter an e-mail notice 

mstockstill on PROD1PC66 with NOTICES6

4766 Federal Register / Vol. 74, No. 16 / Tuesday, January 27, 2009 / Notices 

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 help 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, 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. 

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: July 2, 

2008. 

Description of amendment request: 

The amendments would revise 

Technical Specification (TS) 4.2.2, 

‘‘Control Element Assemblies,’’ to 

support replacement of the full strength 

control element assemblies (CEAs) with 

a new design beginning with the 14th 

refueling outage (U3R14) for Palo Verde 

Nuclear Generating Station (PVNGS), 

Unit 3 in the spring of 2009. 

Additionally, Arizona Public Service 

Company (APS) will be updating the TS 

by removing the registered trademark 

‘‘Inconel’’ while retaining the generic 

terminology ‘‘Alloy 625’’ and deleting 

the references to part-length CEAs in TS 

4.2.2. 

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? 

Replacement of full-strength compression 

sleeve control element assemblies with fullstrength silver (Ag)-indium (In)-Cadmium 

(Cd) control element assemblies. 

Response: No. 

The proposed change involves a new 

design for the full-strength Control Element 

Assemblies (CEA) that replaces a portion of 

B4C pellets (including the compression 

sleeve) in the tips of the CEA fingers with 

hollow silver-indium-cadmium slugs. 

The following events are related to 

inadvertent movement of the CEAs; however, 

they are not initiated by the CEAs. 

• Uncontrolled Control Element Assembly 

Withdrawal from a Subcritical or Low (Hot 

Zero) Power Condition. 

• Uncontrolled Control Element Assembly 

Withdrawal at Power. 

• Single Full-Strength Control Element 

Assembly Drop. 

• Control Element Assembly Ejection. 

These previously analyzed accidents are 

initiated by the failure of plant structures, 

systems, or components (SSC) other than the 

CEA itself. The proposed change to the CEA 

design does not have a detrimental impact on 

the integrity of any plant SSC that initiates 

an analyzed event. Additionally, the CEAs 

mitigate other events. In these events, the 

chrome plating on the portion of the clad 

exterior and the added weight has been 

conservatively accounted for in the SCRAM 

[safety control rod axe man] calculation. The 

change does not adversely affect the 

protective and mitigative capabilities of the 

plant, nor does the change affect the 

initiation or probability of occurrence of any 

accident. The SSCs will continue to perform 

their intended safety functions. 

The proposed change in CEA design has 

resulted in a slight (less than 1%) reduction 

of total reactivity. 

Computer modeling events which exhibit 

sensitivity to time dependent rod worth 

(sheared shaft/seized rotor, loss of flow from 

SAFDL [specified acceptable fuel design 

limits] and total loss of reactor coolant flow) 

demonstrate that all acceptance criteria 

continued to be met. 

Therefore this change will not significantly 

increase the probability or consequences of 

any accident previously evaluated. 

The removal of the registered trademark 

name ‘‘Inconel’’. 

Response: No. 

This change is considered editorial. 

Inconel is a registered trademark of Special 

Metals Corporation, while Alloy 625 is a 

generic alloy designation from the Unified 

Numbering System. Retaining the already 

referenced term ‘‘Alloy 625’’ does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated, as the material properties and 

application of Alloy 625 have not changed. 

Deletion of the references to part-length 

control element assemblies. 

Response: No. 

This change is considered editorial. The 

removal of this information does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated as the part-length CEAs were 

mstockstill on PROD1PC66 with NOTICES6

Federal Register / Vol. 74, No. 16 / Tuesday, January 27, 2009 / Notices 4767 

replaced in accordance with License 

Amendment 152, dated March 23, 2004 

(Agency Document Access and Management 

System (ADAMS) Accession No. 

ML040860573) and the information is no 

longer applicable. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Replacement of full-strength compression 

sleeve control element assemblies with fullstrength silver(Ag)-indium(In)-Cadmium(Cd) 

control element assemblies. 

Response: No. 

There are three differences in the 

replacement CEAs as compared to the current 

CEAs. 

First, there is a very slight change in the 

outside diameter of a portion of the cladding 

on the replacement CEAs due to chrome 

plating on the lower portion of cladding. 

Analysis demonstrates that this change will 

not cause interference between the CEA 

cladding and the guide tube inside diameter 

in the buffer region. Secondly, there is a 

slight increase in weight with the Ag-In-Cd 

CEAs. However, this difference has been 

analyzed with respect to the performance 

capability of the CEDMs [Control Element 

Drive Mechanisms] and found to be within 

design capabilities and design analyses. 

Finally, the upper edges of the spider bosses 

have been chamfered to prevent damage to 

the self-latching mechanisms that can occur 

if the CEA hangs up when lifting through the 

upper guide structure cut outs. This change 

is for ease of maintenance and has no impact 

on operation of the CEAs. 

Therefore, the Ag-In-Cd CEAs are identical 

to the compression sleeve CEAs in terms of 

form, fit and function and the proposed 

change will not introduce any new failure 

mechanisms, malfunctions, or accident 

initiators not already considered in the 

design and licensing bases. The possibility of 

a new or different malfunction of safetyrelated equipment is not created. No new 

accident scenarios, transient precursors, or 

limiting single failures are introduced as a 

result of these changes. There will be no 

adverse effects or challenges imposed on any 

safety-related system as a result of these 

changes. Therefore, the possibility of a new 

or different accident from any accident 

previously evaluated is not created as a result 

of any dimensional change. 

The removal of the registered trademark 

name ‘‘Inconel’’. 

Response: No. 

This change is considered editorial. 

Inconel is a registered trademark of Special 

Metals Corporation, while Alloy 625 is a 

generic alloy designation from the Unified 

Numbering System. Retaining the already 

referenced term ‘‘Alloy 625’’ does not create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated, as the material properties and 

application of Alloy 625 have not changed. 

Deletion of the references to part-length 

control element assemblies. 

Response: No. 

This change is considered editorial. The 

removal of this information does not create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated as the part-length CEAs were 

replaced in accordance with License 

Amendment 152, dated March 23, 2004 

(Agency Document Access and Management 

System (ADAMS) Accession No. 

ML040860573) and the information is no 

longer applicable. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Replacement of full-strength compression 

sleeve control element assemblies with fullstrength silver(Ac)-indium(In)-Cadmium(Cd) 

control element assemblies. 

Response: No. 

Reactor core safety limits are established in 

the PVNGS Technical Specifications to 

prevent overheating of the fuel and cladding 

that would result in the release of fission 

products to the reactor coolant during steady 

state operation, normal operational 

transients, and anticipated operational 

occurrences. The margin to these safety 

limits is not affected by the CEA design 

changes under consideration. 

Overheating of the fuel is prevented by 

maintaining steady state, peak linear heat 

rate (LHR) below the level at which fuel 

centerline melting occurs. If the local LHR is 

high enough to cause the fuel centerline 

temperature to reach the melting point of the 

fuel, expansion of the pellet caused by 

centerline melting may cause the pellet to 

stress the cladding to the point of failure, 

allowing an uncontrolled release of activity 

to the reactor coolant. 

Compliance with the DNBR [departure 

from nucleate boiling ratio] and fuel 

centerline melt specified acceptable fuel 

design limits (SAFDLs) is assured through 

the CEA insertion limits and alignment 

technical specifications, and through the 

power distribution limit technical 

specifications. 

There is no change to the operation of the 

full-strength CEAs due to the change from 

compression sleeve CEAs to Ag-In-Cd CEAs. 

Since the Ag-In-Cd CEAs may be used to 

control power distribution similar to the 

compression sleeve CEAs, power 

distributions will still be controlled and 

maintained within the limits necessary to 

assure SAFDLs are met. 

The proposed change in CEA design has 

resulted in a slight (less than 1%) reduction 

in total reactivity. 

Computer modeling results of events 

which exhibit sensitivity to time dependent 

rod worth (sheared shaft/seized rotor, loss of 

flow from SAFDL and total loss of reactor 

coolant flow) demonstrate that all acceptance 

criteria continued to be met. 

Therefore, since SAFDLs continue to be 

met, the change from compression sleeve 

CEAs to Ag-In-Cd CEAs does not involve a 

significant reduction in a margin of safety. 

The removal of the registered trademark 

name ‘‘Inconel’’. 

Response: No. 

The removal of the registered trademark 

name ‘‘Inconel’’ [ ] is considered editorial. 

Inconel is a registered trademark of Special 

Metals Corporation, while Alloy 625 is a 

generic alloy designation from the Unified 

Numbering System. Retaining the already 

referenced term ‘‘Alloy 625’’ does not involve 

a significant reduction in the margin of safety 

as the material properties and application of 

Alloy 625 have not changed. 

Deletion of the references to part-length 

control element assemblies. 

Response: No. 

This change is considered editorial. The 

removal of this information does not involve 

a significant reduction in the margin of safety 

as the part-length CEAs were replaced in 

accordance with Amendment 152, dated 

March 23, 2004 (Agency Document Access 

and Management System (ADAMS) 

Accession No. ML040860573) and the 

information is no longer applicable. 

The NRC staff has reviewed the 

licensee’s analysis and, based on that 

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 no significant 

hazards consideration. 

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. 

Carolina Power & Light Company, 

Docket Nos. 50–325 and 50–324, 

Brunswick Steam Electric Plant, Units 1 

and 2, Brunswick County, North 

Carolina 

Date of amendments request: October 

6, 2008. 

Description of amendments request: 

The proposed change would remove 

work hour controls and/or references to 

the NRC Generic Letter 82–12 from the 

administrative control sections of the 

technical specifications. On April 17, 

2007, the NRC approved a final rule that 

amended 10 CFR Part 26 and, among 

other changes, established requirements 

for managing worker fatigue at operating 

nuclear power plants. Subpart I, 

‘‘Managing Fatigue,’’ specifically 

addresses managing worker fatigue by 

designating individual break 

requirements, work hour limits, and 

annual reporting requirements. Subpart 

I was published in the Federal Register 

on March 31, 2008 (73 FR 16966), with 

a required implementation period of 18 

months. Compliance is, therefore, 

required by October 1, 2009. In order to 

support compliance with 10 CFR Part 

26, Subpart I, the licensee is proposing 

to remove these work hour controls 

from Technical Specification 5.2.2.e at 

the Brunswick Steam Electric Plant, 

Units 1 and 2. 

