Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 58669 

This meeting, from 9 a.m. to 11:30 

a.m. (ending time is approximate), will 

be open to the public on a space 

available basis. The meeting will begin 

with opening remarks and will include 

a poetry reading by David Lehman, a 

performance from Shakespeare by 

Aquila Theater Company, and a jazz 

performance by pianist Helen Sung. 

After the presentations the Council will 

review and vote on applications and 

guidelines, and the meeting will end 

with remarks and Council members’ 

farewell to the Chairman. 

If, in the course of the open session 

discussion, it becomes necessary for the 

Council to discuss non-public 

commercial or financial information of 

intrinsic value, the Council will go into 

closed session pursuant to subsection 

(c)(4) of the Government in the 

Sunshine Act, 5 U.S.C. 552b. 

Additionally, discussion concerning 

purely personal information about 

individuals, submitted with grant 

applications, such as personal 

biographical and salary data or medical 

information, may be conducted by the 

Council in closed session in accordance 

with subsection (c) (6) of 5 U.S.C. 552b. 

Any interested persons may attend, as 

observers, Council discussions and 

reviews that are open to the public. If 

you need special accommodations due 

to a disability, please contact the Office 

of AccessAbility, National Endowment 

for the Arts, 1100 Pennsylvania Avenue, 

NW., Washington, DC 20506, 202/682– 

5532, TTY–TDD 202/682–5429, at least 

seven (7) days prior to the meeting. 

Further information with reference to 

this meeting can be obtained from the 

Office of Communications, National 

Endowment for the Arts, Washington, 

DC 20506, at 202/682–5570. 

Dated: October 2, 2008. 

Kathy Plowitz-Worden, 

Panel Coordinator, Office of Guidelines and 

Panel Operations. 

[FR Doc. E8–23705 Filed 10–6–08; 8:45 am] 

BILLING CODE 7537–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses; Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a.(2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC 

staff) is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued from September 

11, 2008 to September 24, 2008. The last 

biweekly notice was published on 

September 23, 2008 (73 FR 54862). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

10 CFR 50.92, this means that operation 

of the facility in accordance with the 

proposed amendment would not (1) 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated; or (2) 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated; or (3) 

involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example, 

in derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rulemaking, 

Directives and Editing Branch, Division 

of Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be delivered to Room 6D22, Two 

White Flint North, 11545 Rockville 

Pike, Rockville, Maryland, from 7:30 

a.m. to 4:15 p.m. Federal workdays. 

Copies of written comments received 

may be examined at the Commission’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area 01F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. The filing of 

requests for a hearing and petitions for 

leave to intervene is discussed below. 

Within 60 days after the date of 

publication of this notice, person(s) may 

file a request for a hearing with respect 

to issuance of the amendment to the 

subject facility operating license and 

any person whose interest may be 

affected by this proceeding and who 

wishes to participate as a party in the 

proceeding must file a written request 

via electronic submission through the 

NRC E-Filing system for a hearing and 

a petition for leave to intervene. 

Requests for a hearing and a petition for 

leave to intervene shall be filed in 

accordance with the Commission’s 

‘‘Rules of Practice for Domestic 

Licensing Proceedings’’ in 10 CFR Part 

2. Interested person(s) should consult a 

current copy of 10 CFR 2.309, which is 

available at the Commission’s PDR, 

located at One White Flint North, Public 

File Area 01F21, 11555 Rockville Pike 

(first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

mstockstill on PROD1PC66 with NOTICES

58670 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/ 

requestor to relief. A petitioner/ 

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

A request for hearing or a petition for 

leave to intervene must be filed in 

accordance with the NRC E-Filing rule, 

which the NRC promulgated on August 

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

process requires participants to submit 

and serve documents over the Internet 

or in some cases to mail copies on 

electronic storage media. Participants 

may not submit paper copies of their 

filings unless they seek a waiver in 

accordance with the procedures 

described below. 

To comply with the procedural 

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

days prior to the filing deadline, the 

petitioner/requestor must contact the 

Office of the Secretary by e-mail at 

hearingdocket@nrc.gov, or by calling 

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

ID certificate, which allows the 

participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and/or (2) creation of an 

electronic docket for the proceeding 

(even in instances in which the 

petitioner/requestor (or its counsel or 

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

petitioner/requestor will need to 

download the Workplace Forms 

ViewerTM to access the Electronic 

Information Exchange (EIE), a 

component of the E-Filing system. The 

Workplace Forms ViewerTM is free and 

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

Information about applying for a digital 

ID certificate is available on NRC’s 

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

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

Once a petitioner/requestor has 

obtained a digital ID certificate, had a 

docket created, and downloaded the EIE 

viewer, it can then submit a request for 

hearing or petition for leave to 

intervene. Submissions should be in 

Portable Document Format (PDF) in 

accordance with NRC guidance 

available on the NRC public Web site at 

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

complete at the time the filer submits its 

documents through EIE. To be timely, 

an electronic filing must be submitted to 

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

Eastern Time on the due date. Upon 

receipt of a transmission, the E-Filing 

system time-stamps the document and 

sends the submitter an e-mail notice 

confirming receipt of the document. The 

EIE system also distributes an e-mail 

notice that provides access to the 

document to the NRC Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/ 

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically may 

seek assistance through the ‘‘Contact 

Us’’ link located on the NRC Web site 

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

technical help line, which is available 

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

Eastern Time, Monday through Friday. 

The help line number is (800) 397–4209 

or locally, (301) 415–4737. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file a 

motion, in accordance with 10 CFR 

2.302(g), with their initial paper filing 

requesting authorization to continue to 

submit documents in paper format. 

Such filings must be submitted by: (1) 

First class mail addressed to the Office 

of the Secretary of the Commission, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, Attention: 

Rulemaking and Adjudications Staff; or 

(2) courier, express mail, or expedited 

delivery service to the Office of the 

Secretary, Sixteenth Floor, One White 

Flint North, 11555 Rockville Pike, 

Rockville, Maryland 20852, Attention: 

Rulemaking and Adjudications Staff. 

Participants filing a document in this 

manner are responsible for serving the 

document on all other participants. 

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

the mail, or by courier, express mail, or 

expedited delivery service upon 

depositing the document with the 

provider of the service. 

Non-timely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission, the presiding officer, or 

mstockstill on PROD1PC66 with NOTICES

Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 58671 

the Atomic Safety and Licensing Board 

that the petition and/or request should 

be granted and/or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

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

filings must be submitted no later than 

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

date. 

Documents submitted in adjudicatory 

proceedings will appear in NRC’s 

electronic hearing docket which is 

available to the public at http:// 

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

unless excluded pursuant to an order of 

the Commission, an Atomic Safety and 

Licensing Board, or a Presiding Officer. 

Participants are requested not to include 

personal privacy information, such as 

social security numbers, home 

addresses, or home phone numbers in 

their filings. With respect to copyrighted 

works, except for limited excerpts that 

serve the purpose of the adjudicatory 

filings and would constitute a Fair Use 

application, participants are requested 

not to include copyrighted materials in 

their submission. 

For further details with respect to this 

amendment action, see the application 

for amendment which is available for 

public inspection at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the ADAMS Public 

Electronic Reading Room on the Internet 

at the NRC Web site, http:// 

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

you do not have access to ADAMS or if 

there are problems in accessing the 

documents located in ADAMS, contact 

the PDR Reference staff at 1 (800) 397– 

4209, (301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

Calvert Cliffs Nuclear Power Plant, Inc., 

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

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of amendments request: August 

28, 2008. 

Description of amendments request: 

The amendment would relocate the 

main steam isolation valve times in 

Technical Specification (TS) section 

3.7.2, ‘‘Main Steam Isolation Valves 

(MSIVs)’’ to the licensee controlled 

document that is referenced in the 

Bases. In addition, the valve isolation 

times in the TS are replaced with the 

phrase ‘‘within limits.’’ The changes are 

consistent with the Nuclear Regulatory 

Commission approved Technical 

Specification Task Force (TSTF)–491, 

Revision 2, ‘‘Removal of Main Steam 

and Main Feedwater Valve Isolation 

Times From Technical Specifications.’’ 

The availability of the TS improvement 

was published in the Federal Register 

on December 29, 2006 (71 FR 250) as 

part of the consolidated item 

improvement process (CLIIP). 

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: 

Criterion 1—The Proposed Change Does 

Not Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change allows relocating 

main steam and main feedwater valve 

isolation times to the Licensee Controlled 

Document that is referenced in the Bases. 

The proposed change is described in 

Technical Specification Task Force (TSTF) 

Standard TS Change Traveler TSTF–491 

related to relocating the main steam and 

main feedwater valves isolation times to the 

Licensee Controlled Document that is 

referenced in the Bases and replacing the 

isolation time with the phrase, ‘‘within 

limits.’’ 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed). 

The proposed changes relocate the main 

steam and main feedwater isolation valve 

times to the Licensee Controlled Document 

that is referenced in the Bases. The 

requirements to perform the testing of these 

isolation valves are retained in the TS. Future 

changes to the Bases or 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 minimal 

increase in the probability or consequences 

of an accident previously evaluated. 

The proposed changes do not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, and 

configuration of the facility or the manner in 

which the plant is operated and maintained. 

The proposed changes do not adversely affect 

the ability of structures, systems and 

components (SSCs) to perform their intended 

safety function to mitigate the consequences 

of an initiating event within the assumed 

acceptance limits. The proposed changes do 

not affect the source term, containment 

isolation, or radiological consequences of any 

accident previously evaluated. Further, the 

proposed changes do not increase the types 

and the amounts of radioactive effluent that 

may be released, nor significantly increase 

individual or cumulative occupation/public 

radiation exposures. 

Therefore, the changes do not involve a 

significant increase in the probability or 

consequences of any accident previously 

evaluated. 

Criterion 2—The Proposed Change Does 

Not Create the Possibility of a New or 

Different Kind of Accident From Any 

Previously Evaluated 

The proposed changes relocate the main 

steam and main feedwater valve isolation 

times to the Licensee Controlled Document 

that is referenced in the Bases. In addition, 

the valve isolation times are replaced in the 

TS with the phrase ‘‘within limits’’. The 

changes do not involve a physical altering of 

the plant (i.e., no new or different type of 

equipment will be installed) or a change in 

methods governing normal pant operation. 

The requirements in the TS continue to 

require testing of the main steam and main 

feedwater isolation valves to ensure the 

proper functioning of these isolation valves. 

