Federal Register / Vol. 74, No. 75 / Tuesday, April 21, 2009 / Notices 18251 

FOR FURTHER INFORMATION CONTACT: 

Mary Rupp, Secretary of the Board, 

Telephone: 703-518-6304. 

Mary Rupp, 

Board Secretary. 

[FR Doc. E9–9268 Filed 4–17–09; 4:15 pm] 

BILLING CODE P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2009–0170] 

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 March 26, 

2009, to April 8, 2009. The last biweekly 

notice was published on April 7, 2009 

(74 FR 15765). 

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 and 

Directives Branch, TWB–05–B01M, 

Division of Administrative Services, 

Office of Administration, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001, and should cite the 

publication date and page number of 

this Federal Register notice. Copies of 

written comments received may be 

examined at the Commission’s Public 

Document Room (PDR), located at One 

White Flint North, Public File Area 

O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. 

Within 60 days after the date of 

publication of this notice, any person(s) 

whose interest may be affected by this 

action may file a request for a hearing 

and a petition to intervene with respect 

to issuance of the amendment to the 

subject facility operating license. 

Requests for a hearing and a petition for 

leave to intervene shall be filed in 

accordance with the Commission’s 

‘‘Rules of Practice for Domestic 

Licensing Proceedings’’ in 10 CFR Part 

2. Interested person(s) should consult a 

current copy of 10 CFR 2.309, which is 

available at the Commission’s PDR, 

located at One White Flint North, Public 

File Area 01F21, 11555 Rockville Pike 

(first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/ 

requestor to relief. A petitioner/ 

requestor who fails to satisfy these 

requirements with respect to at least one 

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18252 Federal Register / Vol. 74, No. 75 / Tuesday, April 21, 2009 / Notices 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

All documents filed in NRC 

adjudicatory proceedings, including a 

request for hearing, a petition for leave 

to intervene, any motion or other 

document filed in the proceeding prior 

to the submission of a request for 

hearing or petition to intervene, and 

documents filed by interested 

governmental entities participating 

under 10 CFR 2.315(c), must be filed in 

accordance with the NRC E–Filing rule, 

which the NRC promulgated in August 

28, 2007 (72 FR 49139). The E–Filing 

process requires participants to submit 

and serve all adjudicatory documents 

over the Internet or in some cases to 

mail copies on electronic storage media. 

Participants may not submit paper 

copies of their filings unless they seek 

a waiver in accordance with the 

procedures described below. 

To comply with the procedural 

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

days prior to the filing deadline, the 

petitioner/requestor must contact the 

Office of the Secretary by e-mail at 

hearingdocket@nrc.gov, or by calling 

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

ID certificate, which allows the 

participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and/or (2) creation of an 

electronic docket for the proceeding 

(even in instances in which the 

petitioner/requestor (or its counsel or 

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

petitioner/requestor will need to 

download the Workplace Forms 

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 

electronic filing Help Desk, which is 

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

Eastern Time, Monday through Friday, 

excluding government holidays. The 

help electronic filing Help Desk can be 

contacted by telephone at 1–866–672– 

7640 or by e-mail at 

MSHD.Resource@nrc.gov. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file a 

motion, in accordance with 10 CFR 

2.302(g), with their initial paper filing 

requesting authorization to continue to 

submit documents in paper format. 

Such filings must be submitted by: (1) 

First class mail addressed to the Office 

of the Secretary of the Commission, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, Attention: 

Rulemaking and Adjudications Staff; or 

(2) courier, express mail, or expedited 

delivery service to the Office of the 

Secretary, Sixteenth Floor, One White 

Flint North, 11555 Rockville, Pike, 

Rockville, Maryland, 20852, Attention: 

Rulemaking and Adjudications Staff. 

Participants filing a document in this 

manner are responsible for serving the 

document on all other participants. 

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

the mail, or by courier, express mail, or 

expedited delivery service upon 

depositing the document with the 

provider of the service. 

Non-timely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission, the presiding officer, or 

the Atomic Safety and Licensing Board 

that the petition and/or request should 

be granted and/or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

CFR 2.309(c)(1)(i)–(viii). 