Basis for proposed no significant 

hazards consideration determination 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

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4768 Federal Register / Vol. 74, No. 16 / Tuesday, January 27, 2009 / Notices 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes remove TS 

[technical specification] controls on working 

hours for personnel who perform safety 

related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. Removal of 

the TS requirements will be performed 

concurrently with the implementation of the 

10 CFR Part 26, Subpart I requirements. The 

proposed changes do not impact the physical 

configuration or function of plant structures, 

systems, or components (SSCs) or the manner 

in which SSCs are operated, maintained, 

modified, tested, or inspected. The proposed 

changes do not impact the initiators or 

assumptions of analyzed events, nor do they 

impact the mitigation of accidents or 

transient events. 

Therefore, it is concluded that these 

changes do not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes remove TS controls 

on working hours for personnel who perform 

safety related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. Work hours 

will continue to be controlled in accordance 

with NRC requirements. The new rule allows 

for deviations from controls to mitigate or 

prevent a condition adverse to safety or as 

necessary to maintain the security of the 

facility. This ensures that the new rule will 

not restrict work hours and thereby create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed changes do not alter plant 

configuration, require that new plant 

equipment be installed, alter assumptions 

made about accidents previously evaluated, 

add any initiators, or effect the function of 

plant systems or the manner in which 

systems are operated, maintained, modified, 

tested, or inspected. 

Therefore, it is concluded that this change 

does not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes remove TS controls 

on working hours for personnel who perform 

safety related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. The 

proposed changes do not involve any 

physical changes to plant or the manner in 

which plant systems are operated, 

maintained, modified, tested, or inspected. 

The proposed changes do not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed changes will not result 

in plant operation in a configuration outside 

the design basis. The proposed changes will 

not adversely affect systems that respond to 

safely shutdown the plant and to maintain 

the plant in a safe shutdown condition. 

Removal of plant-specific TS 

administrative requirements will not reduce 

a margin of safety because the requirements 

in 10 CFR Part 26 are adequate to ensure that 

worker fatigue is managed. Therefore, it is 

concluded that these 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: David T. 

Conley, Associate General Counsel II— 

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, NC 27602. 

NRC Branch Chief: Thomas H. Boyce. 

Carolina Power & Light Company, 

Docket No. 50–261, H. B. Robinson 

Steam Electric Plant, Unit No. 2, 

Darlington County, South Carolina 

Date of amendment request: October 

6, 2008. 

Description of amendments request: 

The proposed change would remove 

work hour controls and/or references to 

the NRC Generic Letter 82–12 from the 

administrative control sections of the 

technical specifications. On April 17, 

2007, the NRC approved a final rule that 

amended 10 CFR Part 26 and, among 

other changes, established requirements 

for managing worker fatigue at operating 

nuclear power plants. Subpart I, 

‘‘Managing Fatigue,’’ specifically 

addresses managing worker fatigue by 

designating individual break 

requirements, work hour limits, and 

annual reporting requirements. Subpart 

I was published in the Federal Register 

on March 31, 2008 (73 FR 16966), with 

a required implementation period of 18 

months. Compliance is, therefore, 

required by October 1, 2009. In order to 

support compliance with 10 CFR Part 

26, Subpart I, the licensee is proposing 

to remove these work hour controls 

from Technical Specification 5.2.2.e at 

the H. B. Robinson Steam Electric Plant, 

Unit 2. 

Basis for proposed no significant 

hazards consideration determination 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes remove TS 

[technical specification] controls on working 

hours for personnel who perform safety 

related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. Removal of 

the TS requirements will be performed 

concurrently with the implementation of the 

10 CFR Part 26, Subpart I requirements. The 

proposed changes do not impact the physical 

configuration or function of plant structures, 

systems, or components (SSCs) or the manner 

in which SSCs are operated, maintained, 

modified, tested, or inspected. The proposed 

changes do not impact the initiators or 

assumptions of analyzed events, nor do they 

impact the mitigation of accidents or 

transient events. 

Therefore, it is concluded that these 

changes do not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes remove TS controls 

on working hours for personnel who perform 

safety related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. Work hours 

will continue to be controlled in accordance 

with NRC requirements. The new rule allows 

for deviations from controls to mitigate or 

prevent a condition adverse to safety or as 

necessary to maintain the security of the 

facility. This ensures that the new rule will 

not restrict work hours and thereby create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed changes do not alter plant 

configuration, require that new plant 

equipment be installed, alter assumptions 

made about accidents previously evaluated, 

add any initiators, or affect the function of 

plant systems or the manner in which 

systems are operated, maintained, modified, 

tested, or inspected. 

Therefore, it is concluded that this change 

does not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes remove TS controls 

on working hours for personnel who perform 

safety related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. The 

proposed changes do not involve any 

physical changes to the plant or the manner 

in which plant systems are operated, 

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Federal Register / Vol. 74, No. 16 / Tuesday, January 27, 2009 / Notices 4769 

maintained, modified, tested, or inspected. 

The proposed changes do not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed changes will not result 

in plant operation in a configuration outside 

the design basis. The proposed changes will 

not adversely affect systems that respond to 

safely shut down the plant and to maintain 

the plant in a safe shutdown condition. 

Removal of plant-specific TS 

administrative requirements will not reduce 

a margin of safety because the requirements 

in 10 CFR Part 26 are adequate to ensure that 

worker fatigue is managed. Therefore, it is 

concluded that these 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: David T. 

Conley, Associate General Counsel II— 

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, NC 27602. 

NRC Branch Chief: Thomas H. Boyce. 

Carolina Power & Light Company, et al., 

Docket No. 50–400, Shearon Harris 

Nuclear Power Plant, Unit 1, Wake and 

Chatham Counties, North Carolina 

Date of amendment request: October 

6, 2008. 

Description of amendment request: 

The proposed change would remove 

work hour controls and/or references to 

the NRC Generic Letter 82–12 from the 

administrative control sections of the 

technical specifications. On April 17, 

2007, the NRC approved a final rule that 

amended 10 CFR Part 26 and, among 

other changes, established requirements 

for managing worker fatigue at operating 

nuclear power plants. Subpart I, 

‘‘Managing Fatigue,’’ specifically 

addresses managing worker fatigue by 

designating individual break 

requirements, work hour limits, and 

annual reporting requirements. Subpart 

I was published in the Federal Register 

on March 31, 2008 (73 FR 16966), with 

a required implementation period of 18 

months. Compliance is, therefore, 

required by October 1, 2009. In order to 

support compliance with 10 CFR Part 

26, Subpart I, the licensee is proposing 

to remove these work hour controls 

from Technical Specification 6.2.2.f at 

the Shearon Harris Nuclear Power Plant, 

Unit 1. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes remove TS 

[technical specification] controls on working 

hours for personnel who perform safety 

related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. Removal of 

the TS requirements will be performed 

concurrently with the implementation of the 

10 CFR Part 26, Subpart I requirements. The 

proposed changes do not impact the physical 

configuration or function of plant structures, 

systems, or components (SSCs) or the manner 

in which SSCs are operated, maintained, 

modified, tested, or inspected. The proposed 

changes do not impact the initiators or 

assumptions of analyzed events, nor do they 

impact the mitigation of accidents or 

transient events. 

Therefore, it is concluded that these 

changes do not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes remove TS controls 

on working hours for personnel who perform 

safety related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. Work hours 

will continue to be controlled in accordance 

with NRC requirements. The new rule allows 

for deviations from controls to mitigate or 

prevent a condition adverse to safety or as 

necessary to maintain the security of the 

facility. This ensures that the new rule will 

not restrict work hours and thereby create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed changes do not alter plant 

configuration, require that new plant 

equipment be installed, alter assumptions 

made about accidents previously evaluated, 

add any initiators, or affect the function of 

plant systems or the manner in which 

systems are operated, maintained, modified, 

tested, or inspected. 

Therefore, it is concluded that this change 

does not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes remove TS controls 

on working hours for personnel who perform 

safety related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. The 

proposed changes do not involve any 

physical changes to the plant or the manner 

in which plant systems are operated, 

maintained, modified, tested, or inspected. 

The proposed changes do not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed changes will not result 

in plant operation in a configuration outside 

the design basis. The proposed changes will 

not adversely affect systems that respond to 

safely shut down the plant and to maintain 

the plant in a safe shutdown condition. 

Removal of plant-specific TS 

administrative requirements will not reduce 

a margin of safety because the requirements 

in 10 CFR Part 26 are adequate to ensure that 

worker fatigue is managed. Therefore, it is 

concluded that these 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: David T. 

Conley, Associate General Counsel II— 

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, NC 27602. 

NRC Branch Chief: Thomas H. Boyce. 

Entergy Operations, Inc., Docket No. 50– 

368, Arkansas Nuclear One, Unit No. 2, 

Pope County, Arkansas 

Date of amendment request: 

November 13, 2008. 

Description of amendment request: 

The proposed change will modify 

Technical Specification (TS) 3.3.1.1, 

‘‘Reactor Protective Instrumentation.’’ 

Specifically, Table 4.3–1 and the 

associated Notes 7 and 8 will be revised 

to clarify and streamline the reactor 

coolant system (RCS) flow verification 

requirements associated with the 

departure from nucleate boiling ratio 

(DNBR) reactor trip signal. 

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 CPC [Core Protection Calculator] 

reactor protective function is not considered 

an accident initiator. The primary function is 

to initiate an automatic reactor trip signal 

when specific plant conditions are reached, 

thereby limiting the consequences of an 

accident. The proposed change acts to 

eliminate unnecessary conservatisms and 

accordingly increase operational margin by 

eliminating the requirement to use 

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4770 Federal Register / Vol. 74, No. 16 / Tuesday, January 27, 2009 / Notices 

calorimetric flow measurement in the CPC 

flow verification. This method of verification 

will normally only be used in the future 

during periods when the COLSS [Core 

Operating Limits Supervisory System] RCP 

[Reactor Coolant Pump] D p flow 

measurement is unavailable. Regardless of 

the method of verification used, the CPC will 

continue to be verified to have an indicated 

RCS flow equal to or conservative relative to 

the measured RCS flow on a once per 12- 

hour basis. In so doing, the CPC will 

continue to act to generate a reactor trip on 

low DNBR as originally designed in order to 

ensure the DNBR reactor core Safety Limit is 

not exceeded. 