Therefore, the changes do not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

Criterion 3—The Proposed Change Does 

Not Involve a Significant Reduction in the 

Margin of Safety 

The proposed changes relocate the main 

steam and main feedwater valve isolation 

times to the Licensee Controlled Document 

that is referenced in the Bases. In addition, 

the valve isolation times are replaced in the 

TS with the phrase ‘‘within limits.’’ 

Instituting the proposed changes will 

continue to ensure the testing of main steam 

and main feedwater isolation valves. Changes 

to the Bases are license controlled document 

are performed in accordance with 10 CFR 

50.59. This approach provides an effective 

level of regulatory control and ensures that 

main steam and feedwater isolation valve 

testing is conducted such that there is no 

significant reduction in the margin of safety. 

The margin of safety provided by the 

isolation valves is unaffected by the proposed 

changes since there continue to be TS 

requirements to ensure the testing of main 

steam and main feedwater isolation valves. 

The proposed changes maintain sufficient 

controls to preserve the current margins 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 

amendments request involves no 

significant hazards consideration. 

Attorney for licensee: Carey Fleming, 

Sr. Counsel–Nuclear Generation, 

Constellation Generation Group, LLC, 

750 East Pratt Street, 17th floor, 

Baltimore, MD 21202. 

NRC Branch Chief: Mark G. Kowal. 

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: June 19, 

2008. 

Description of amendments request: 

The proposed change would: (1) Revise 

Technical Specifications (TS) control 

rod notch surveillance requirement (SR) 

frequency in TS 3.1.3, ‘‘Control Rod 

Operability,’’ and (2) revise Example 

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

clarify the applicability of the 1.25 

surveillance test extension. The licensee 

mstockstill on PROD1PC66 with NOTICES

58672 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 

is proposing to adopt the approved 

Technical Specification Task Force 

(TSTF) change traveler TSTF–475, 

Revision 1, ‘‘Control Rod Notch Testing 

Frequency.’’ A notice of availability of 

TSTF–475, Revision 1, was published in 

the Federal Register on November 13, 

2007 (72 FR 63935). 

In addition, the proposed amendment 

would remove Note 2 associated with 

SR 3.1.3.3 for Unit 1, which is a cyclespecific note and has expired. This 

change is administrative in nature and 

does not affect the no significant 

hazards consideration (NSHC) 

determination. 

Basis for proposed NSHC 

determination: As required by 10 CFR 

50.91(a), the licensee, in its application 

dated June 19, 2008, affirmed the 

applicability of the published model 

NSHC determination, which is 

presented below: 

Criterion 1—The Proposed Change Does 

Not Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change generically 

implements TSTF–475, Revision 1, ‘‘Control 

Rod Notch Testing Frequency and SRM 

Insert Control Rod Action.’’ TSTF–475, 

Revision 1 modifies NUREG–1433 (BWR/4) 

and NUREG–1434 (BWR/6) STS. The 

changes: (1) revise TS testing frequency for 

surveillance requirement (SR) 3.1.3.2 in TS 

3.1.3, ‘‘Control Rod OPERABILITY’’, (2) [not 

applicable to BSEP], and (3) revise Example 

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

the applicability of the 1.25 surveillance test 

interval extension. The consequences of an 

accident after adopting TSTF–475, Revision 

1 are no different than the consequences of 

an accident prior to adoption. Therefore, this 

change does not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

Criterion 2—The Proposed Change Does 

Not Create the Possibility of a New or 

Different Kind of Accident From Any 

Accident Previously Evaluated 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. The proposed change will 

not introduce new failure modes or effects 

and will not, in the absence of other 

unrelated failures, lead to an accident whose 

consequences exceed the consequences of 

accidents previously analyzed. Thus, this 

change does not create the possibility of a 

new or different kind of accident from any 

accident previously evaluated. 

Criterion 3—The Proposed Change Does 

Not Involve a Significant Reduction in the 

Margin of Safety 

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

TS SR 3.1.3.2 frequency in TS 3.1.3, ‘‘Control 

Rod OPERABILITY’’, (2) [not applicable to 

BSEP], and (3) revise Example 1.4–3 in 

Section 1.4 ‘‘Frequency’’ to clarify the 

applicability of the 1.25 surveillance test 

interval extension. The GE Nuclear Energy 

Report, ‘‘CRD Notching Surveillance Testing 

for Limerick Generating Station,’’ dated 

November 2006, concludes that extending 

the control rod notch test interval from 

weekly to monthly is not expected to impact 

the reliability of the scram system and that 

the analysis supports the decision to change 

the surveillance frequency. Therefore, the 

proposed changes in TSTF–475, Revision 1 

[. . .] do not involve a significant reduction 

in a margin of safety. 

Based on the review of the above 

analysis, the NRC staff finds that the 

three standards in 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves NSHC. 

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 Nos. 50–325 and 50–324, 

Brunswick Steam Electric Plant, Units 

1 and 2, Brunswick County, North 

Carolina 

Date of amendments request: June 19, 

2008. 

Description of amendments request: 

The proposed change would revise 

Limiting Condition for Operation (LCO) 

3.10.1, and the associated Bases, to 

expand its scope to include provisions 

for temperature excursions greater than 

212 degrees Fahrenheit (°F) as a 

consequence of inservice leak and 

hydrostatic testing, and as a 

consequence of scram time testing 

initiated in conjunction with an 

inservice leak or hydrostatic test, while 

considering operational conditions to be 

in Mode 4. 

The NRC issued a ‘‘Notice of 

Availability of Model Application on 

Technical Specification Improvement to 

Modify Requirements Regarding LCO 

3.10.1, Inservice Leak and Hydrostatic 

Testing Operation Using the 

Consolidated Line Item Improvement 

Process,’’ associated with Technical 

Specification Task Force (TSTF) 

Improved Standard Technical 

Specification Change Traveler, TSTF– 

484, Revision 0, in the Federal Register 

on October 27, 2006 (71 FR 63050). The 

NRC also issued a Federal Register 

notice on August 21, 2006 (71 FR 48561) 

that provided a model safety evaluation 

and a model no significant hazards 

consideration (NSHC) determination 

relating to modification of requirements 

regarding LCO 3.10.1, ‘‘Inservice Leak 

and Hydrostatic Testing Operation.’’ In 

its application dated June 19, 2008, the 

licensee affirmed the applicability of the 

model NSHC determination. 

Basis for proposed NSHC 

determination: As required by 10 CFR 

Part 50.91(a), an analysis of the issue of 

NSHC determination is presented 

below: 

Criterion 1: The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

Technical Specifications currently allow 

for operation at greater than 212 °F while 

imposing MODE 4 requirements in addition 

to the secondary containment requirements 

required to be met. Extending the activities 

that can apply this allowance will not 

adversely impact the probability or 

consequences of an accident previously 

evaluated. Therefore, the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

Criterion 2: The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Accident 

Previously Evaluated 

Technical Specifications currently allow 

for operation at greater than 212 °F while 

imposing MODE 4 requirements in addition 

to the secondary containment requirements 

required to be met. No new operational 

conditions beyond those currently allowed 

by LCO 3.10.1 are introduced. The changes 

do not involve a physical alteration of the 

plant (i.e., no new or different type of 

equipment will be installed) or a change in 

the methods governing normal plant 

operation. In addition, the changes do not 

impose any new or different requirements or 

eliminate any existing requirements. The 

changes do not alter assumptions made in the 

safety analysis. The proposed changes are 

consistent with the safety analysis 

assumptions and current plant operating 

practice. Therefore, the proposed change 

does not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

Criterion 3: The Proposed Change Does Not 

Involve a Significant Reduction in a Margin 

of Safety 

Technical Specifications currently allow 

for operation at greater than 212 °F while 

imposing MODE 4 requirements in addition 

to the secondary containment requirements 

required to be met. Extending the activities 

that can apply this allowance will not 

adversely impact any margin of safety. 

Allowing completion of inspections and 

testing and supporting completion of scram 

time testing initiated in conjunction with an 

inservice leak or hydrostatic test prior to 

power operation results in enhanced safe 

operations by eliminating unnecessary 

maneuvers to control reactor temperature and 

pressure. Therefore, the proposed change 

does not involve a significant reduction in a 

margin of safety. 

Based on the above, the proposed 

change presents NSHCs under the 

standards set forth in 10 CFR 50.92(c). 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves NSHC. 

Attorney for licensee: David T. 

Conley, Associate General Counsel II– 

mstockstill on PROD1PC66 with NOTICES

Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 58673 

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, NC 27602. 

NRC Branch Chief: Thomas H. Boyce. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–3, Indian Point Energy 

Center, Unit 1 Westchester County, New 

York 

Date of amendment request: June 26, 

2008. 

Description of amendment request: 

The proposed amendment would delete 

license conditions and Technical 

Specification (TS) requirements which 

relate to the storage of spent nuclear fuel 

in the Indian Point Unit 1 (IP1) Fuel 

Handling Building Spent Fuel Pool. The 

spent fuel is to be transferred to, and 

stored at, the existing Indian Point 

Independent Spent Fuel Storage 

Installation (ISFSI), Docket No. 72–51. 

The removal of the stored spent fuel and 

drain down of the spent fuel pools 

renders many of the license conditions 

and TS requirements unnecessary and 

burdensome. 

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. Will operation of the facility in 

accordance with this proposed change 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated? 

No. The proposed changes are all 

contingent on the prior removal of the stored 

spent fuel from the IP1 Spent Fuel Pool (SFP) 

to the Indian Point Energy Center (IPEC) 

ISFSI. The accidents previously evaluated in 

the IP1 Final Safety Analysis Report (FSAR), 

which consists of the IP1 Decommission Plan 

and Supplemental Environmental 

Information, are stored fuel related accidents. 

The removal of the stored fuel from the IP1 

facility to the IPEC ISFSI precludes the 

possibility of these accidents. 

Consequently, the proposed changes to the 

license do not involve a significant increase 

in the probability or the consequences of an 

accident previously evaluated. 

2. Does the proposed change create the 

probability of a new or different accident 

from any accident previously evaluated? 

The proposed changes are all contingent on 

the prior removal of the stored spent fuel 

from the IP1 SFP to the IPEC ISFSI. With the 

removal of the stored spent fuel from the IP1 

facility, and considering the IP1 has been in 

a SAFESTOR mode for over thirty years, no 

significant source term remains which could 

result in any postulated radiological event 

that would impact the health and safety of 

the public. 