Documents submitted in adjudicatory 

proceedings will appear in NRC’s 

electronic hearing docket which is 

available to the public at http:// 

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

unless excluded pursuant to an order of 

the Commission, an Atomic Safety and 

Licensing Board, or a Presiding Officer. 

Participants are requested not to include 

personal privacy information, such as 

social security numbers, home 

addresses, or home phone numbers in 

their filings, unless an NRC regulation 

or other law requires submission of such 

information. With respect to 

copyrighted works, except for limited 

excerpts that serve the purpose of the 

adjudicatory filings and would 

constitute a Fair Use application, 

participants are requested not to include 

copyrighted materials in their 

submission. 

For further details with respect to this 

amendment action, see the application 

for amendment which is available for 

public inspection at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the ADAMS Public 

Electronic Reading Room on the Internet 

at the NRC Web site, http:// 

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

you do not have access to ADAMS or if 

there are problems in accessing the 

documents located in ADAMS, contact 

the PDR Reference staff at 1 (800) 397– 

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Federal Register / Vol. 74, No. 75 / Tuesday, April 21, 2009 / Notices 18253 

4209, (301) 415–4737 or by e-mail to 

pdr@nrc.gov. 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–369 and 50–370, McGuire 

Nuclear Station, Units 1 and 2, 

Mecklenburg County, North Carolina 

Date of amendment request: August 

21, 2008. 

Description of amendment request: 

The proposed amendments would 

revise the proposed license amendment 

implements Technical Specification 

Task Force (TSTF) Changes Travelers 

TSTF–479, Revision 0, ‘‘Changes to 

Reflect Revision of [Title 10 of the Code 

of Federal Regulations] 10 CFR 50.55a’’ 

and TSTF–497, Revision 0, ‘‘Limit 

Inservice Testing [IST] Program SR 3.0.2 

Application to Frequencies of 2 Years or 

Less’’. TSTF–479 and TSTF–497 revise 

the technical specification 

Administrative Controls section 

pertaining to requirements for the IST 

Program, consistent with the 

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

pumps and valves which are classified 

as American Society of Mechanical 

Engineers (ASME) Code Class 1, Class 2, 

and Class 3. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises TS [Technical 

Specification] 5.5.8, ‘‘Inservice Testing 

Program,’’ for consistency with the 

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

the inservice testing of pumps and valves 

which are classified as ASME Code Class 1, 

Class 2, and Class 3. The proposed change 

incorporates revisions to the ASME 

[American Society of Mechanical Engineers] 

Code as identified in the TSTFs [Technical 

Specification Task Force] referenced above. 

The proposed change does not impact any 

accident initiators or analyzed events or 

assumed mitigation of accident or transient 

events. The proposed change does not 

involve the addition or removal of any 

equipment, or any design changes to the 

facility. Additionally, there is no change in 

the types or increases in the amounts of any 

effluent that may be released offsite and there 

is no increase in individual or cumulative 

occupational exposure. 

Therefore, this 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 revises TS 5.5.8, 

‘‘Inservice Testing Program,’’ for consistency 

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

regarding the inservice testing of pumps and 

valves which are classified as ASME Code 

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

change incorporates revisions to the ASME 

Code as identified in the TSTFs referenced 

above. The proposed change does not involve 

a modification to the physical configuration 

of the plant nor does it involve a change in 

the methods governing normal plant 

operation. The proposed change will not 

impose any new or different requirements or 

introduce a new accident initiator, accident 

precursor, or malfunction mechanism. 

Additionally, there is no change in the types 

or increases in the amounts of any effluent 

that may be released offsite and there is no 

increase in individual or cumulative 

occupational exposure. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No 

The proposed change revises TS 5.5.8, 

‘‘Inservice Testing Program,’’ for consistency 

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

regarding the inservice testing of pumps and 

valves which are classified as ASME Code 

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

change does not involve a modification to the 

physical configuration of the plant nor does 

it change the methods governingnormal plant 

operation. The proposed change incorporates 

revisions to the ASME Code as identified in 

the TSTFs referenced above. 

The safety function of the affected pumps 

and valves will be maintained. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ms. Lisa F. 

Vaughn, Associate General Counsel and 

Managing Attorney, Duke Energy 

Carolinas, LLC, 526 South Church 

Street, EC07H, Charlotte, NC 28202. 