The relocation of measurement uncertainty 

references to the TS Bases does not reduce 

the requirements to account for uncertainties 

in any Limiting Safety System Setting (LSSS) 

designed to protect reactor core Safety 

Limits. The necessary uncertainties will 

continue to be applied as required and will 

be controlled in accordance with TS 6.5.14, 

Technical Specification Bases Control 

Program, and station procedures. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not result in 

any physical plant modifications or changes 

in the way the plant is operated. In addition, 

the CPCs are unrelated to any type of 

accident initiator previously evaluated. 

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 increases operating 

margin when the COLSS RCP Dp flow 

measurement is available for use while 

unaffecting the CPC ability to initiate an 

automatic reactor trip on low DNBR prior to 

the DNBR reactor core safety limit being 

exceeded. Relocating the references to 

measurement uncertainties to the TS Bases 

likewise has no impact on the CPC design 

function and the uncertainties will continue 

to be applied as required and controlled in 

accordance with TS 6.5.14, Technical 

Specification Bases Control Program, and 

station procedures. 

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 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: 

December 8, 2008 

Description of amendment request: 

The proposed amendment adds a 

license condition to allow a one-time 

extension of surveillance requirements 

involving the 18-month channel 

calibration and logic system functional 

tests for one channel of the reactor water 

level instrumentation system. The 

extension is to account for the effects of 

rescheduling the next refueling outage 

from early to late 2009. 

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 requested action is a one-time 

extension to the performance interval of 

certain TS [Technical Specification] 

surveillance requirements. The performance 

of the surveillances, or the failure to perform 

the surveillances, is not a precursor to an 

accident. Performing the surveillances or 

failing to perform the surveillances does not 

affect the probability of an accident. 

Therefore, the proposed delay in 

performance of the surveillance requirements 

in this amendment request does not increase 

the probability of an accident previously 

evaluated. 

A delay in performing the surveillances 

does not result in a system being unable to 

perform its required function. Additionally, 

the defense in depth of the system design 

provides additional confidence that the 

safety function is maintained. In the case of 

this one-time extension request, the relatively 

short period of additional time that the 

systems and components will be in service 

before the next performance of the 

surveillance will not affect the ability of 

those systems to operate as designed. 

Therefore, the systems required to mitigate 

accidents will remain capable of performing 

their required function. No new failure 

modes have been introduced because of this 

action and the consequences remain 

consistent with previously evaluated 

accidents. Therefore, the proposed delay in 

performance of the surveillance requirement 

in this amendment request does not involve 

a significant increase in the consequences of 

an accident. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment does not involve 

a physical alteration of any system, structure, 

or component (SSC), or a change in the way 

any SSC is operated. The surveillance 

intervals of the level instrumentation are 

currently evaluated for 30 months, which 

bounds the requested interval extension. The 

proposed amendment does not involve 

operation of any SSCs in a manner or 

configuration different from those previously 

recognized or evaluated. No new failure 

mechanisms will be introduced by the onetime surveillance extension being requested. 

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 amendment is a one-time 

extension of the performance-interval of 

certain TS surveillance requirements. 

Extending the surveillance requirements does 

not involve a modification of any TS 

Limiting Conditions for Operation. Extending 

the surveillance frequency does not involve 

a change to any limit on accident 

consequences specified in the license or 

regulations. Extending the surveillance 

frequency does not involve a change to how 

accidents are mitigated or a significant 

increase in the consequences of an accident. 

Extending the surveillance frequency does 

not involve a change in a methodology used 

to evaluate consequences of an accident. 

Extending the surveillance frequency does 

not involve a change in any operating 

procedure or process. The surveillance 

intervals of the level instrumentation are 

currently evaluated for 30 months which 

bounds the requested interval extension. The 

components involved in this request have 

exhibited reliable operation based on the 

results of the most recent performances of 

their 18-month surveillance requirements 

and the associated functional surveillances. 

Based on the limited additional period of 

time that the systems and components will 

be in service before the surveillance is next 

performed, as well as the operating 

experience that these surveillances are 

typically successful when performed, it is 

reasonable to conclude that the margin of 

safety associated with the surveillance 

requirement will not be affected by the 

requested extension. 

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. 

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Federal Register / Vol. 74, No. 16 / Tuesday, January 27, 2009 / Notices 4771 

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. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and STN 50– 

457, Braidwood Station, Units 1 and 2, 

Will County, Illinois 

Docket Nos. STN 50–454 and STN 50– 

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

Ogle County, Illinois. 

Date of amendment request: 

December 4, 2008. 

Description of amendment request: 

The proposed amendments would 

revise Technical Specifications (TSs) 

1.1, ‘‘Definitions,’’ and 3.4.16, ‘‘RCS 

Specific Activity,’’ and Surveillance 

Requirements 3.4.16.1 and 3.4.16.3. The 

proposed changes would replace the 

current TS 3.4.16 limit on reactor 

coolant system (RCS) gross specific 

activity with a new limit on RCS noble 

gas specific activity. The noble gas 

specific activity limit would be based on 

a new dose equivalent Xe–133 

definition that would replace the 

current E Bar average disintegration 

energy definition. In addition, the 

current dose equivalent I–131 definition 

would be reformatted. The availability 

of this TS revision was announced in 

the Federal Register on March 15, 2007 

(72 FR 12217) as part of the 

consolidated line item improvement 

process. The licensee affirmed the 

applicability of the model no significant 

hazards consideration determination in 

its application. 

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 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. 

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 

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. 

The Nuclear Regulatory Commission 

(NRC) staff has reviewed the analysis 

adopted by the licensee 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 

amendments involve no significant 

hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Russell Gibbs. 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station, Unit No. 1, DeWitt County, 

Illinois 

Date of amendment request: 

September 2, 2008. 

Description of amendment request: 

The proposed amendments would 

relocate Surveillance Requirements (SR) 

3.8.3.6 from the technical specifications 

(TSs) to a licensee-controlled document. 

SR 3.8.3.6 requires Emergency Diesel 

Generator fuel oil storage tanks to be 

drained, sediment removed, and 

cleaned on a 10-year interval. The 

change is consistent with the current 

revision (i.e., Rev. 3) of the Improved 

Standard Technical Specifications 

(ISTS), NUREG 1434, ‘‘Standard 

Technical Specifications General 

Electric Plants, BWR/6.’’ The SR was 

removed from the ISTS under Technical 

Specification Task Force (TSTF) 

Traveler No. 2, ‘‘Relocate the 10–Year 

Sediment Cleaning of the Fuel Oil 

Storage Tank to Licensee Control,’’ 

approved by the Nuclear Regulatory 

Commission on July 16, 1998. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration which is presented below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The FOSTs [fuel oil storage tanks] provide 

the storage for the DG [diesel generator] fuel 

oil, assuring an adequate volume is available 

for each DG to operate for seven days in the 

event of a loss of offsite power concurrent 

with a loss of coolant accident. The 

relocation of the SR to drain and clean the 

FOSTs to a licensee-controlled document 

will not impact any of the previously 

analyzed accidents. Sediment in the tank, or 

failure to perform this SR, does not 

necessarily result in an inoperable storage 

tank. Fuel oil quantity and quality are 

assured by other TS SRs that remain 

unchanged. These SRs help ensure tank 

sediment is minimized and ensure that any 

degradation of the tank wall surface that 

results in fuel oil volume reduction is 

detected and corrected in a timely manner. 

Future changes to the licensee-controlled 

document will be evaluated pursuant to the 

requirements of 10 CFR 50.59, ‘‘Changes, 

tests, and experiments,’’ to ensure that such 

changes do not result in more than a minimal 

increase in the probability or consequences 

of an accident previously evaluated. 

The proposed change does not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, and 

configuration or the manner in which the 

plant is operated and maintained. The 

proposed change does not adversely affect 

the ability of structures, systems or 

components (SSCs) to perform their intended 

safety function to mitigate the consequences 

of an initiating event within the assumed 

acceptance limits. 

The proposed change does not affect the 

source term, containment isolation, or 

radiological release assumptions used in 

evaluating the radiological consequences of 

any accident previously evaluated. Further, 

the proposed change does not increase the 

types and amounts of radiological effluent 

that may be released offsite, nor significantly 

increase individual or cumulative 

occupational/public radiation exposures. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed TS change does not involve 

the addition or modification of any plant 

equipment. Also, the proposed change will 

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not alter the design configuration, or method 

of operation of plant equipment beyond its 

normal functional capabilities. The 

requirements retained in the TS continue to 

require testing of the diesel fuel oil to ensure 

the proper functioning of the DGs. The 

proposed TS change does not create any new 

credible failure mechanisms, malfunctions or 

accident initiators. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change does not alter or 

exceed a design basis or safety limit. The 

requirements retained in the TS continue to 

require testing of the diesel fuel oil to ensure 

the DGs are able to perform their intended 

function. 

Therefore, the proposed changes 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. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Nuclear, 4300 Winfield Road, 

Warrenville, IL 60555. 

NRC Branch Chief: Russell Gibbs. 

FirstEnergy Nuclear Operating 

Company, et al., Docket Nos. 50–334 

and 50–412, Beaver Valley Power 

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

2), Beaver County, Pennsylvania 

Date of amendment request: 

September 24, 2008. 

Description of amendment request: 

The proposed amendment would 

modify Technical Specifications (TSs) 

to allow the BVPS–2 containment spray 

additive sodium hydroxide (NaOH) to 

be replaced by sodium tetraborate 

(NaTB). 

Basis for proposed no significant 

hazards consideration determination: 

As required by10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

Use of NaTB in lieu of NaOH would not 

involve a significant increase in probability 

of a previously evaluated accident because 

the containment spray additive is not an 

initiator of any analyzed accident. The NaTB 

would be stored and delivered by a passive 

method that does not have potential to affect 

plant operations. Any existing NaOH 

delivery system equipment which remains in 

place but is removed from service would 

meet existing seismic, electrical and 

containment isolation requirements. 

Therefore the change in additive, including 

removal of NaOH equipment from service, 

would not result in any failure modes that 

could initiate an accident. 