Therefore, the proposed changes to the IP1 

license consequently do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

No. The proposed changes are all 

contingent on the prior removal of the stored 

spent fuel from the IP1 SFP to the IPEC 

ISFSI. Upon the removal of spent fuel, the 

Technical Specifications being deleted no 

longer are required to protect the health and 

safety of the public or occupational workers 

from the potential adverse conditions, 

hazards or accidents as discussed in the 

FSAR. 

Therefore, operation of the facility in 

accordance with the proposed amendments 

would not involve a significant reduction in 

the margin of safety. 

Based upon the reasoning presented 

above 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: Assistant 

General Counsel, Entergy Nuclear 

Operations, Inc., 440 Hamilton Avenue, 

White Plains, NY 10601. 

NRC Branch Chief: Theodore Smith, 

Acting. 

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 amendment request: 

September 11, 2008. 

Description of amendment request: 

The amendment would revise several 

surveillance requirements (SRs) and add 

SR 3.8.1.21 in Technical Specification 

(TS) 3.8.1, ‘‘AC [alternating current] 

Sources—Operating,’’ and TS 3.8.2, ‘‘AC 

Sources—Shutdown.’’ The amendment 

would allow the slow-start testing 

sequence of the diesel generators in 

order to reduce the stress and wear on 

the equipment. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change affects the 

surveillance requirements for the Diesel 

Generators (DGs). The DGs are onsite standby 

power sources intended to provide 

redundant and reliable power to ESF 

[Engineered Safety Feature] systems credited 

as accident mitigating features in design basis 

[accident] analyses. Per NRC Regulatory 

Guide (RG) 1.9, Revision 3, which is 

referenced in Grand Gulf Nuclear Station 

(GGNS) UFSAR [Updated Final Safety 

Analysis Report Section] 8.3.1.2.1, the 

proposed change is intended to allow slower 

starts of the DGs during testing in order to 

reduce DG aging effects due to excessive 

testing conditions. As such, the proposed 

change will result in improved DG reliability 

and availability, thereby providing additional 

assurance that the DGs will be capable of 

performing their safety function. The method 

of starting the emergency diesel generators 

for testing purposes does not affect the 

probability of any previously evaluated 

accident. Although the change allows slower 

starts for the monthly tests, the more rapid 

start function, assumed in the accident 

analysis, is unchanged and will be verified 

on a 184 day frequency. Therefore the 

accident analysis consequences are not 

affected [by the proposed change]. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability [or] consequences of any 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 affects the 

surveillance requirements for the onsite ac 

sources, i.e. the Diesel Generators. 

Accordingly, the proposed change does not 

involve any change to the configuration or 

method of operation of any plant equipment 

that could cause an accident. In addition, no 

new failure modes have been created nor has 

any new limiting failure been introduced as 

a result of the proposed surveillance changes. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Will operation of the facility in 

accordance with this proposed change 

involve a significant reduction in a margin of 

safety? 

Response: No. 

The proposed change is intended to bring 

the existing GGNS TS requirements for the 

onsite AC sources in line with regulatory 

guidance. Under the proposed change, the 

DGs will remain capable of performing their 

safety function, and the effects of aging on 

the DGs will be reduced by eliminating 

unnecessary testing. The DG start times 

assumed in the current accident analyses are 

unchanged and will be verified on a 184 day 

frequency. 

Therefore, the proposed change does not 

involve a significant reduction in the margin 

of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Terence A. 

Burke, Associate General Counsel— 

Nuclear Entergy Services, Inc., 1340 

mstockstill on PROD1PC66 with NOTICES

58674 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 

Echelon Parkway, Jackson, Mississippi 

39213. 

NRC Branch Chief: Michael T. 

Markley. 

GE Hitachi Nuclear Energy (GEH), 

License No. DR–10, Docket No. 50–183, 

ESADA Vallecitos Experimental 

Superheat Reactor (EVESR) 

Date of amendment request: June 23, 

2008. 

Description of amendment request: 

The proposed license amendment 

would modify the Technical 

Specification (TS) requirements to 

revise the scope of dismantling 

activities that GEH can perform under 

The Vallecitos Nuclear Center Liabilities 

Reduction Project and specify 

radiological control requirements of 10 

CFR Part 20. Two TS changes are 

proposed. The proposed changes to the 

TS: 

• Allow GEH to conduct dismantling 

activities below the 549-ft elevation 

level within the containment building; 

and 

• Revise the physical security 

requirements for access to areas below 

the 549-ft elevation level within the 

containment building. 

The application for license 

amendment is available electronically at 

the NRC’s Electronic Reading Room at 

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

adams.html. From this site, you can 

access the NRC’s Agencywide 

Document Access and Management 

System (ADAMS), which provides text 

and image files of NRC’s public 

documents. The ADAMS accession 

number for the June 23, 2008, request is 

ML081780099. 

If you do not have access to ADAMS, 

or if there are problems in accessing the 

documents located in ADAMS, contact 

the NRC Public Document Room (PDR) 

Reference staff at 1–800–397–4209, 301– 

415–4737, or by e-mail to pdr@nrc.gov. 

These documents may also be viewed 

electronically on the public computers 

located at the NRC’s PDR, 01F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, MD 20852. The PDR 

reproduction contractor will copy 

documents for a fee. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

(1) Does the change involve a significant 

increase in the probability or consequences 

of an accident previously evaluated? 

Response: No. 

Proposed change one is an administrative 

change submitted to clarify the area where 

dismantling activities will occur as 

authorized by the facility license. The 

majority of the component removal activities 

will occur in areas below the 549-ft. 

elevation. Proposed change two removes the 

specific shielding and covering requirements 

for the reactor vessel, shield plug storage pit 

and the empty spent fuel storage pit and 

modifies the access control requirements to 

be consistent with 10 CFR 20. The EVESR 

reactor was shutdown in 1967 and has 

remained in a ‘‘Possess Only’’ status. All fuel 

bundles were removed from the facility and 

the radiation and contamination levels have 

been reduced by the removal of radioactive 

material and natural decay. No aspect of the 

proposed changes will involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

(2) Does the change create the possibility 

of a new or different kind of accident from 

any accident evaluated? 

Response: No. 

Proposed change one is an administrative, 

therefore there it cannot create a new or 

different kind of accident. Removal of the 

specific shielding and covering requirements 

for the reactor vessel, shield plug storage pit 

and the empty spent fuel storage pit and 

modification of the access control 

requirements as described in proposed 

change two will not impact the function or 

integrity of the reactor pressure vessel, which 

is the primary safety system required to be 

maintained by the license. The proposed 

changes do not create the possibility of a new 

or different kind of accident from any 

accident evaluated. 

(3) Does the change involve a significant 

reduction in a margin of safety? 

Response: No. 

Removal of the specific shielding, covering 

and access control requirements will not 

result in a reduction of the margin of the 

safety for the EVESR facility. These controls 

were implemented to provide shielding and 

access controls to High Radiation Areas. 

Since the reactor is no longer operating and 

the radiological conditions have been 

significantly reduced, the specific controls 

specified in the current technical 

specifications are not required. All areas in 

the EVESR containment will be controlled in 

accordance with 10 CFR 20. High Radiation 

areas will be controlled in a manner 

consistent with the requirements of 10 CFR 

20.1601. The proposed changes do not affect 

the margins of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based upon the 

staff’s review of the licensee’s analysis, 

as well as the staff’s own evaluation, the 

staff concludes 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. 

GEH, Manager, Regulatory 

Compliance & EHS: LaTonya L. 

Mahlahla. 

NRC Branch Chief: Andrew Persinko. 

Nine Mile Point Nuclear Station, LLC, 

(NMPNS) Docket Nos. 50–220 and 50– 

410, Nine Mile Point Nuclear Station 

Unit Nos. 1 and 2 (NMP 1 and 2), 

Oswego County, New York 

Date of amendment request: June 24, 

2008. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specifications (TSs) by (1) 

replacing the references to Section XI of 

the American Society of Mechanical 

Engineers (ASME) Boiler and Pressure 

Vessel Code with references to the 

ASME Code for Operation and 

Maintenance of Nuclear Power Plants 

(OM Code); and (2) revising the 

allowance to extend Inservice Testing 

(IST) frequencies by 25 percent to 

clearly state that the allowance is 

applicable to IST frequencies of 2 years 

or less. The proposed changes are based 

on TS Task Force (TSTF) Standard 

Technical Specification Change Traveler 

479–A, 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: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes revise the IST 

Program sections of the NMP1 and NMP2 TS 

to maintain consistency with the 

requirements of 10 CFR 50.55a(f)(4) regarding 

the IST of pumps and valves that are 

classified as ASME Code Class 1, Class 2, and 

Class 3. The proposed changes incorporate 

revisions to the ASME Code that result in a 

net improvement in the measures for testing 

pumps and valves. The proposed changes 

also revise the allowance to extend IST 

frequencies by 25 percent to clearly state that 

this allowance is applicable to IST 

frequencies of 2 years or less. 

The proposed TS changes are 

administrative in nature. They do not impact 

any accident initiators, the ability to mitigate 

previously evaluated accidents, or the 

assumptions used in evaluating the 

radiological consequences of previously 

evaluated accidents. The proposed changes 

do not involve the addition or removal of any 

equipment, or any design changes to the 

facilities. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

mstockstill on PROD1PC66 with NOTICES

Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 58675 

The proposed changes revise the IST 

Program sections of the NMP1 and NMP2 TS 

to maintain consistency with the 

requirements of 10 CFR 50.55a(f)(4) regarding 

the IST of pumps and valves that are 

classified as ASME Code Class 1, Class 2, and 

Class 3. The proposed changes incorporate 

revisions to the ASME Code that result in a 

net improvement in the measures for testing 

pumps and valves. The proposed changes 

also revise the allowance to extend IST 

frequencies by 25 percent to clearly state that 

this allowance is applicable to IST 

frequencies of 2 years or less. 

The proposed TS changes are 

administrative in nature. They do not involve 

a modification to the physical configuration 

of the plants (i.e., no new equipment will be 

installed) or involve a change in the methods 

governing normal plant operation. The 

proposed changes will not impose any new 

or different requirements or introduce a new 

accident initiator, accident precursor, or 

failure mechanism. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed TS changes are 

administrative in nature. They do not involve 

a modification to the physical configuration 

of the plants (i.e., no new equipment will be 

installed) or change the methods governing 

normal plant operation. The proposed 

changes do not modify the safety limits or 

setpoints at which protective actions are 

initiated, and do not change the requirements 

governing operation or availability of safety 

equipment assumed to operate to preserve 

margins of safety. The incorporation of 

revisions to the ASME Code results 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 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: Mark J. 

Wetterhahn, Esquire, Winston & Strawn, 

1700 K Street, NW., Washington, DC 

20006. 

NRC Branch Chief: Mark G. Kowal. 

PPL Susquehanna, LLC, Docket Nos. 

50–387 and 50–388, Susquehanna 

Steam Electric Station, Units 1 and 2, 

Luzerne County, Pennsylvania 

Date of amendment request: July 7, 

2008. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specification (TS) testing 

frequency for the Surveillance 

Requirement (SR) in TS 3.1.4, ‘‘Control 

Rod Scram Times.’’ The proposed 

change revises 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.’’ The NRC staff issued a notice 

of availability of a model no significant 

hazards consideration determination 

(NSHCD) for referencing in licensing 

amendment applications in the Federal 

Register on August 23, 2004 (69 FR 

51864) using the consolidated line item 

improvement process (CLIIP). The 

licensee affirmed the applicability of the 

model NSHC determination and the 

model safety evaluation in its 

application dated July 7, 2008. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of NSHC, based on 

the model NSHCD published in the 

Federal Register on August 23, 2004 (69 

FR 51864), is presented below: 

1. Does the change involve a significant 

increase in the probability or consequences 

of an accident, previously evaluated? 

Response: No. 

The proposed change extends the 

frequency for testing control rod scram time 

testing from every 120 days of cumulative 

Mode 1 operation to 200 days of cumulative 

Mode 1 operation. The frequency of 

surveillance testing is not an initiator of any 

accident previously evaluated. The frequency 

of surveillance testing does not affect the 

ability to mitigate any accident previously 

evaluated, as the tested component is still 

required to be operable. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

Response: No. 

The proposed change extends the 

frequency for testing control rod scram time 

testing from every 120 days of cumulative 

Mode 1 operation to 200 days of cumulative 

Mode 1 operation. The proposed change does 

not result in any new or different modes of 

plant operation. 

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 continues to test the 

control rod scram time to ensure the 

assumptions in the safety analysis are 

protected. Therefore, the proposed change 

does not involve a significant reduction in a 

margin of safety. 

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Bryan A. Snapp, 

Esquire, Assoc. General Counsel, PPL 

Services Corporation, 2 North Ninth St., 

GENTW3, Allentown, PA 18101–1179. 

NRC Branch Chief: Mark Kowal. 

PPL Susquehanna, LLC, Docket Nos. 

50–387 and 50–388, Susquehanna 

Steam Electric Station, Units 1 and 2, 

Luzerne County, Pennsylvania 

Date of amendment request: July 7, 

2008. 

Description of amendment request: 

PPL Susquehanna, LLC (the licensee) 

requests adoption of 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. 

The NRC staff issued a notice of 

availability in the Federal Register on 

November 13, 2007, (72 FR 63935), 

which included a model safety 

evaluation (SE) and model no 

significant hazards consideration 

determination (NSHCD), using the 

consolidated line-item improvement 

process (CLIIP), of possible amendments 

to revise the plant specific TS, to allow: 

(1) Revising the frequency of SR 3.1.3.2, 

notch testing of fully withdrawn control 

rod, from ‘‘7 days after the control rod 

mstockstill on PROD1PC66 with NOTICES

58676 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 

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 of the RWM’’, (2) adding the 

word ‘‘fully’’ to LCO 3.3.1.2 Required 

Action E.2 to clarify the requirement to 

fully insert all insertable control rods in 

core cells containing one or more fuel 

assemblies when the associated SRM 

instrument is inoperable, and (3) 

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. The licensee 

affirmed the applicability of the model 

SE and model NSHC determination in 

its application dated July 7, 2008. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of NSHC, based on 

the model NSHCD published in the 

Federal Register on November 13, 2007 

(72 FR 63935), is presented below: 

Criterion 1—The Proposed Change 

Does Not Involve a Significant Increase 

in the Probability or Consequences of an 

Accident Previously Evaluated 

The proposed change generically 

implements TSTF–475, Revision 1, 

‘‘Control Rod Notch Testing Frequency 

and SRM Insert Control Rod Action.’’ 

TSTF–475, Revision 1 modifies 

NUREG–1433 (BWR/4) and NUREG– 

1434 (BWR/6) STS. The changes: (1) 

revise TS testing frequency for 

surveillance requirement (SR) 3.1.3.2 in 

TS 3.1.3, ‘‘Control Rod OPERABILITY’’, 

(2) clarify the requirement to fully insert 

all insertable control rods for the 

limiting condition for operation (LCO) 

in TS 3.3.1.2, Required Action E.2, 

‘‘Source Range Monitoring 

Instrumentation’’ (NUREG–1434 only), 

and (3) revise Example 1.4–3 in Section 

1.4 ‘‘Frequency’’ to clarify the 

applicability of the 1.25 surveillance 

test interval extension. Implementing 

TSTF–475, Revision 1 does not change 

the control rod notch test method. 

Implementing TSTF–475, Revision 1 

decreases the performance frequency of 

the control rod notch test. Therefore, the 

proposed change does not involve a 

significant increase in the probability of 

an accident previously evaluated. The 

consequences of an accident after 

adopting TSTF–475, Revision 1 are no 

different than the consequences of an 

accident prior to adoption. Therefore, 

this change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Criterion 2—The Proposed Change 

Does Not Create the Possibility of a New 

or Different Kind of Accident from any 

Accident Previously Evaluated 

The proposed change does not 

involve a physical alteration of the plant 

(no new or different type of equipment 

will be installed) or a change in the 

methods governing normal plant 

operation. The proposed change will not 

introduce new failure modes or effects 

and will not, in the absence of other 

unrelated failures, lead to an accident 

whose consequences exceed the 

consequences of accidents previously 

analyzed. Thus, this change does not 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated. 

Criterion 3—The Proposed Change 

Does Not Involve a Significant 

Reduction in the Margin of Safety 

The proposed amendment will: (1) 

Revise the TS SR 3.1.3.2 frequency in 

TS 3.1.3, ‘‘Control Rod OPERABILITY’’, 

and (2) revise Example 1.4–3 in Section 

1.4 ‘‘Frequency’’ to clarify the 

applicability of the 1.25 surveillance 

test interval extension. The GE Nuclear 

Energy Report, ‘‘CRD Notching 

Surveillance Testing for Limerick 

Generating Station,’’ dated November 

2006, concludes that extending the 

control rod notch test interval from 

weekly to monthly is not expected to 

impact the reliability of the scram 

system and that the analysis supports 

the decision to change the surveillance 

frequency. Therefore, the proposed 

changes in TSTF–475, Revision 1 do not 

involve a significant reduction in a 

margin of safety. 

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Bryan A. Snapp, 

Esquire, Assoc. General Counsel, PPL 

Services Corporation, 2 North Ninth St., 

GENTW3, Allentown, PA 18101–1179. 

NRC Branch Chief: Mark Kowal. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of amendment request: July 21, 

2008. 

Description of amendment request: 

The proposed amendments would 

delete the requirements related to plant 

staff working hours from Section 6.0, 

‘‘Administrative Controls’’ of the 

respective plants’ Technical 

Specifications (TSs). The current 

working hour requirements were 

incorporated into the TSs as a result of 

the guidance in Nuclear Regulatory 

Commission (NRC) Generic Letter (GL) 

82–12, ‘‘Nuclear Power Plant Staff 

Working Hours.’’ The guidance in GL 

82–12 has been superseded by the 

requirements of Title 10 of the Code of 

Federal Regulations (10 CFR), Part 26, 

‘‘Fitness for Duty Programs,’’ Subpart I, 

‘‘Managing Fatigue’’ which was 

published in the Federal Register on 

March 31, 2008, as part of the final 

rulemaking for Part 26. As discussed in 

the Federal Register notice for the final 

rule (73 FR 16966), Subpart I must be 

implemented by licensees no later than 

October 1, 2009. The licensee stated that 

the proposed amendments would 

support implementation of the new 

requirements in Subpart I. 

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 removal of GL 82–12 administrative 

controls will not remove the requirement to 

control work hours and manage fatigue. 

Removal of TS controls required by GL 82– 

12 will be performed concurrently with the 

implementation of the more conservative [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. 

Because these new requirements are more 

conservative with respect to work hour 

controls and fatigue management, this will 

not significantly increase the probability or 

consequence 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 GL 82–12 

administrative controls from [the] TS to 

support the implementation of Subpart I to 

[10 CFR Part 26]. The Subpart I regulations 

are more restrictive than the current guidance 

in [the] TS and would add conservatism to 

work hour controls and fatigue management. 

Work hours will continue to be controlled in 

accordance with NRC requirements. The new 

rule continues to allow for deviations from 

controls to mitigate or prevent a condition 

adverse to safety or necessary to maintain the 

security of the facility. This ensures that the 

new rule will not restrict work hours at the 

mstockstill on PROD1PC66 with NOTICES

Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 58677 

expense of the health and safety of the public 

as well as plant personnel. 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 impact the function of plant 

SSCs or the manner in which SSCs are 

operated, maintained, modified, tested, or 

inspected. 

Because the proposed changes do not 

remove the station’s requirement to control 

work hours and increases the conservatism of 

work hour controls by changing 

administrative scheduling requirements, the 

proposed changes do not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

An input to maintaining the margin of 

safety is the control of work hours in 

managing fatigue. Salem and Hope Creek 

Generating Stations will continue their 

fitness-for-duty and behavioral observation 

programs, both of which will be strengthened 

by compliance with the new Part 26 

regulation. The proposed changes add 

conservatism to fatigue management and 

contribute to the margin of safety. The 

proposed changes do not involve any 

physical changes to plant SSCs or the manner 

in which SSCs are operated, maintained, 

modified, tested, or inspected. The proposed 

changes do not involve a change to any safety 

limits, limiting safety system settings, 

limiting conditions of operation, or design 

parameters for any SSC. The proposed 

changes do not impact any safety analysis 

assumptions and do not involve a change in 

initial conditions, system response times, or 

other parameters affecting an accident 

analysis. Therefore, the proposed changes do 

not involve a significant reduction in the 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, with changes in the areas noted 

above, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jeffrie J. Keenan, 

Esquire, Nuclear Business Unit—N21, 

P.O. Box 236, Hancocks Bridge, NJ 

08038. 