The spray additive is used to mitigate the 

consequences of a LOCA [loss-of-coolant 

accident]. Use of NaTB as an additive in lieu 

of NaOH would not involve a significant 

increase in the consequences of a previously 

evaluated accident because the amount of 

NaTB specified in the proposed TS would 

achieve a pH of 7 or greater, consistent with 

the current licensing basis. This pH is 

sufficient to achieve long-term retention of 

iodine by the containment sump fluid for the 

purpose of reducing accident related 

radiation dose following a LOCA. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Regarding the proposed use of NaTB in 

lieu of NaOH, the NaTB would be stored and 

delivered by a passive method that does not 

have potential to affect plant operations. Any 

existing NaOH delivery system equipment 

remaining in place but which is removed 

from service would meet existing seismic, 

electrical and containment isolation 

requirements. Hydrogen generation would 

not be significantly impacted by the change. 

Therefore, no new failure mechanisms, 

malfunctions, or accident initiators would be 

introduced by the proposed change and it 

would 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 quantity of NaTB specified in the 

amended TS would reduce the potential for 

undesirable chemical effects while achieving 

radiation dose reductions, corrosion control 

and hydrogen generation effects that are 

comparable to NaOH, the proposed change 

does not involve a significant reduction in a 

margin of safety. The primary function of an 

additive is to reduce loss of coolant accident 

consequences by controlling the amount of 

iodine fission products released to 

containment atmosphere from reactor coolant 

accumulating in the sump during a LOCA. 

Because the amended technical 

specifications would achieve a pH of 7 or 

greater using NaTB, dose related safety 

margins would not be significantly reduced. 

Use of NaTB reduces the potential for 

undesirable chemical effects that could 

interfere with recirculation flow through the 

sump strainers. Any existing NaOH delivery 

system equipment which remains in place 

but is removed from service would meet 

existing seismic, electrical and containment 

isolation requirements and would not 

interfere with operation of the existing 

containment or containment spray system. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David W. 

Jenkins, FirstEnergy Nuclear Operating 

Company, FirstEnergy Corporation, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: Mark G. Kowal. 

FirstEnergy Nuclear Operating Company 

(FENOC), et al., Docket No. 50–440, 

Perry Nuclear Power Plant, Unit No. 1 

(PNPP), Lake County, Ohio 

Date of amendment request: 

November 18, 2008 

Description of amendment request: 

The proposed amendment would 

modify Technical Specification (TS) 

5.5.6 to incorporate Technical 

Specification Task Force (TSTF) 

Travelers TSTF–479, ‘‘Changes to 

Reflect Revision of 10 CFR 50.55a,’’ and 

TSTF 497, ‘‘Limit Inservice Testing 

Program SR [Surveillance Requirement] 

3.0.2 Application to Frequencies of 2 

Years or Less.’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment revises TS 5.5.6, 

‘‘Inservice Testing Program,’’ for consistency 

with 10 CFR 50.55a(f)(4) requirements 

regarding inservice testing of pumps and 

valves. The proposed amendment 

incorporates revisions to the ASME Code that 

result in a net improvement in the measures 

for testing pumps and valves. The proposed 

changes do not impact any accident initiators 

or analyzed events or assumed mitigation of 

accident or transient events. They do not 

involve the addition or removal of any 

equipment, or any design changes to the 

facility. Therefore, the proposed changes do 

not represent a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not involve a 

modification to the physical configuration of 

the plant. There is no new equipment to be 

installed or a change in the methods 

governing normal plant operation. The 

proposed change will not impose any new or 

different requirements or introduce a new 

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accident initiator, accident precursor, or 

malfunction mechanism. Additionally, there 

is no change in the types or increases in the 

amounts of any effluent that may be released 

off-site and there is no increase in individual 

cumulative occupational exposure. 

Therefore, the proposed change does not 

create the possibility of an accident of a 

different kind than previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed amendment revises TS 5.5.6, 

‘‘Inservice Testing Program,’’ for consistency 

with the requirements of 10 CFR 50.55a(f)(4) 

regarding the inservice testing of pumps and 

valves. The proposed amendment 

incorporates revisions to the ASME Code that 

result in a net improvement in the measures 

for testing pumps and valves. The safety 

function of the affected pumps and valves 

will be maintained. 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: David W. 

Jenkins, Attorney, FirstEnergy 

Corporation, Mail Stop A–GO–15, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: Russell Gibbs. 

Florida Power Corporation, et al., 

Docket No. 50–302, Crystal River Unit 3 

Nuclear Generating Plant, Citrus 

County, Florida 

Date of amendment request: October 

6, 2008. 

Description of amendment request: 

The proposed change would remove 

work hour controls and/or references to 

the NRC Generic Letter 82–12 from the 

administrative control sections of the 

technical specifications. On April 17, 

2007, the NRC approved a final rule that 

amended 10 CFR Part 26 and, among 

other changes, established requirements 

for managing worker fatigue at operating 

nuclear power plants. Subpart I, 

‘‘Managing Fatigue,’’ specifically 

addresses managing worker fatigue by 

designating individual break 

requirements, work hour limits, and 

annual reporting requirements. Subpart 

I was published in the Federal Register 

on March 31, 2008 (73 FR 16966), with 

a required implementation period of 18 

months. Compliance is, therefore, 

required by October 1, 2009. In order to 

support compliance with 10 CFR Part 

26, Subpart I, the licensee is proposing 

to remove these work hour controls 

from Technical Specification 5.2.2.e at 

the Crystal River Unit 3 Nuclear 

Generating Plant. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes remove TS 

[technical specification] controls on working 

hours for personnel who perform safety 

related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. Removal of 

the TS requirements will be performed 

concurrently with the implementation of the 

10 CFR Part 26, Subpart I requirements. The 

proposed changes do not impact the physical 

configuration or function of plant structures, 

systems, or components (SSCs) or the manner 

in which SSCs are operated, maintained, 

modified, tested, or inspected. The proposed 

changes do not impact the initiators or 

assumptions of analyzed events, nor do they 

impact the mitigation of accidents or 

transient events. 

Therefore, it is concluded that these 

changes do not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes remove TS controls 

on working hours for personnel who perform 

safety related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. Work hours 

will continue to be controlled in accordance 

with NRC requirements. The new rule allows 

for deviations from controls to mitigate or 

prevent a condition adverse to safety or as 

necessary to maintain the security of the 

facility. This ensures that the new rule will 

not restrict work hours and thereby create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed changes do not alter plant 

configuration, require that new plant 

equipment be installed, alter assumptions 

made about accidents previously evaluated, 

add any initiators, or effect the function of 

plant systems or the manner in which 

systems are operated, maintained, modified, 

tested, or inspected. 

Therefore, it is concluded that this change 

does not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes remove TS controls 

on working hours for personnel who perform 

safety related functions. The TS controls are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. The 

proposed changes do not involve any 

physical changes to plant or the manner in 

which plant systems are operated, 

maintained, modified, tested, or inspected. 

The proposed changes do not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed changes will not result 

in plant operation in a configuration outside 

the design basis. The proposed changes will 

not adversely affect systems that respond to 

safely shut down the plant and to maintain 

the plant in a safe shutdown condition. 

Removal of plant-specific TS 

administrative requirements will not reduce 

a margin of safety because the requirements 

in 10 CFR Part 26 are adequate to ensure that 

worker fatigue is managed. Therefore, it is 

concluded that these 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: David T. 

Conley, Associate General Counsel II— 

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, NC 27602. 

NRC Branch Chief: Thomas H. Boyce. 

Florida Power Corporation, et al., 

Docket No. 50–302, Crystal River Unit 3 

Nuclear Generating Plant, Citrus 

County, Florida 

Date of amendment request: 

December 17, 2008. 

Description of amendments request: 

The proposed change would revise the 

Crystal River Unit 3 Improved Technical 

Specifications Administrative Controls, 

Section 5.6, to revise the Inservice 

Testing Program to incorporate the 

Technical Specification Task Force 

(TSTF) Standard TS Change Traveler, 

TSTF–479, Revision 0, ‘‘Changes to 

Reflect Revision of 10 CFR 50.55a,’’ and 

TSTF–497, Revision 0, ‘‘Limit Inservice 

Testing Program SR 3.0.2 Application to 

Frequencies of 2 Years or Less.’’ 

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: 

4. Does not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

The proposed change revises the CR–3 

[Crystal River Unit 3] ITS [Improved 

Technical Specifications], Section 5.6.2.9, 

‘‘Inservice Testing Program,’’ for consistency 

with the requirements of 10 CFR 50.55a(f)(4) 

regarding the inservice testing of pumps and 

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valves which are classified as ASME 

[American Society of Mechanical Engineers] 

Code Class 1, Class 2, and Class 3. The 

proposed change incorporates revisions to 

the ASME Code that result in a net 

improvement in the measures for testing 

pumps and valves. 

The proposed change does not impact any 

accident initiators or analyzed events or 

assumed mitigation of accident or transient 

events. The proposed change does not 

involve the addition or removal of any 

equipment, or any design changes to the 

facility. Therefore, this proposed change does 

not involve an increase in probability or 

consequences of an accident previously 

evaluated. 

5. Does not create the possibility of a new 

or different kind of accident from any 

accident previously evaluated. 

The proposed change revises the CR–3 ITS, 

Section 5.6.2.9, ‘‘Inservice Testing Program,’’ 

for consistency with the requirements of 10 

CFR 50.55a(f)(4) regarding the inservice 

testing of pumps and valves which are 

classified as ASME Code Class 1, Class 2, and 

Class 3. The proposed change incorporates 

revisions to the ASME Code that result in a 

net improvement in the measures for testing 

pumps and valves. 

The proposed change does not involve a 

modification to the physical configuration of 

the plant (i.e., no new equipment will be 

installed) or involve a change in the methods 

governing normal plant operation. The 

proposed change will not introduce a new 

accident initiator, accident precursor, or 

malfunction mechanism. Additionally, there 

is no change in types or increases in the 

amounts of any effluents that may be released 

offsite and there is no increase in individual 

or cumulative occupational exposure. 

Therefore, the proposed change does not 

create the possibility of an accident of a 

different kind than previously evaluated. 

6. Does not involve a significant reduction 

in a margin of safety[.] 

The proposed change revises the CR–3 ITS, 

Section 5.6.2.9, ‘‘Inservice Testing Program,’’ 

for consistency with the requirements of 10 

CFR 50.55a(f)(4) regarding the inservice 

testing of pumps and valves which are 

classified as ASME Code Class 1, Class 2, and 

Class 3. The proposed change does not 

involve a modification to the physical 

configuration of the plant (i.e., no new 

equipment will be installed) or change the 

methods governing normal plant operation. 