NRC Branch Chief: Harold K. 

Chernoff. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of amendment request: July 30, 

2008. 

Description of amendment request: 

The proposed amendment would 

relocate Technical Specification (TS) 3/ 

4.7.5, ‘‘Snubbers,’’ to the Hope Creek 

Generating Station (HCGS) Technical 

Requirements Manual (TRM). TS 

6.10.3.l, which specifies retention 

requirements for records of snubber 

service life monitoring pursuant to TS 

4.7.5, would also be relocated to the 

TRM. In addition, the amendment 

would add new TS Limiting Condition 

for Operation (LCO) 3.0.8, 

‘‘Inoperability of Snubbers,’’ and would 

modify LCO 3.0.1 to reference LCO 

3.0.8. 

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 to relocate TS 3/4.7.5 

to the TRM is administrative in nature and 

does not involve the modification of any 

plant equipment or affect basic plant 

operation. Snubber operability and 

surveillance requirements will be contained 

in the TRM to ensure design assumptions for 

accident mitigation are maintained. 

The proposed change to add LCO 3.0.8 

allows a delay time for entering a supported 

system technical specification (TS) when the 

inoperability is due solely to an inoperable 

snubber if risk is assessed and managed. 

Entrance into TS actions or delaying entrance 

into actions is not an initiator of any accident 

previously evaluated. Consequently, the 

probability of an accident previously 

evaluated is not significantly increased. The 

consequences of an accident while relying on 

[the] allowance provided by proposed LCO 

3.0.8 are no different than the consequences 

of an accident while relying on the current 

TS required actions in effect without the 

allowance provided by proposed LCO 3.0.8. 

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 change to relocate TS 3/4.7.5 

to the TRM is administrative and does not 

involve any physical alteration of plant 

equipment. The proposed change does not 

change the method by which any safetyrelated system performs its function. As 

such, no new or different types of equipment 

will be installed, and the basic operation of 

installed equipment is unchanged. The 

methods governing plant operation and 

testing remain consistent with current safety 

analysis assumptions. Therefore, the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed change to add LCO 3.0.8 

does not involve a physical alteration of the 

plant (no new or different type of equipment 

will be installed). Allowing delay times for 

entering supported system TS when 

inoperability is due solely to inoperable 

snubbers, if risk is assessed and managed, 

will not introduce new failure modes or 

effects. 

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 to relocate TS 3/4.7.5 

to the TRM is administrative in nature, does 

not negate any existing requirement, and 

does not adversely affect existing plant safety 

margins or the reliability of the equipment 

assumed to operate in the safety analysis. As 

such, there are no changes being made to 

safety analysis assumptions, safety limits or 

safety system settings that would adversely 

affect plant safety as a result of the proposed 

change. Margins of safety are unaffected by 

requirements that are retained, but relocated 

from the TS to the TRM. 

The proposed change to add LCO 3.0.8 to 

[the] TS allows a delay time before declaring 

supported TS systems inoperable when the 

associated snubber(s) cannot perform the 

required safety function. The proposed 

change retains an allowance in the current 

HCGS TS while upgrading it to be more 

conservative for snubbers supporting 

multiple trains or sub-systems of an 

associated system. The updated TS will 

continue to provide an adequate margin of 

safety for plant operation upon incorporation 

of LCO 3.0.8. The station design and safety 

analysis assumptions provide margin in the 

form of redundancy to account for periods of 

time when system capability is reduced. 

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, with changes in the areas noted 

above, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jeffrie J. Keenan, 

Esquire, Nuclear Business Unit—N21, 

P.O. Box 236, Hancocks Bridge, NJ 

08038. 

NRC Branch Chief: Harold K. 

Chernoff. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–424 and 50–425, 

Vogtle Electric Generating Plant, Units 

1 and 2, Burke County, Georgia 

Date of amendment request: August 

12, 2008. 

Description of amendment request: 

The proposed amendment deletes 

License Condition 2.H, which requires 

reporting of violations of operating 

license requirements found in license 

condition 2.C. 

mstockstill on PROD1PC66 with NOTICES

58678 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the change involve a significant 

increase in the probability or consequences 

of an accident previously evaluated? 

Response: No. 

The proposed change involves the deletion 

of a reporting requirement. The change does 

not affect plant equipment or operating 

practices and therefore does not significantly 

increase the probability or consequences of 

an accident previously evaluated. 

2. Does the change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

Response: No. 

The proposed change is administrative in 

that it deletes a reporting requirement. The 

change does not add new plant equipment, 

change existing plant equipment, or affect the 

operating practices of the facility. Therefore, 

the change does not create the possibility of 

a new or different kind of accident from any 

accident previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change deletes a reporting 

requirement. The change does not affect 

plant equipment or operating practices and 

therefore does not involve a significant 

reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Arthur H. 

Domby, Troutman Sanders, 

NationsBank Plaza, Suite 5200, 600 

Peachtree Street, NE., Atlanta, Georgia 

30308–2216. 

NRC Branch Chief: Melanie C. Wong. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: June 3, 

2008. 

Description of amendment request: 

The proposed changes will revise 

Technical Specifications (TSs) 3.3.7, 

3.3.8, 3.7.10, 3.7.13, 3.8.2, 3.8.5, 3.8.8, 

and 3.8.10. This amendment will (1) 

delete MODES 5 and 6 from the Control 

Room Emergency Ventilation System 

and its actuation instrumentation in TS 

3.7.10 and TS 3.3.7; (2) adopt U.S. 

Nuclear Regulatory Commission (NRC)- 

approved traveler TSTF–36–A for TSs 

3.3.8, 3.7.13, 3.8.2, 3.8.5, 3.8.8, and 

3.8.10; and (3) add a more restrictive 

change to the Limiting Condition for 

Operation (LCO) Applicability for TSs 

3.8.2, 3.8.5, 3.8.8, and 3.8.10 such that 

these LCOs apply not only during 

MODES 5 and 6, but also during the 

movement of irradiated fuel assemblies 

regardless of the MODE in which the 

plant is operating. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes to delete MODES 5 

and 6 from the LCO Applicability of 

Technical Specifications (TSs) 3.3.7 and 

3.7.10, adopt TSTF–36–A, and revise the 

LCO Applicability of the shutdown electrical 

specifications to be more restrictive does not 

alter plant design or operation; therefore, 

these changes will not increase the 

probability of any accident. 

Overall protection system performance will 

remain within the bounds of the previously 

performed accident analyses since there are 

no design changes. All design, material, and 

construction standards that were applicable 

prior to this amendment request will be 

maintained. There will be no changes to any 

design or operating limits. 

The proposed changes will not adversely 

affect accident initiators or precursors nor 

adversely alter the design assumptions, 

conditions, and configuration of the facility 

or the manner in which the plant is operated 

and maintained. The proposed changes will 

not alter or prevent the ability of structures, 

systems, and components (SSCs) from 

performing their intended functions to 

mitigate the consequences of an initiating 

event within the assumed acceptance limits. 

The proposed changes do not physically 

alter safety-related systems nor affect the way 

in which safety-related systems perform their 

functions. 

Deleting MODES 5 and 6 from the LCO 

Applicability of TSs 3.3.7 and 3.7.10 does not 

significantly increase the consequences of 

any accident since it has been demonstrated 

that the radiological consequences to control 

room occupants from a waste gas decay tank 

rupture will remain much less than the 

regulatory limits with no mitigation from the 

Control Room Emergency Ventilation System 

(CREVS) in MODES 5 and 6. The acceptance 

criteria for this event will continue to be met. 

The adoption of TSTF–36–A will not affect 

the equipment and LCOs needed to mitigate 

the consequences of a fuel handling accident 

in the fuel building; however, this change 

will reduce the chances of an unnecessary 

plant shutdown due to activities in the fuel 

building that have no bearing on the 

operation of the rest of the plant and the 

reactor core inside the containment building. 

The changes to the shutdown electrical 

specifications will add an additional 

restriction that is consistent with the 

objective of being able to mitigate a fuel 

handling accident during all situations, 

including a full core offload, in which such 

an accident could occur. 

All accident analysis acceptance criteria 

will continue to be met with the proposed 

changes. The proposed changes will not 

affect the source term, containment isolation, 

or radiological release assumptions used in 

evaluating the radiological consequences of 

an accident previously evaluated. After a 

postulated release from a waste gas decay 

tank rapture no CREVS mitigation is 

required. The applicable radiological dose 

criteria will continue to be met. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

There are no proposed design changes nor 

are there any changes in the method by 

which any safety-related plant structure, 

system, or component (SSC) performs its 

specified safety function. The proposed 

changes will not affect the normal method of 

plant operation or change any operating 

parameters. Equipment performance 

necessary to fulfill safety analysis missions 

will be unaffected. The proposed changes 

will not alter any assumptions required to 

meet the safety analysis acceptance criteria. 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures will be introduced as a result 

of this amendment. There will be no adverse 

effect or challenges imposed on any safetyrelated system as a result of this amendment. 

The proposed amendment will not alter the 

design or performance of the 7300 Process 

Protection System, Nuclear Instrumentation 

System, or Solid State Protection System 

used in the plant protection systems. 

The proposed changes to delete MODES 5 

and 6 from the LCO Applicability of TSs 

3.3.7 and 3.7.10, adopt TSTF–36–A, and 

revise the LCO Applicability of the shutdown 

electrical specifications to be more restrictive 

do not, therefore, create the possibility of a 

new or different accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

There will be no effect on those plant 

systems necessary to assure the 

accomplishment of protection functions. 

There will be no impact on the overpower 

limit, departure from nucleate boiling ratio 

(DNBR) limits, heat flux hot channel factor 

(FQ), nuclear enthalpy rise hot channel-factor 

(FAH), loss of coolant accident peak cladding 

temperature (LOCA PCT), peak local power 

density, or any other margin of safety. The 

applicable radiological dose consequence 

acceptance criteria will continue to be met. 

It has been demonstrated that the CREVS and 

its actuation instrumentation are not required 

to mitigate the control room radiological 

consequences of a waste gas decay tank 

rupture. 

The proposed changes do not eliminate 

any surveillances or alter the frequency of 

mstockstill on PROD1PC66 with NOTICES

Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 58679 

surveillances required by the Technical 

Specifications. None of the acceptance 

criteria for any accident analysis will be 

changed. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: John O’Neill, 

Esq., Pillsbury Winthrop Shaw Pittman 

LLP, 2300 N Street, NW., Washington, 

DC 20037. 

NRC Branch Chief: Michael T. 

Markley. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: August 

14, 2008. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specification (TS) 3.3.2, 

‘‘Engineered Safety Feature Actuation 

System (ESFAS)’’ to extend the 

Surveillance Frequency on selected 

ESFAS slave relays from 92 days to 18 

months. Justification for extending the 

slave relay Surveillance Frequency is 

based on information contained in the 

Westinghouse Electric Corporation 

reports WCAP–13878–P–A, Revision 2 

(proprietary version), and WCAP– 

14117–NP–A, Revision 2 

(nonproprietary version), ‘‘Reliability 

Assessment of Potter & Brumfield MDR 

Series Relays,’’ dated August 2000. 

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 (NSHC), 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 will not result in a 

condition where the design, material, and 

construction standards that were applicable 

prior to the change are altered. The same 

Engineered Safety Feature Actuation System 

(ESFAS) instrumentation will be used and 

the same ESFAS system reliability is 

expected. Overall protection system 

performance will remain within the bounds 

of the previously performed accident 

analyses since there are no design changes. 

There will be no changes to any design or 

operating limits. 

The proposed changes will not change 

accident initiators or precursors assumed or 

postulated in the Updated Safety Analysis 

Report (USAR) described accident analyses, 

nor will they alter the design assumptions, 

conditions, and configuration of the facility 

or the manner in which the plant is operated 

and maintained. The proposed changes will 

not alter or prevent the ability of structures, 

systems, and components (SSCs) from 

performing their intended functions to 

mitigate the consequences of an initiating 

event within the assumed acceptance limits. 

The proposed changes do not physically 

alter safety related systems, nor do they affect 

the way in which safety related systems 

perform their functions. All accident analysis 

acceptance criteria will continue to be met 

with the proposed changes. The proposed 

changes will not affect the source term, 

containment isolation, or radiological release 

assumptions used in evaluating the 

radiological consequences of an accident 

previously evaluated. The proposed changes 

will not alter any assumptions or change any 

mitigation actions in the radiological 

consequence evaluations in the USAR. The 

applicable radiological dose acceptance 

criteria will continue to be met. 

Based on the above considerations, the 

proposed changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

There are no proposed design changes, nor 

are there any changes in the method by 

which any safety-related plant SSC performs 

its specified safety function. Changing the 

interval for periodically verifying the ESFAS 

slave relays will not create any new accident 

initiators or scenarios. The proposed changes 

will not affect the normal method of plant 

operation or change any operating 

parameters. No equipment performance 

requirements will be affected. The proposed 

changes will not alter any assumptions made 

in the safety analyses. 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures will be introduced as a result 

of this amendment. There will be no adverse 

effect or challenges imposed on any safetyrelated system as a result of this amendment. 

The proposed amendment will not alter the 

design or performance of the 7300 Process 

Protection System, Nuclear Instrumentation 

System, or Solid State Protection System 

used in the plant protection systems. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

accident from any accident previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in the margin of 

safety? 

Response: No. 

The proposed change will not affect the 

total ESFAS response assumed in the safety 

analysis because the reliability of the slave 

relays will not be significantly affected by the 

increased surveillance interval. The relays 

have demonstrated a high reliability and 

insensitivity to short term wear and aging 

effects. The overall reliability, redundancy, 

and diversity assumed available for the 

protection and mitigation of accident and 

transient conditions is unaffected by this 

proposed change. 

There will be no effect on those plant 

systems necessary to assure the 

accomplishment of protection functions. 

There will be no impact on the overpower 

limit, departure from nucleate boiling ratio 

(DNBR) limits, heat flux hot channel factor 

(F2), nuclear enthalpy rise hot channel factor 

(F∆H), loss of coolant accident peak cladding 

temperature (LOCA PCT), peak local power 

density, or any other margin of safety. The 

applicable radiological dose consequence 

acceptance criteria for design-basis transients 

and accidents will continue to be met. 

None of the acceptance criteria for any 

accident analysis will be changed. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves NSHC. 

Attorney for licensee: Jay Silberg, Esq., 

Pillsbury Winthrop Shaw Pittman LLP, 

2300 N Street, NW., Washington, DC 

20037. 

NRC Branch Chief: Michael T. 

Markley. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: August 

14, 2008. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specification (TS) 3.3.2, 

‘‘Engineered Safety Feature Actuation 

System (ESFAS) Instrumentation,’’ TS 

3.7.2, ‘‘Main Steam Isolation Valves 

(MSIVs),’’ and add New TS 3.7.19, 

‘‘Secondary System Isolation Valves 

(SSIVs).’’ TS 3.7.2 is being revised to 

add MSIV bypass valves to the scope of 

TS 3.7.2. TS Table 3.3.2–1 is being 

revised to reflect the addition of the 

MSIV bypass valves to TS 3.7.2 and the 

associated applicability to be consistent 

with Westinghouse Standard Technical 

Specifications (NUREG–1431, Revision 

31). TS 3.7.19 is being added to include 

a Limiting condition for Operation 

(LCO), Conditions/Required Actions 

and Surveillance Requirements for the 

steam generator blowdown isolation 

valves and steam generator blowdown 

sample isolation valves. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

mstockstill on PROD1PC66 with NOTICES

58680 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration (NSHC), 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 adds requirements to 

the TS to ensure that systems and 

components are maintained consistent with 

the safety analysis and licensing basis. 

Requirements are incorporated into the TS 

for secondary system isolation valves. These 

changes do not involve any design or 

physical changes to the facility, including the 

SSIVs themselves. The design and functional 

performance requirements, operational 

characteristics, and reliability of the SSIVs 

are unchanged. 

Overall protection system performance will 

remain within the bounds of the previously 

performed accident analyses since there are 

no design changes. All design, material, and 

construction standards that were applicable 

prior to this amendment request will be 

maintained. There will be no changes to any 

design or operating limits. 

The proposed changes will not change 

accident initiators or precursors assumed or 

postulated in the Updated Safety Analysis 

Report (USAR) described accident analyses, 

nor will they alter the design assumptions, 

conditions, and configuration of the facility 

or the manner in which the plant is operated 

and maintained. The proposed changes will 

not alter or prevent the ability of structures, 

systems, and components (SSCs) from 

performing their intended functions to 

mitigate the consequences of an initiating 

event within the assumed acceptance limits. 

The proposed changes do not physically 

alter safety related systems, nor do they affect 

the way in which safety related systems 

perform their functions. All accident analysis 

acceptance criteria will continue to be met 

with the proposed changes. The proposed 

changes will not affect the source term, 

containment isolation, or radiological release 

assumptions used in evaluating the 

radiological consequences of an accident 

previously evaluated. The proposed changes 

will not alter any assumptions or change any 

mitigation actions in the radiological 

consequence evaluations in the USAR. The 

applicable radiological dose acceptance 

criteria will continue to be met. 

Based on the above considerations, the 

proposed changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

There are no proposed design changes, nor 

are there any changes in the method by 

which any safety related plant SSC performs 

its specified safety function. The proposed 

changes will not affect the normal method of 

plant operation or change any operating 

parameters. No equipment performance 

requirements will be affected. The proposed 

changes will not alter any assumptions made 

in the safety analyses. 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures will be introduced as a result 

of this amendment. There will be no adverse 

effect or challenges imposed on any safety 

related system as a result of this amendment. 

The proposed amendment will not alter the 

design or performance of the [Analog Series] 

7300 Process Protection System, Nuclear 

Instrumentation System, or Solid State 

Protection System used in the plant 

protection systems. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

accident from any accident previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in the margin of 

safety? 

Response: No. 

There will be no effect on those plant 

systems necessary to assure the 

accomplishment of protection functions. 

There will be no impact on the overpower 

limit, departure from nucleate boiling ratio 

(DNBR) limits, heat flux hot channel factor 

(FQ), nuclear enthalpy rise hot channel factor 

(FAH), loss of coolant accident peak cladding 

temperature (LOCA PCT), peak local power 

density, or any other margin of safety. The 

applicable radiological dose consequence 

acceptance criteria for design-basis transients 

and accidents will continue to be met. 

The proposed changes do not eliminate 

any surveillances or alter the frequency of 

surveillances required by the Technical 

Specifications. None of the acceptance 

criteria for any accident analysis will be 

changed. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves NSHC. 

Attorney for licensee: Jay Silberg, Esq., 

Pillsbury Winthrop Shaw Pittman LLP, 

2300 N Street, NW., Washington, DC 

20037. 

NRC Branch Chief: Michael T. 

Markley. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: August 

18, 2008. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specification (TS) 3.5.2, 

‘‘ECCS [Emergency Core Cooling 

System]—Operating’’ requirements. The 

change is in accordance with Technical 

Specification Task Force (TSTF) TSTF– 

325–A, Revision 0, ‘‘ECCS Conditions 

and Required Actions with <100% 

Equivalent ECCS Flow.’’ 

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 (NSHC), 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 corrects the structure 

of the ACTIONS table to assure its correct 

application. There is no change or intent in 

the way the Conditions are actually applied. 

The literal interpretation of the existing 

Conditions structure could, under some 

circumstances, provide longer than intended 

Completion Times for restoration of 

OPERABILITY. Since the proposed change 

affects neither the Conditions intent nor its 

application, the proposed change will 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 change corrects the structure 

of the ACTIONS table to assure its correct 

application. The proposed change does not 

result in any physical alterations to the plant 

configuration, no new equipment additions, 

no equipment interface modifications, and no 

changes to any equipment function or the 

method of operating the equipment are being 

made. As the proposed change would not 

change the design, configuration or operation 

of the plant, no new or different kinds of 

accident modes are created. 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 the margin of 

safety? 

Response: No. 

The proposed change corrects the structure 

of the LCO [Limiting Condition for 

Operation] to assure its correct application. 

The proposed change is consistent with the 

requirements of the Technical Specifications. 

There is no change in intent or in the way 

the LCO is applied. Therefore, the proposed 

change does not involve a significant 

reduction in the margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves NSHC. 