The proposed change incorporates revisions 

to the ASME Code that result in a net 

improvement in the measures for testing 

pumps and valves. The safety function of the 

affected pumps and valves will be 

maintained. 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: David T. 

Conley, Associate General Counsel II— 

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, NC 27602. 

NRC Branch Chief: Thomas H. Boyce. 

Northern States Power Company— 

Minnesota, Docket Nos. 50–282 and 50– 

306, Prairie Island Nuclear Generating 

Plant, Units 1 and 2, Goodhue County, 

Minnesota 

Date of amendment request: 

November 4, 2008. 

Description of amendment request: 

The proposed amendments would make 

changes to the Technical Specifications 

to increase the 24 month test load for 

the Unit 1 Emergency Diesel Generators 

(EDGs), D1 and D2, reduce the monthly 

test load for the Unit 2 EDGs, D5 and 

D6, and reduce the 24 month test loads 

for the Unit 2 EDGs. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This license amendment request proposes 

to increase a portion of the Prairie lsland 

Nuclear Generating Plant Unit 1 emergency 

diesel generator’s 24-month test loading, 

reduce the Unit 2 emergency diesel 

generators’ monthly test loading which 

demonstrates Technical Specification 

operability and revise the 24-month test to 

require the Unit 2 emergency diesel 

generators to operate for at least 2 hours at 

100–110% of the continuous rated loading 

and the remainder of the 24-hour test at or 

above 4000 kW. The proposed test loads will 

continue to assure that the emergency diesel 

generators have the necessary reliability and 

availability for the design basis accidents and 

station blackout events. 

The emergency diesel generators are 

required to be operable in the event of a 

design basis accident coincident with a loss 

of offsite power to mitigate the consequences 

of the accident. They are also the alternate 

AC source for a station blackout on the other 

Prairie lsland Nuclear Generating Plant unit. 

The emergency diesel generators are not 

accident initiators and therefore these 

changes do not involve a significant increase 

in the probability of an accident previously 

evaluated. 

The accident analyses assume that at least 

one safeguards bus is provided with power 

either from the offsite sources or the 

emergency diesel generators. The Technical 

Specification changes proposed in this 

license amendment request will continue to 

assure that the emergency diesel generators 

have the capacity and capability to assume 

their maximum auto-connected loads. Thus, 

the changes proposed in this license 

amendment request do not involve a 

significant increase in the consequences of an 

accident previously evaluated. 

The changes proposed in this license 

amendment do not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

This license amendment request proposes 

to increase a portion of the Prairie Island 

Nuclear Generating Plant Unit 1 emergency 

diesel generator’s 24-month test loading, 

reduce the Unit 2 emergency diesel 

generators’ monthly test loading which 

demonstrates Technical Specification 

operability and revise the 24-month test to 

require the Unit 2 emergency diesel 

generators to operate for at least 2 hours at 

100–110% of the continuous rated loading 

and the remainder of the 24-hour test at or 

above 4000 kW. The proposed test loads will 

continue to assure that the emergency diesel 

generators have the necessary reliability and 

availability for the design basis accidents and 

station blackout events. 

The proposed Technical Specification 

changes do not involve a change in the plant 

design, system operation, or the use of the 

emergency diesel generators. The proposed 

changes require the Unit 1 emergency diesel 

generators to be tested at increased loads and 

allow the Unit 2 emergency diesel generator 

to be tested at reduced loads which envelope 

the required safety function loads. These 

revised loads continue to demonstrate the 

capability and capacity of the emergency 

diesel generators to perform their required 

functions. There are no new failure modes or 

mechanisms created due to testing the 

emergency diesel generators at the proposed 

test loading. Testing of the emergency diesel 

generators at the proposed test loadings does 

not involve any modification in the 

operational limits or physical design of plant 

systems. There are no new accident 

precursors generated due to the proposed test 

loadings. 

The Technical Specification changes 

proposed in this license amendment do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

This license amendment request proposes 

to increase a portion of the Prairie Island 

Nuclear Generating Plant Unit 1 emergency 

diesel generator’s 24-month test loading, 

reduce the Unit 2 emergency diesel 

generators’ monthly test loading which 

demonstrates Technical Specification 

operability and revise the 24-month test to 

require the Unit 2 emergency diesel 

generators to operate for at least 2 hours at 

100–110% of the continuous rated loading 

and the remainder of the 24-hour test at or 

above 4000 kW. The proposed test loads will 

continue to assure that the emergency diesel 

generators have the necessary reliability and 

availability for the design basis accidents and 

station blackout events. 

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The proposed Technical Specification 

changes will continue to demonstrate that the 

emergency diesel generators meet the 

Technical Specification definition of 

operability, that is, the proposed tests will 

demonstrate that the emergency diesel 

generators will perform their safety function 

and the necessary emergency diesel generator 

attendant instrumentation, controls, cooling, 

lubrication and other auxiliary equipment 

required for the emergency diesel generators 

to perform their safety function loads are also 

tested at these proposed loadings. The 

proposed testing will also continue to 

demonstrate the capability and capacity of 

the emergency diesel generators to supply 

their required loss of offsite power loads 

coincident with station blackout loads from 

the opposite unit. Since the proposed 

surveillance testing will continue to 

demonstrate operability, and the capability 

and capacity to supply their required loss of 

offsite power coincident with opposite unit 

station blackout loads, the proposed 

Technical Specification changes do not 

involve a significant reduction in a margin of 

safety. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment requests involve no 

significant hazards consideration. 

Attorney for licensee: Peter M. Glass, 

Assistant General Counsel, Xcel Energy 

Services, Inc., 414 Nicollet Mall, 

Minneapolis, MN 55401. 

NRC Branch Chief: Lois M. James. 

Tennessee Valley Authority, Docket 

Nos. 50–259, 50–260 and 50–296, 

Browns Ferry Nuclear Plant, Units 1, 2 

and 3, Limestone County, Alabama 

Date of amendment request: March 

27, 2008, as supplemented by a letter 

December 19, 2008. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specifications (TS) 

requirements related to control building 

envelope habitability in TS Section 

3.7.3 Control Room Emergency 

Ventilation (CREV) System, and add TS 

Section 5.5.13, Control Building 

Envelope Habitability Program, to the 

Administrative Section of the TSs. The 

licensee has included conforming 

technical changes to the TS Bases. The 

proposed revision to the Bases also 

includes editorial and administrative 

changes to reflect applicable changes to 

the corresponding TS Bases, which were 

made to improve clarity, conform to the 

latest information and references, 

correct factual errors, and achieve more 

consistency with the standard TS 

NUREGs. The proposed revision to the 

TS and associated Bases is similar to the 

TSTF–448, Revision 3. The supplement 

contains additional information related 

to smoke and chemical effects and 

addresses the associated proposed 

revision to TS Section 3.7.3, TS Section 

5.5.13 and TS Bases 3.7.3. 

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 Technical 

Specification change involve a significant 

increase in the probability or consequences 

of an accident previously evaluated? 

No. The proposed change does not 

adversely affect accident initiators or 

precursors nor alter the design assumptions, 

conditions, or configuration of the facility. 

The proposed change does not alter or 

prevent the ability of structures, systems, and 

components (SSCs) to perform their intended 

function to mitigate the consequences of an 

initiating event within the assumed 

acceptance limits. The proposed change 

revises the TS for the CRE emergency 

ventilation system, which is a mitigation 

system designed to minimize unfiltered air 

leakage into the CRE and to filter the CRE 

atmosphere to protect the CRE occupants in 

the event of accidents previously analyzed. 

An important part of the CRE emergency 

ventilation system is the CRE boundary. The 

CRE emergency ventilation system is not an 

initiator or precursor to any accident 

previously evaluated. Therefore, the 

probability of any accident previously 

evaluated is not increased. Performing tests 

to verify the operability of the CRE boundary 

and implementing a program to assess and 

maintain CRE habitability ensure that the 

CRE emergency ventilation system is capable 

of adequately mitigating radiological 

consequences to CRE occupants during 

accident conditions, and that the CRE 

emergency ventilation system will perform as 

assumed in the consequence analyses of 

design basis accidents. Thus, the 

consequences of any accident previously 

evaluated are not increased. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed Technical 

Specification change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

No. The proposed change does not impact 

the accident analysis. The proposed change 

does not alter the required mitigation 

capability of the CRE emergency ventilation 

system, or its functioning during accident 

conditions as assumed in the licensing basis 

analyses of design basis accident radiological 

consequences to CRE occupants. No new or 

different accidents result from performing the 

new surveillance or following the new 

program. The proposed change does not 

involve a physical alteration of the plant (i.e., 

no new or different type of equipment will 

be installed) or a significant change in the 

methods governing normal plant operation. 

The proposed change does not alter any 

safety analysis assumptions and is consistent 

with current plant operating practice. 

Therefore, this change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed Technical 

Specification change involve a significant 

reduction in a margin of safety? 

The proposed change does not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The proposed 

change does not affect safety analysis 

acceptance criteria. The proposed change 

will not result in plant operation in a 

configuration outside the design basis for an 

unacceptable period of time without 

compensatory measures. The proposed 

change does not adversely affect systems that 

respond to safely shut down the plant and to 

maintain the plant in a safe shutdown 

condition. 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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Branch Chief: Thomas H. Boyce. 

Tennessee Valley Authority, Docket No. 

50–260, Browns Ferry Nuclear Plant, 

Unit 2, Limestone County, Alabama 

Date of amendment request: 

December 22, 2008 (TS–463–T). 

Description of amendment request: 

The proposed amendment would, on a 

one-time basis, extend several Technical 

Specification (TS) surveillance 

frequencies approximately 45 days. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The requested action is a one-time 

extension to the performance interval of a 

limited number of TS surveillance 

requirements. The performance of these 

surveillances, or the failure to perform these 

surveillances, is not a precursor to an 

accident. Performing these surveillances or 

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4776 Federal Register / Vol. 74, No. 16 / Tuesday, January 27, 2009 / Notices 

failing to perform these surveillances does 

not affect the probability of an accident. 

Therefore, the proposed delay in 

performance of the surveillance requirements 

in this amendment request does not increase 

the probability of an accident previously 

evaluated. 