Attorney for licensee: Jay Silberg, Esq., 

Pillsbury Winthrop Shaw Pittman LLP, 

2300 N Street, NW., Washington, DC 

20037. 

NRC Branch Chief: Michael T. 

Markley. 

mstockstill on PROD1PC66 with NOTICES

Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 58681 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

Systems (ADAMS) Public Electronic 

Reading Room on the internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

AmerGen Energy Company, LLC, 

Docket No. 50–461, Clinton Power 

Station, Unit No. 1, DeWitt County, 

Illinois 

Date of application for amendment: 

December 12, 2006, as supplemented by 

letters dated November 16, 2007, and 

May 16 and June 27, 2008. 

Brief description of amendment: The 

amendment would increase the interval 

between the local power range monitor 

(LPRM) calibrations from 1000 

megawatt-days/ton (MWD/T) to 2000 

MWD/T as required by the Clinton 

Power Station technical specification 

surveillance requirements 3.3.1.1.8 and 

SR 3.3.1.2.2. 

Date of issuance: September 12, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 181. 

Facility Operating License No. NPF– 

62: The amendment revised the 

Technical Specifications and License. 

Date of initial notice in Federal 

Register: May 22, 2007 (72 FR 28718). 

The November 16, 2007, and May 16 

and June 27, 2008, supplements, 

contained clarifying information and 

did not change the NRC staff’s initial 

proposed finding of no significant 

hazards consideration. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 12, 

2008. 

No significant hazards consideration 

comments received: No. 

Dominion Nuclear Connecticut, Inc., 

Docket Nos. 50–336 and 50–423 

Millstone Power Station, Unit Nos. 2 

and 3, New London County, 

Connecticut 

Date of application for amendments: 

July 13, 2007, as supplemented by 

letters dated December 7, 2007, March 

5, March 25, April 28, June 9, June 26, 

and July 28, 2008. 

Brief description of amendments: The 

amendment changed the Millstone 

Power Station, Unit Nos. 2 and 3 

Technical Specifications. This 

amendment established more effective 

and appropriate action, surveillance, 

and administrative requirements related 

to ensuring the habitability of the 

control room envelope in accordance 

with the Nuclear Regulatory 

Commission-approved Technical 

Specification Task Force (TSTF) 

Standard Technical Specification 

change traveler TSTF–448, Revision 3, 

‘‘Control Room Habitability.’’ 

Additionally, the amendment changed 

the ‘‘irradiated fuel movement’’ 

terminology and adopted ‘‘movement of 

recently irradiated fuel assemblies’’ 

terminology with TSTF–448, Revision 3. 

Date of issuance: September 18, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 180 days from the date of 

issuance. 

Amendment Nos.: 305 and 243. 

Renewed Facility Operating License 

No. DPR–65 and NPF–49: Amendments 

revised the License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: May 16, 2008 and July 1, 2008 

(73 FR 28534 and 73 FR 37506, 

respectively). The supplements dated 

June 9, June 26, and July 28, 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 amendment is contained in a 

Safety Evaluation dated September 18, 

2008. 

No significant hazards consideration 

comments received: No. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of application for amendment: 

July 30, 2007, as supplemented August 

28, 2008. 

Brief description of amendment: The 

amendment revises Technical 

Specifications 3.3.3.1, ‘‘Post Accident 

Monitoring (PAM) Instrumentation,’’ 

3.3.6.1, ‘‘Primary Containment Isolation 

Instrumentation,’’ 3.6.1.3, ‘‘Primary 

Containment Isolation Valves (PCIVs),’’ 

and 3.6.4.2, ‘‘Secondary Containment 

Isolation Valves (SCIVs).’’ The proposed 

changes adopt the following TS Task 

Force (TSTF) Travelers that have been 

previously approved by the NRC: TSTF– 

45–A, Revision 2, TSTF–46–A, Revision 

1, TSTF–207–A, Revision 5, TSTF–269– 

A, Revision 2, TSTF–295–A, Revision 0, 

TSTF–306–A, Revision 2, and TSTF– 

323–A, Revision 0. 

Date of issuance: September 15, 2008. 

Effective date: As of its date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 208. 

Facility Operating License No. NPF– 

21: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 28, 2007 (72 FR 

49573). 

The supplemental letter dated August 

28, 2008, provided additional 

information that clarified the 

mstockstill on PROD1PC66 with NOTICES

58682 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 

application, did not expand the scope of 

the application originally noticed, and 

did not change the staff’s original 

proposed no significant hazards 

consideration determination. The 

Commission’s related evaluation of the 

amendment is contained in a Safety 

Evaluation dated September 15, 2008, 

No significant hazards consideration 

comments received: No. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of application for amendment: 

May 7, 2008 

Brief description of amendment: The 

amendment revises Technical 

Specification Limiting Condition for 

Operation 3.10.1, and approves the 

associated Bases, to expand its scope to 

include provisions for temperature 

excursions greater than 200 degrees 

Fahrenheit as a consequence of 

inservice leak and hydrostatic testing, 

and as a consequence of scram time 

testing initiated in conjunction with an 

inservice leak or hydrostatic test, while 

considering operational conditions to be 

in Mode 4. 

Date of issuance: September 16, 2008. 

Effective date: As of its date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment No.: 209. 

Facility Operating License No. NPF– 

21: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 15, 2008 (73 FR 40630). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated September 16, 2008. 

No significant hazards consideration 

comments received: No. 

FPL Energy Duane Arnold, LLC, Docket 

No. 50–331, Duane Arnold Energy 

Center, Linn County, Iowa 

Date of application for amendment: 

February 19, 2008. 

Brief description of amendment: The 

amendment revises the Technical 

Specification Actions for the Emergency 

Diesel Generators (EDG) to remove the 

conditional surveillance requirement to 

test the alternate EDG whenever one 

EDG is taken out of service for preplanned preventive maintenance and 

testing. 

Date of issuance: September 9, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 270. 

Facility Operating License No. DPR– 

49: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: June 13, 2008 (73 FR 33853). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated September 9, 2008. 

No significant hazards consideration 

comments received: No. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Steam Electric Station, 

Unit Nos. 1 and 2, Somervell County, 

Texas 

Date of amendment request: 

November 29, 2007. 

Brief description of amendments: The 

amendments revised the Technical 

Specification (TS) 3.6.7, ‘‘Spray 

Additive System,’’ to allow 

modifications to the facility potentially 

required to address U.S. Nuclear 

Regulatory Commission (NRC) Generic 

Letter 2004–02, ‘‘Potential Impact of 

Debris Blockage on Emergency 

Recirculation during Design Basis 

Accident at Pressurized-Water Reactors’’ 

and authorized changes to TS 3.6.7 to 

remove the current surveillances for 

sodium hydroxide and insert a 

surveillance to ensure equilibrium sump 

pH is greater than or equal to 7.1. 

Date of issuance: September 12, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: Unit 1–147, Unit 

2–147. 

Facility Operating License Nos. NPF– 

87 and NPF–89: The amendments 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in Federal 

Register: December 31, 2007 (72 FR 

74360). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

September 12, 2008. 

No significant hazards consideration 

comments received: No. 

Nine Mile Point Nuclear Station, LLC, 

Docket No. 50–220, Nine Mile Point 

Nuclear Station, Unit No. 1, Oswego 

County, New York 

Date of application for amendment: 

July 23, 2007, as supplemented by letter 

dated January 24, 2008. 

Brief description of amendment: The 

amendment revises Technical 

Specification (TS) Section 3.1.1, 

‘‘Control Rod System,’’ to incorporate a 

provision that should the rod worth 

minimizer (RWM) become inoperable 

before a reactor startup is commenced or 

before the first 12 control rods have 

been withdrawn, startup will be allowed 

to continue. This provision will rely on 

the RWM function being performed 

manually and will require a double 

check of compliance with the control 

rod program by a second licensed 

operator or other qualified member of 

the technical staff. The use of this 

allowance will be limited to one startup 

in the last calendar year. 

Date of issuance: July 29, 2008. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment No.: 196. 

Renewed Facility Operating License 

No. DPR–63: Amendment revised the 

License and TSs. 

Date of initial notice in Federal 

Register: September 11, 2007 (72 FR 

51863). 

The supplemental letter dated January 

24, 2008, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the Nuclear 

Regulatory Commission staff’s initial 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 29, 2008. 

No significant hazards consideration 

comments received: No. 

Northern States Power Company, 

Docket No. 50–263, Monticello Nuclear 

Generating Plant, Wright County, 

Minnesota 

Date of application for amendment: 

April 16, 2008, as supplemented by 

letter dated August 6, 2008. 

Brief description of amendment: The 

amendment conforms Renewed Facility 

Operating License No. DPR–22 to reflect 

the fact that Northern States Power 

Company holds the operating authority 

of the unit as of the date of this 

amendment. This license transfer was 

previously approved by an Order dated 

September 15, 2008. 

Date of issuance: September 22, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 1 year. 

Amendment No.: 156. 

Facility Operating License No. DPR– 

22: Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: June 5, 2008 (73 FR 32057). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated September 15, 2008. 

No significant hazards consideration 

comments received: As provided in 10 

CFR 2.1315, no public comments with 

respect to significant hazards 

considerations were solicited. 

mstockstill on PROD1PC66 with NOTICES

Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 58683 

Northern States Power Company, 

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

Island Nuclear Generating Plant, Units 

1 and 2, Goodhue County, Minnesota 

Date of application for amendments: 

April 16, 2008, as supplemented by 

letter dated August 6, 2008. 

Brief description of amendments: The 

amendments conform the Technical 

Specifications and Facility Operating 

License Nos. DPR–42 and DPR–60 to 

reflect the fact that Northern States 

Power Company holds the operating 

authority of the units as of the date of 

these amendments. This license transfer 

was previously approved by an Order 

dated September 15, 2008. 

Date of issuance: September 22, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 1 year. 

Amendment Nos.: 188 (for Unit 1) and 

177 (for Unit 2). 

Facility Operating License Nos. DPR– 

42 and DPR–60: Amendments revised 

the Facility Operating Licenses and the 

Technical Specifications. 

Date of initial notice in Federal 

Register: June 5, 2008 (73 FR 32055). 

The Commission’s related evaluation of 

the amendments is contained in a Safety 

Evaluation dated September 15, 2008. 

No significant hazards consideration 

comments received: As provided in 10 

CFR 2.1315, no public comments with 

respect to significant hazards 

considerations were solicited. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of application for amendments: 

March 11, 2008, as supplemented on 

June 17, and July 23, 2008. 

Brief description of amendments: The 

amendments revise the Technical 

Specification (TS) requirements for fuel 

decay time prior to commencing 

movement of irradiated fuel in the 

reactor pressure vessel. 

Date of issuance: September 24, 2008. 

Effective date: As of the date of 

issuance, to be implemented within 30 

days. 

Amendment Nos.: 289 and 273. 

Facility Operating License Nos. DPR– 

70 and DPR–75: The amendments revise 

the TSs and the license. 

Date of initial notice in Federal 

Register: July 15, 2008 (73 FR 40631). 

The letters dated June 17, and July 23, 

2008, provided clarifying information 

that did not change the initial proposed 

no significant hazards consideration 

determination or expand the application 

beyond the scope of the original Federal 

Register notice. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 24, 

2008. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Georgia Power Company, 

Oglethorpe Power Corporation, 

Municipal Electric Authority of 

Georgia, City of Dalton, Georgia, Docket 

Nos. 50–321 and 50–366, Edwin I. 

Hatch Nuclear Plant, Units 1 and 2, 

Appling County, Georgia 

Date of application for amendments: 

April 29, 2008. 

Brief description of amendments: The 

amendments revise Technical 

Specification Figure 3.1.7–1, ‘‘Sodium 

Penataborate Solution Volume Versus 

Concentration Requirements,’’ by 

implementing an editorial change to 

improve the readability of the figure. 

Date of issuance: September 23, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: Unit 1–257, Unit 

2–201. 

Renewed Facility Operating License 

Nos. DPR–57 and NPF–5: Amendments 

revised the licenses and the technical 

specifications. 

Date of initial notice in Federal 

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

The Commission’s related evaluation of 

the amendments is contained in a Safety 

Evaluation dated September 23, 2008. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–424 and 50–425, 

Vogtle Electric Generating Plant, Units 

1 and 2, Burke County, Georgia 

Date of application for amendments: 

June 27, 2008. 

Brief description of amendments: The 

amendments revised the combined 

Vogtle Electric Generating Plant, Units 1 

and 2 Technical Specifications (TS) 

5.5.9, ‘‘Steam Generator (SG) Program’’ 

and TS 5.6.10, ‘‘Steam Generator Tube 

Inspection Report,’’ to incorporate a 

one-cycle interim alternate repair 

criterion in the provisions for SG tube 

repair criteria for VEGP Unit 2 during 

refueling outage 2R13 and the 

subsequent operating cycle. 

Date of issuance: September 16, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: Unit 1–152, Unit 

2–133. 

Facility Operating License Nos. NPF– 

68 and NPF–81: Amendments revised 

the technical specifications. 

Date of initial notice in Federal 

Register: July 14, 2008 (73 FR 40394). 

The Commission’s related evaluation of 

the amendments is contained in a Safety 

Evaluation dated September 16, 2008. 

No significant hazards consideration 

comments received: No. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project, Units 1 and 2, Matagorda 

County, Texas 

Date of amendment request: August 

27, 2007, as supplemented by letters 

dated March 27 and September 5, 2008. 

Brief description of amendments: The 

amendments revised the South Texas 

Project, Units 1 and 2 fire protection 

program to allow the performance of 

operator manual actions to achieve and 

maintain safe shutdown in the event of 

a fire, in lieu of meeting circuit 

separation requirements specified in 

Title 10 of the Code of Federal 

Regulations, Part 50, Appendix R, 

Section III.G.2, for a fire in Fire Area 32 

located in the Mechanical/Electrical 

Auxiliary Building. License Condition 

2.E of the operating licenses is revised. 

Date of issuance: September 16, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment Nos.: Unit 1–186, Unit 

2–173. 

Facility Operating License Nos. NPF– 

76 and NPF–80: The amendments 

revised the Facility Operating Licenses. 

Date of initial notice in Federal 

Register: November 20, 2007 (72 FR 

65373). The supplemental letters dated 

March 27 and September 5, 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 September 16, 

2008. 

No significant hazards consideration 

comments received: No. 

Virginia Electric and Power Company, 

et al., Docket Nos. 50–280 and 50–281, 

Surry Power Station, Units 1 and 2, 

Surry County, Virginia 

Date of application for amendments: 

September 19, 2007, as supplemented 

on April 11, 2008. 

Brief Description of amendments: 

These amendments revised various 

Technical Specification (TS) setting 

mstockstill on PROD1PC66 with NOTICES

58684 Federal Register / Vol. 73, No. 195 / Tuesday, October 7, 2008 / Notices 

limits and the overtemperature DT/ 

overpower DT time constants in TS 2.3 

and TS 3.7. The methodology for 

determining the revised setting limits 

and time constants is in agreement with 

methods 1 and 2 in ‘‘The 

Instrumentation, Systems, and 

Automation Society (ISA),’’ Standard 

ISA–R67.04, Part II, ‘‘Methodologies for 

the Determination of Setpoints for 

Nuclear Safety-Related 

Instrumentation.’’ 

Date of issuance: September 17, 2008. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 261 and 261. 

Renewed Facility Operating License 

Nos. DPR–32 and DPR–37: Amendments 

changed the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: October 23, 2007 (72 FR 

60036). The supplement dated April 11, 

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 published in the Federal 

Register on October 23, 2007. The 

Commission’s related evaluation of the 

amendments is contained in a Safety 

Evaluation dated September 17, 2008. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 29th day 

of September 2008. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E8–23342 Filed 10–6–08; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Sunshine Federal Register Notice 

AGENCY HOLDING THE MEETINGS: Nuclear 

Regulatory Commission. 

DATES: Weeks of October 6, 13, 20, 27, 

November 3, 10, 2008. 

PLACE: Commissioners’ Conference 

Room, 11555 Rockville Pike, Rockville, 

Maryland. 

STATUS: Public and Closed. 

Week of October 6, 2008 

Monday, October 6, 2008 

12:55 p.m. Affirmation Session (Public 

Meeting) (Tentative) 

a. Oyster Creek, Indian Point, Pilgrim, 

and Vermont Yankee License 

Renewals, Docket Nos. 50–219–LR, 

50–247–LR, 50–286–LR, 50–293– 

LR, 50–271–LR, Petition to Suspend 

Proceedings (Tentative). 

b. Pacific Gas and Electric Co. (Diablo 

Canyon ISFSI), Docket No. 72–26– 

ISFSI, Decision on the Merits of San 

Luis Obispo Mothers for Peace’s 

Contention 2 (Tentative). 

c. EnergySolutions (Radioactive Waste 

Import/Export)—EnergySolutions’ 

Applications for Low-Level 

Radioactive Waste Import and 

Export Licenses (Tentative). 

1 p.m. Discussion of Security Issues 

(Closed—Ex. 1 and 3). 

Week of October 13, 2008—Tentative 

There are no meetings scheduled for 

the week of October 13, 2008. 

Week of October 20, 2008—Tentative 

Wednesday, October 22, 2008 

9:30 a.m. Briefing on New Reactor 

Issues—Construction Readiness, 

Part 1 (Public Meeting) (Contact: 

Roger Rihm, 301 415–7807). 

1:30 p.m. Briefing on New Reactor 

Issues—Construction Readiness, 

Part 2 (Public Meeting) (Contact: 

Roger Rihm, 301 415–7807). 

Both parts of this meeting will be 

Webcast live at the Web address— 

http://www.nrc.gov. 

Week of October 27, 2008—Tentative 

There are no meetings scheduled for 

the week of October 27, 2008. 

Week of November 3, 2008—Tentative 

Thursday, November 6, 2008 

1:30 p.m. Briefing on NRC 

International Activities (Public 

Meeting) (Contact: Karen 

Henderson, 301 415–0202). 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov. 

Friday, November 7, 2008 

2 p.m. Meeting with Advisory 

Committee on Reactor Safeguards 

(Public Meeting) (Contact: Tanny 

Santos, 301 415–7270). 

This meeting will be Webcast live at 

the Web address—http://www.nrc.gov. 

Week of November 10, 2008—Tentative 

There are no meetings scheduled for 

the week of November 10, 2008. 

*The schedule for Commission 

meetings is subject to change on short 

notice. To verify the status of meetings, 

call (recording)—(301) 415–1292. 

Contact person for more information: 

Michelle Schroll, (301) 415–1662. 

The NRC Commission Meeting 

Schedule can be found on the Internet 

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

The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g. 

braille, large print), please notify the 

NRC’s Disability Program Coordinator, 

Rohn Brown, at 301–492–2279, TDD: 

301–415–2100, or by e-mail at 

rohn.brown@nrc.gov. Determinations on 

requests for reasonable accommodation 

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

This notice is distributed by mail to 

several hundred subscribers; if you no 

longer wish to receive it, or would like 

to be added to the distribution, please 

contact the Office of the Secretary, 

Washington, DC 20555 (301–415–1969). 

In addition, distribution of this meeting 

notice over the Internet system is 

available. If you are interested in 

receiving this Commission meeting 

schedule electronically, please send an 

electronic message to 

darlene.wright@nrc.gov. 

Dated: October 2, 2008. 

R. Michelle Schroll, 

Office of the Secretary. 

[FR Doc. E8–23844 Filed 10–3–08; 4:15 pm] 

BILLING CODE 7590–01–P 

OFFICE OF PERSONNEL 

MANAGEMENT 

[OMB Control No. 3206–0206; Form RI 25– 

37] 

Submission for OMB Review; Request 

for Comments on a Revised 

Information Collection 

AGENCY: Office of Personnel 

Management. 

ACTION: Notice. 

SUMMARY: In accordance with the 

Paperwork Reduction Act of 1995 (Pub. 

L. 104–13, May 22, 1995), this notice 

announces that the Office of Personnel 

Management (OPM) has submitted to 

the Office of Management and Budget 

(OMB) a request for review of a revised 

information collection. This information 

collection, ‘‘Evidence to Prove 

Dependency of a Child’’ (OMB Control 

No. 3206–0206; form RI 25–37), is 

designed to collect sufficient 

information for OPM to determine 

whether the surviving child of a 

deceased federal employee is eligible to 

receive benefits as a dependent child. 

Approximately 250 forms are 

completed annually. We estimate it 

takes approximately 60 minutes to 

assemble the needed documentation. 

mstockstill on PROD1PC66 with NOTICES