A delay in performing these surveillances 

does not result in a system being unable to 

perform its required function. In the case of 

this one-time extension request, the relatively 

short period of additional time that the 

systems and components will be in service 

before the next performance of the 

surveillance will not affect the ability of 

those systems to operate as designed. 

Therefore, the systems required to mitigate 

accidents will remain capable of performing 

their required function. No new failure 

modes have been introduced because of this 

action and the consequences remain 

consistent with previously evaluated 

accidents. Therefore, the proposed delay in 

performance of the surveillance requirements 

in this amendment request does not involve 

a significant increase in the consequences of 

an accident. 

Therefore, operation of the facility in 

accordance with the proposed license 

amendment would not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment does not involve 

a physical alteration of any system, structure, 

or component (SSC) or a change in the way 

any SSC is operated. The proposed 

amendment does not involve operation of 

any SSCs in a manner or configuration 

different from those previously recognized or 

evaluated. No new failure mechanisms will 

be introduced by the one-time surveillance 

requirement extensions being requested. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed amendment is a one-time 

extension of the performance interval of a 

limited number of TS surveillance 

requirements. Extending these surveillance 

requirements does not involve a modification 

of any TS Limiting Conditions for Operation. 

Extending these surveillance requirements 

does not involve a change to any limit on 

accident consequences specified in the 

license or regulations. Extending these 

surveillance requirements does not involve a 

change to how accidents are mitigated or a 

significant increase in the consequences of an 

accident. Extending these surveillance 

requirements does not involve a change in a 

methodology used to evaluate consequences 

of an accident. Extending these surveillance 

requirements does not involve a change in 

any operating procedure or process. 

The instrumentation and components 

involved in this request have exhibited 

reliable operation based on the results of the 

most recent performance of their 24-month 

surveillance requirements. 

Based on the limited additional period of 

time that the systems and components will 

be in service before the surveillances are next 

performed, as well as the operating 

experience that these surveillances are 

typically successful when performed, it is 

reasonable to conclude that the margins of 

safety associated with these surveillance 

requirements will not be affected by the 

requested extension. 

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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Branch Chief: Thomas Boyce. 

Tennessee Valley Authority, Docket No. 

50 390, Watts Bar Nuclear Plant, Unit 1, 

Rhea County, Tennessee 

Date of amendment request: August 1, 

2008, as supplemented November 25 

and December 31, 2008 (2 letters). 

Description of amendment request: 

The proposed amendment would revise 

the following: (1) Technical 

Specification (TS) 4.2.1, ‘‘Fuel 

Assemblies,’’ and TS Surveillance 

Requirements 3.5.1.4, ‘‘Accumulators,’’ 

and 3.5.4.3, ‘‘RWST [Refueling Water 

Storage Tank],’’ to increase the 

maximum number of Tritium Producing 

Burnable Absorber Rods (TPBARs) that 

can be irradiated per cycle from 400 to 

704. 

An application that addressed similar 

issues was previously submitted on 

August 1, 2008, and notice of that 

application was provided in the Federal 

Register on November 12, 2008 (73 FR 

66946). Due to certain changes in the 

specifics of the December 31, 2008, 

revision from those proposed in the 

August 1, 2008, application, as 

supplemented on November 25 and 

December 31, 2008, the application is 

being renoticed in its entirety. This 

notice supersedes the notice published 

in the Federal Register on November 12, 

2008. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change modifies the 

maximum number of TPBARs in the core. 

The required boron concentration for the 

cold leg accumulators (CLAs) and RWST 

remains unchanged. The current boron 

concentration has been demonstrated to 

maintain the required accident mitigation 

safety function for the CLAs and RWST with 

the higher number of TPBARs and this will 

be verified for each core that contains 

TPBARs as part of the normal reload 

analysis. The CLAs and RWST safety 

function is to mitigate accidents that require 

the injection of borated water to cool the core 

and to control reactivity. These functions are 

not potential sources for accident generation 

and the modification of the number of 

TPBARs will not increase the potential for an 

accident. Therefore, the possibility of an 

accident is not increased by the proposed 

changes. The current boron concentration 

levels are supported by the proposed number 

of TPBARs in the core. Since the current 

boron concentration levels will continue to 

maintain the safety function of the CLAs and 

RWST in the same manner as currently 

approved, the consequences of an accident 

are not increased by the proposed changes. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change only modifies the 

maximum number of TPBARs in the core. 

The boron concentrations for accident 

mitigation functions of the CLAs and RWST 

remain unchanged. These functions do not 

have a potential to generate accidents as they 

only serve to perform mitigation functions 

associated with an accident. The proposed 

modification will maintain the mitigation 

function in an identical manner as currently 

approved. There are no plant equipment or 

operational changes associated with the 

proposed revision. Therefore, since the CLA 

and RWST functions are not altered and the 

plant will continue to operate without 

change, the possibility of a new or different 

kind of an accident is not created. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

This change proposes a change to the 

maximum number of TPBARs in the core. 

The boron concentration requirements that 

support the accident mitigation functions of 

the CLAs and RWST remain unchanged. The 

proposed change does not alter any plant 

equipment or components and does not alter 

any setpoints utilized for the actuation of 

accident mitigation system or control 

functions. The proposed number of TPBARs, 

in conjunction with the current boron 

concentration values, has been demonstrated 

to provide an adequate level of reactivity 

control for accident mitigation and this will 

be verified for each core that contains 

TPBARs as part of the normal reload 

analysis. Therefore, the proposed change will 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

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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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Acting Branch Chief: P. Milano. 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Plant, Van 

Buren County, Michigan 

Date of amendment request: 

November 25, 2008. 

Brief description of amendment 

request: The proposed amendment 

would revise Appendix A, Technical 

Specifications (TS), as they apply to the 

spent fuel pool (SFP) storage 

requirements in TS section 3.7.16 and 

the criticality requirements for the 

Region I SFP and north tilt pit fuel 

storage racks, in TS section 4.3.1.1. 

Date of publication of individual 

notice in Federal Register: January 2, 

2009 (74 FR 123). 

Expiration date of individual notice: 

February 3, 2009. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: 

December 1, 2008. 

Description of amendment request: By 

letter dated October 31, 2008, the 

Nuclear Regulatory Commission issued 

Amendment No. 186, to Callaway Plant, 

Unit 1, Facility Operating License No. 

NPF–30. The amendment allowed a 

one-time extension of the allowed 

outage time (completion time) for each 

of the two essential service water (ESW) 

trains (ESW Train A and Train B) from 

72 hours to 14 days. The extended 

completion time was requested to 

support planned replacement of the 

underground carbon steel piping with 

new high density polyethylene (HDPE) 

piping for ESW Train A and ESW Train 

B during plant operation. The 

amendment was issued with a 

requirement to complete the 

replacement of carbon steel piping with 

HDPE piping for both ESW trains by 

December 31, 2008. By its application 

dated December 1, 2008, the licensee 

informed NRC that it had experienced 

significant delays in completing the 

replacement of underground piping/ 

conduit due, in part, to underground 

obstructions during excavation, a longer 

refueling outage (Refuel 16) than 

anticipated, a forced outage at the 

beginning of Cycle 17, switchyard 

maintenance, and other equipment and 

personnel issues. However, the 

replacement of ESW Train A carbon 

steel piping was completed by the 

required date of December 31, 2008, but 

the replacement of ESW Train B carbon 

steel piping was deferred. Consequently, 

the licensee proposed to extend the 

implementation date for completion of 

replacement of carbon steel piping for 

ESW Train B from December 31, 2008, 

to April 30, 2009. 

Date of publication of individual 

notice in Federal Register: December 23, 

2008 (73 FR 78858). 

Expiration date of individual notice 

comment period: January 22, 2009. 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

of 1954, as amended (the Act), and the 

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

amendment, (2) the amendment, and (3) 

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

Systems (ADAMS) Public Electronic 

Reading Room on the internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr.resource@nrc.gov. 

AmerGen Energy Company, LLC, Docket 

No. 50–461, Clinton Power Station, Unit 

No. 1, DeWitt County, Illinois 

AmerGen Energy Company, LLC, et al., 

Docket No. 50–219, Oyster Creek 

Nuclear Generating Station, Ocean 

County, New Jersey 

AmerGen Energy Company, LLC, Docket 

No. 50–289, Three Mile Island Nuclear 

Station, Unit 1 (TMI–1), Dauphin 

County, Pennsylvania 

Date of application for amendments: 

June 20, 2008. 

Brief description of amendments: The 

amendments conform the licenses to 

reflect the direct transfer of AmerGen 

Energy Company, LLC’s ownership and 

operating authority for Clinton Power 

Station, Unit No. 1, Oyster Creek 

Nuclear Generating Station (Oyster 

Creek), and Three Mile Island Nuclear 

Station, Unit 1, to Exelon Generation 

Company, LLC, (ECG) as approved by 

Commission Order dated December 23, 

2008. Transfer of the license for Oyster 

Creek will also authorize EGC to store 

spent fuel in the Oyster Creek 

independent spent fuel storage 

installation. 

Date of issuance: January 8, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

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Amendment Nos.: CPS–183, Oyster 

Creek–271, and TMI–1–267. 

Facility Operating License Nos. NPF– 

62, DPR–16, and DPR–50: The 

amendments revised the Technical 

Specifications and Licenses. 

Date of initial notice in Federal 

Register: August 26, 2008 (73 FR 

50368). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

December 23, 2008. 

No significant hazards consideration 

comments received: The NRC received 

three comments on August 27, 2008, 

one for each plant’s initial notice. The 

comments did not provide any 

information additional to that in the 

application, nor did they provide any 

information contradictory to that 

provided in the application. 

Dominion Energy Kewaunee, Inc. 

Docket No. 50–305, Kewaunee Power 

Station, Kewaunee County, Wisconsin 

Date of application for amendment: 

April 4, 2008. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications by removing the 

operability and surveillance 

requirements for the shield building 

ventilation (SBV) and auxiliary building 

special ventilation filter train heaters, 

and reducing the operating time 

required to verify the SBV system 

operability from 10 hours to 15 minutes. 

Date of issuance: December 30, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 201. 

Facility Operating License No. DPR– 

43: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: June 3, 2008 (73 FR 31720) 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated December 30, 2008. 

No significant hazards consideration 

comments received: No. 

Dominion Energy Kewaunee, Inc. 

Docket No. 50–305, Kewaunee Power 

Station, Kewaunee County, Wisconsin 

Date of application for amendment: 

April 14, 2008, as supplemented by 

letter dated October 17, 2008. 

Brief description of amendment: The 

amendment adds a new footnote to 

Kewaunee Technical Specifications 

Table 3.5–4, ‘‘Instrument Operating 

Conditions for Isolation Functions.’’ The 

new footnote allows the main steam line 

isolation circuitry to be inoperable 

when both main steam isolation valves 

are closed and deactivated. 

Date of issuance: January 12, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 202. 

Facility Operating License No. DPR– 

43: Amendment revised the operating 

license and Technical Specifications. 

Date of initial notice in Federal 

Register: June 17, 2008 (73 FR 34340) 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated January 12, 2009. 

No significant hazards consideration 

comments received: No. 

Duke Energy Carolinas, LLC, et. al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina 

Date of application for amendments: 

December 11, 2007, as supplemented 

December 18, 2008. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications sections to allow the 

bypass test times and Completion Times 

(CTs) for Limiting Condition for 

Operation (LCOs) 3.3.1, ‘‘Reactor Trip 

System (RTS) Instrumentation;’’ 3.3.2, 

‘‘Engineered Safety Feature Actuation 

System (ESFAS) Instrumentation;’’ 

3.3.6, ‘‘Containment Air Release and 

Addition Isolation Instrumentation,’’ 

and 3.3.9, ‘‘Boron Dilution Mitigation 

System (BDMS).’’ 

The proposed license amendment 

request (LAR) adopts changes as 

described in Westinghouse Commercial 

Atomic Power (WCAP) topical report 

WCAP–14333–P–A, Revision 1, 

‘‘Probabilistic Risk Analysis of the 

Reactor Protection System and 

Engineered Safety Features Actuation 

System Test Times and Completion 

Times,’’ issued October 1998 and 

approved by U.S. Nuclear Regulatory 

Commission (NRC) letter dated July 15, 

1998. Implementation of the proposed 

changes is consistent with Technical 

Specification Task Force (TSTF) 

Traveler TSTF–418, Revision 2, ‘‘RPS 

[Reactor Protection System] and ESFAS 

Test Times and Completion Times 

(WCAP–14333).’’ The NRC approved 

TSTF–418, Revision 2, by letter dated 

April 2, 2003. 

In addition, the proposed LAR adopts 

changes as described in WCAP–15376– 

P–A, Revision 1, ‘‘Risk-Informed 

Assessment of the RTS and ESFAS 

Surveillance Test Intervals and Reactor 

Trip Breaker Test and Completion 

Times,’’ issued March 2003, as 

approved by NRC letter dated December 

20, 2002. Implementation of the 

proposed changes is consistent with 

TSTF Traveler # TSTF–411, Revision 1, 

‘‘Surveillance Test Interval Extension 

for Components of the Reactor 

Protection System (WCAP–15376).’’ The 

NRC approved TSTF–411, Revision 1, 

by letter dated August 30, 2002. The 

licensee also requested additional 

changes not specifically included in the 

above topical reports. These changes 

will be evaluated in a future 

amendment. 

Date of issuance: December 22, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment Nos.: 247 and 240. 

Facility Operating License Nos. NPF– 

35 and NPF–52: Amendments revised 

the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: March 25, 2008 (73 FR 

15783). The supplement dated 

December 18, 2008, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated December 22, 

2008. 

No significant hazards consideration 

comments received: No. 

Duke Energy Carolinas, LLC, et. al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina 

Date of application for amendments: 

December 11, 2007, as supplemented by 

letter dated December 18, 2008. 

Brief description of amendments: The 

amendments revised the Technical 

Specification sections to allow the 

bypass test times and Completion Times 

for Limiting Condition for Operation 

3.3.1, ‘‘Reactor Trip System (RTS) 

Instrumentation’’ and 3.3.2, 

‘‘Engineered Safety Feature Actuation 

System (ESFAS) Instrumentation.’’ 

By letter dated December 30, 2008 

(Agencywide Documents Access and 

Management System Accession No. 

ML0083460216), the NRC issued 

Amendment No. 247 and Amendment 

No. 240 for Catawba Units 1 and 2, 

respectively, for all the proposed 

changes approved by the NRC in TSTFs 

411 and 418. The December 30, 2008, 

amendment stated that the following 

changes would be evaluated in a future 

amendment: 

Surveillance requirement (SR) 3.3.1.5, 

Safety injection input from ESFAS, 

Condition J, Feedwater isolation with 

low average core temperature coincident 

with reactor trip P–4, SR 3.3.2.2, turbine 

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trip and feedwater isolation for steam 

generator water level high high. 

(P–14), SR 3.3.2.4 turbine trip and 

feedwater isolation for steam generator 

water level high high (P–14), and SR 

3.3.2.5 turbine trip and feedwater 

isolation for low average core 

temperature trip coincident with reactor 

trip P–4. 

This amendment approves the above 

changes. 

Date of issuance: January 9, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment Nos.: 248 and 241. 

Facility Operating License Nos. NPF– 

35 and NPF–52: Amendments revised 

the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: March 25, 2008 (73 FR 

15783). The supplement dated 

December 18, 2008, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the NRC staff’s 

original proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated January 9, 2009. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of application for amendment: 

February 6, 2008, as supplemented by 

letter dated July 29, 2008. 

Brief description of amendment: The 

amendment revised the Surveillance 

Requirements (SRs) for control rod 

exercising from weekly to monthly in 

Technical Specification (TS) 4.3.A.2, 

revise verification of control rod 

coupling integrity as described in TS 

4.3.B.1, revise the scram insertion time 

Limiting Conditions for Operation 

(LCOs) and SRs as described in TS 3.3.C 

and 4.3.C, and enhance TS 3.3.D and 

4.3.D, the LCO and SR for Control Rod 

Accumulators. 

Date of issuance: January 7, 2009. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 233. 

Facility Operating License No. DPR– 

28: Amendment revised the License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: March 11, 2008 (73 FR 

13024). The supplemental letter dated 

July 29, 2008, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. The 

Commission’s related evaluation of this 

amendment is contained in a Safety 

Evaluation dated January 7, 2009. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50– 

313, Arkansas Nuclear One, Unit No. 1, 

Pope County, Arkansas 

Date of amendment request: July 30, 

2008, as supplemented by letter dated 

October 2, 2008. 

Brief description of amendment: The 

amendment revises the current TS 

3.6.6.3 surveillance requirements for 

sodium hydroxide (NaOH) 

concentration. Specifically, the 

amendment changes the surveillance 

requirements of the NaOH tank solution 

concentration from between 5.0 weight 

(wt.) percent and 16.5 wt. percent to 

between 6.0 wt. percent and 8.5 wt. 

percent. 

Date of issuance: January 13, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: Unit 1—234. 

Renewed Facility Operating License 

No. DPR–51: Amendment revised the 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: November 4, 2008, (73 FR 

65694). The supplement dated October 

2, 2008, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 13, 

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: 

June 30, 2008. 

Brief description of amendment: The 

amendment (1) deleted Technical 

Specification (TS) surveillance 

requirement (SR) 3.1.3.2 and revised SR 

3.1.3.3; (2) removed the reference to SR 

3.1.3.2 from Required Action A.2 of TS 

3.1.3, ‘‘Control Rod OPERABILITY’’; (3) 

clarified the requirement to fully insert 

all insertable rods for the limiting 

condition for operation in TS 3.3.1.2 

Required Action E.2, ‘‘Source Range 

Monitoring Instrumentation’’; and (4) 

revised Example 1.4–3 in Section 1.4, 

‘‘Frequency,’’ to clarify the applicability 

of the 1.25 surveillance test interval 

extension. 

Date of issuance: December 31, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No: 180. 

Facility Operating License No. NPF– 

29: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 26, 2008 (73 FR 

50359). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated December 31, 

2008. 

No significant hazards consideration 

comments received: No. 

Florida Power Corporation, et al., 

Docket No. 50–302, Crystal River Unit 

No. 3, Nuclear Generating Plant, Citrus 

County, Florida 

Date of application for amendment: 

January 17, 2008. 

Brief description of amendment: The 

amendment revises the Crystal River, 

Unit 3 Improved Technical 

Specification Surveillance Requirement 

3.7.5.2, ‘‘Emergency Feedwater 

System,’’ to align the text for the 

emergency feedwater system 

surveillance frequency with the text in 

the Technical Specifications Task Force 

Standard Technical Specification 

Change Traveler-101, Revision 0 and the 

NRC technical report, NUREG–1430, 

Volume 1, Revision 3, ‘‘Standard 

Technical Specifications Babcock and 

Wilcox Plants—Specification.’’ 

Date of issuance: January 9, 2009. 

Effective date: Date of issuance, to be 

implemented within 60 days. 

Amendment No.: 231. 

Facility Operating License No. DPR– 

72: Amendment revises the technical 

specifications. 

Date of initial notice in Federal 

Register: May 20, 2008 (73 FR 29163). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 9, 2009. 

No significant hazards consideration 

comments received: No. 

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4780 Federal Register / Vol. 74, No. 16 / Tuesday, January 27, 2009 / Notices 

Pacific Gas and Electric Company, 

Docket Nos. 50–275 and 50–323, Diablo 

Canyon Nuclear Power Plant, Unit Nos. 

1 and 2, San Luis Obispo County, 

California 

Date of application for amendments: 

December 17, 2007, as supplemented by 

letters dated October 2, and November 

18, 2008. 

Brief description of amendments: The 

amendments increase the completion 

times (CTs) for required actions related 

to Technical Specifications (TS) 3.5.2, 

regarding the Emergency Core Cooling 

System, and 3.6.6, regarding the 

Containment Spray and Cooling 

Systems from 72 hours to 14 days. In 

addition, invalid notes were deleted 

from TSs 3.5.2 and 3.6.6 and new notes 

were added to specify the limitations on 

the use of the 14-day extended CT. 

Date of issuance: December 31, 2008. 

Effective date: As of its date of 

issuance and shall be implemented 

within 180 days from the date of 

issuance. 

Amendment Nos.: Unit 1—202; Unit 

2—203. 

Facility Operating License Nos. DPR– 

80 and DPR–82: The amendments 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in Federal 

Register: January 29, 2008 (73 FR 

5227). The supplement(s) dated October 

2 and November 18, 2008, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated December 31, 

2008. 

No significant hazards consideration 

comments received: No. 

PPL Susquehanna, LLC, Docket Nos. 

50–387 and 50–388, Susquehanna 

Steam Electric Station, Units 1 and 2, 

Luzerne County, Pennsylvania 

Date of application for amendments: 

July 7, 2008. 

Brief description of amendments: The 

amendments revised the Technical 

Specification (TS) testing frequency for 

the Surveillance Requirement (SR) in TS 

3.1.4, ‘‘Control Rod Scram Times.’’ The 

change revised the frequency of SR 

3.1.4.2, control rod scram time testing, 

from ‘‘120 days cumulative operation in 

Mode 1’’ to ‘‘200 days cumulative 

operation in Mode 1.’’ These changes 

are based on TS Task Force (TSTF) 

change traveler TSTF–460 (Revision 0) 

that has been approved generically for 

the Boiling-Water Reactor (BWR) 

Standard TS, NUREG–1433 (BWR/4) 

and NUREG–1434 (BWR/6) by revising 

the frequency of SR 3.1.4.2, control rod 

scram time testing, from ‘‘120 days 

cumulative operation in MODE 1’’ to 

‘‘200 days cumulative operation in 

MODE 1.’’ 

Date of issuance: January 2, 2009. 

Effective date: January 2, 2009. 

Amendment Nos.: 249 for Unit 1 and 

228 for Unit 2 

Facility Operating License Nos. NPF– 

14 and NPF–22: The amendments 

revised the License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: October 7, 2008 (73 FR 

58675). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated January 2, 2009. 

No significant hazards consideration 

comments received: No. 

PPL Susquehanna, LLC, Docket Nos. 

50–387 and 50–388, Susquehanna 

Steam Electric Station, Units 1 and 2, 

Luzerne County, Pennsylvania 

Date of application for amendments: 

July 7, 2008. 

Brief description of amendments: The 

amendment adopted the Nuclear 

Regulatory Commission (NRC) approved 

Technical Specification Task Force 

(TSTF) change traveler TSTF–475, 

(Revision 1), ‘‘Control Rod Notch 

Testing Frequency and SRM [Source 

Range Monitor] Insert Control Rod 

Action,’’ to change the Standard 

Technical Specifications (STS) for 

General Electric (GE) Plants (NUREG– 

1433, BWR/4 to the plant-specific TS, 

that allows: (1) Revising the frequency 

of Surveillance Requirement (SR) 

3.1.3.2, notch testing of fully withdrawn 

control rod, from ‘‘7 days after the 

control rod is withdrawn and 

THERMAL POWER is greater than the 

LPSP of RWM’’ to ‘‘31 days after the 

control rod is withdrawn and 

THERMAL POWER is greater than the 

LPSP [Low Power Set Point] of the 

RWM [Rod With Minimizer]’’, and (2) 

revising Example 1.4–3 in Section 1.4 

‘‘Frequency’’ to clarify that the 1.25 

surveillance test interval extension in 

SR 3.0.2 is applicable to time periods 

discussed in NOTES in the 

‘‘SURVEILLANCE’’ column in addition 

to the time periods in the 

‘‘FREQUENCY’’ column. 

Date of issuance: January 2, 2009. 

Effective date: January 2, 2009. 

Amendment Nos.: 250 for Unit 1 and 

229 for Unit 2. 

Facility Operating License Nos. NPF– 

14 and NPF–22: The amendments 

revised the License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: October 7, 2008 (73 FR 

58675). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated January 2, 2009. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket No. 

50 390, Watts Bar Nuclear Plant, Unit 1, 

Rhea County, Tennessee 

Date of application for amendment: 

September 19, 2008. 

Brief description of amendment: The 

amendment modifies the Final Safety 

Analysis Report by requiring an 

inspection of the ice condenser within 

24 hours of experiencing a seismic event 

greater than or equal to an operating 

basis earthquake within the 5-week 

period after ice basket replenishment 

has been completed to confirm that 

adverse ice fallout has not occurred that 

could impede the ability of the ice 

condenser lower inlet doors to open. 

Date of issuance: January 6, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 45 days of issuance. 

Amendment No.: 73. 

Facility Operating License No. NPF– 

90: Amendment authorizes revision to 

the FSAR. 

Date of initial notice in Federal 

Register: November 4, 2008 (73 FR 

65698). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 6, 2009. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket No. 

50 390, Watts Bar Nuclear Plant, Unit 1, 

Rhea County, Tennessee 

Date of application for amendment: 

March 27, 2008, as supplemented 

September 26, 2008. 

Brief description of amendment: The 

amendment revises the allowable value 

listed for Function 3, ‘‘Containment 

Purge Exhaust Radiation Monitors,’’ in 

Table 3.3.6–1, ‘‘Containment Vent 

Isolation Instrumentation,’’ of the 

limited condition for operation 3.3.6. 

Date of issuance: January 8, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 74. 

Facility Operating License No. NPF– 

90: Amendment revises the Technical 

Specifications and License. 

Date of initial notice in Federal 

Register: May 6, 2008 (73 FR 25047). 

The supplement dated September 26, 

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Federal Register / Vol. 74, No. 16 / Tuesday, January 27, 2009 / Notices 4781 

2008, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 8, 2009. 

No significant hazards consideration 

comments received: No. 

Virginia Electric and Power Company, 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Units 1 and 2, 

Louisa County, Virginia 

Date of application for amendment: 

December 17, 2007, as supplemented on 

July 22, 2008, September 26, 2008, and 

November 25, 2008. 

Brief description of amendment: 

These amendments revised Technical 

Specification (TS) 3.8.3 to allow a onetime extended 14-day completion time 

(CT) for each of the two underground 

diesel fuel oil storage tanks (FOST) to 

permit removal of the current coating 

and to recoat the tanks in preparation 

for use of ultra-low sulfur diesel fuel oil. 

The change revised the TS to extend the 

CT associated with an inoperable 

emergency diesel generator FOST from 

7 days to 14 days, applicable once for 

each of the two tanks. 

Date of issuance: December 31, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 254 and 235. 

Renewed Facility Operating License 

Nos. NPF–4 and NPF–7: Amendments 

change the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: January 15, 2008 (73 FR 

2552). The supplements dated July 22, 

2008, September 26, 2008, and 

November 25, 2008, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. The 

Commission’s related evaluation of the 

amendments is contained in a Safety 

Evaluation dated December 31, 2008. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 15th day 

of January 2009. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E9–1568 Filed 1–26–09; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[EA–09–001; NRC–2009–0017] 

In the Matter of Certain General 

Material Licensees; Demand for 

Information 

I 

The Nuclear Regulatory Commission 

(NRC or Commission) is issuing this 

Demand for Information because it is 

our understanding that you possess 

radioactive material in the form of 

tritium in exit signs. Because you 

possess radioactive material in this 

form, you hold what is referred to as a 

‘‘general license’’ to possess such 

material. In this case, your general 

license has been issued by the NRC 

pursuant to section 31.5 in Part 10 of the 

Code of Federal Regulations (10 CFR 

31.5). This general license authorizes 

you, the licensee, to receive, possess, 

use, or transfer, in accordance with the 

provisions of paragraphs (b), (c) and (d) 

of 10 CFR 31.5, radioactive material 

contained in devices designed and 

manufactured for the purpose of 

producing light. 

II 

On December 7, 2006, NRC issued 

Regulatory Issue Summary (RIS) 2006– 

25, ‘‘Requirements for the Distribution 

and Possession of Tritium Exit Signs 

and the Requirements in 10 CFR 31.5 

and 32.51a.’’ This RIS was issued in part 

to remind general licensees of the 

requirements in 10 CFR 31.5 regarding 

transfer and disposal of tritium exit 

signs. It was NRC’s intent that issuance 

of this RIS would minimize the chances 

of improper disposal of tritium exit 

signs. 

Despite the publication of the RIS in 

2006, NRC has reason to believe that 

certain general licensees may lack 

awareness of their responsibility to 

account for and properly dispose of 

tritium exit signs. Therefore, the NRC 

needs further information to determine 

whether we can have reasonable 

assurance that general licensees are 

complying with NRC regulations 

applying to the possession, transfer, and 

disposal of tritium exit signs. 

III 

Accordingly, pursuant to sections 

161c, 161o, 182 and 186 of the Atomic 

Energy Act of 1954, as amended, and 

the Commission’s regulations in 10 CFR 

2.204 and 10 CFR 31.5, the NRC seeks 

information in order to determine 

whether additional regulatory action 

should be taken to ensure compliance 

with NRC requirements. Within 60 days 

of the date of this Demand for 

Information, you must submit a written 

answer to the Director, Office of Federal 

and State Materials and Environmental 

Management Programs, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001. Your answer must be 

submitted under oath or affirmation, 

and it must provide the following 

information: 

A. Explain how you ensure 

compliance with the NRC requirements 

applying to the possession, transfer, and 

disposal of tritium exit signs you have 

acquired. Identify and provide contact 

information for the individual you have 

appointed who is responsible for 

ensuring day-to-day compliance with 

these requirements. 

B. State the number of tritium exit 

signs you currently possess and the 

number of signs that, according to your 

records, should be in your possession. 

C. Explain the reasons for any 

discrepancy between the number of 

tritium exit signs you currently possess 

and the number of signs that should be 

in your possession. 

D. Describe any actions you have 

taken, or plan to take, to locate tritium 

exit signs that should be, but are not, in 

your possession. 

E. Describe any actions you have 

taken, or plan to take, to prevent future 

losses of tritium exit signs. 

After reviewing your response, the 

NRC will determine whether further 

action is necessary to ensure 

compliance with regulatory 

requirements. 

The Director, Office of Federal and 

State Materials and Environmental 

Management Programs, may, in writing, 

relax or rescind any of the above 

conditions upon demonstration by the 

Licensee of good cause, such as a 

particularly large number of signs 

spread over multiple locations. If you 

believe you cannot report the results 

within the 60-day deadline, you may 

forward a request to extend the 

deadline. Extensions will be granted if 

you can reasonably demonstrate an 

inability to meet the deadline. 

Additionally, any other requirement can 

be relaxed or rescinded, as long as you 

can reasonably demonstrate why that 

requirement should be relaxed or 

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