NRC Branch Chief: Melanie Wong. 

Entergy Operations, Inc., Docket No. 50– 

313, Arkansas Nuclear One, Unit No. 1, 

Pope County, Arkansas 

Date of amendment request: February 

16, 2009. 

Description of amendment request: 

The Arkansas Nuclear One, Unit No. 1 

(ANO–1) Technical Specification (TS) 

5.5.16, ‘‘Reactor Building Leakage Rate 

Testing Program,’’ contains reactor 

building leak rate criteria for overall 

Type A, B, and C testing. However, TS 

5.5.16 does not specify criteria for Type 

B air lock leakage testing. Entergy 

Operations, Inc., proposes to modify TS 

5.5.16 to add criteria for overall air lock 

leakage testing and to adopt a low 

pressure test method relevant to the air 

lock door seals. This change is 

consistent with NUREG 1430, Revision 

3.1, ‘‘Standard Technical Specifications 

(STS) for Babcock & Wilcox Plants.’’ 

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 reactor building air locks are passive 

components integral to the reactor building 

structure and are not associated with 

accident initiators. Each air lock door is rated 

for and tested to the maximum calculated 

post-accident pressure of the reactor 

building. The air lock door seal pressure test 

is performed any time the air lock is used for 

reactor building access during modes of 

operation when reactor building integrity is 

required and prior to establishing reactor 

building integrity. The door seal test is 

intended to be a gross test to verify that the 

door seals were not damaged during the 

opening and closing cycle(s). This test does 

not replace the required overall barrel 

leakage test. Based on information provided 

by the air lock vendor, a test pressure of 10 

psig [pounds per square inch gauge] is 

conservatively sufficient to perform this gross 

seal verification. This new acceptable leakage 

rate and test criteria are consistent with 

NUREG 1430, Rev. 3.1, Standard Technical 

Specifications for Babcock & Wilcox Plants 

(STS) and are applicable to ANO–1. While 

new to the TSs, the ANO–1 program for 

ensuring compliance with 10 CFR 50, 

Appendix J has verified leakage within the 

proposed limiting values. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

No physical changes to the facility are 

initiated by the proposed change. In addition, 

the proposed change has no affect on plant 

configuration, or method of operation of 

plant structures, systems, or components. 

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 does not increase the 

allowable overall air lock leakage rate, nor 

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18254 Federal Register / Vol. 74, No. 75 / Tuesday, April 21, 2009 / Notices 

affect the acceptance criteria of the overall 

integrated containment leakage rate as 

currently tested to in accordance with the 

ANO–1 containment leakage rate test 

program. All of the changes are bounded by 

existing analyses for all evaluated accidents 

and do not create any situations that alter the 

assumptions used in these analyses. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Terence A. 

Burke, Associate General Council— 

Nuclear Entergy Services, Inc., 1340 

Echelon Parkway, Jackson, Mississippi 

39213. 

NRC Branch Chief: Michael T. 

Markley. 

Entergy Operations, Inc., Docket No. 50– 

313, Arkansas Nuclear One, Unit No. 1, 

Pope County, Arkansas 

Date of amendment request: March 

10, 2009. 

Description of amendment request: 

The proposed amendment consists of 

changes to Technical Specification (TS) 

3.4.9, ‘‘Pressurizer,’’ which contains a 

maximum and minimum level for the 

pressurizer. The licensee proposes to 

delete the minimum level requirement. 

This change is consistent with NUREG 

1430, Rev. 3.1, ‘‘Standard Technical 

Specifications [STS] for Babcock and 

Wilcox Plants.’’ 

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 minimum Pressure level limit 

currently specified in the TSs does not act to 

ensure specified fuel design limits are 

protected. Accident and transient analyses 

assume lowering or a loss of Pressurizer 

level. Safety systems are designed and 

maintained available to mitigate the 

consequences of an accident or transient that 

may involve a loss of Reactor Coolant System 

(RCS) inventory. None of these systems rely 

upon a predetermined minimum Pressurizer 

level in order to perform their intended 

function. Furthermore, the minimum 

Pressure level limit is unrelated to any 

anticipated accident initiator. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

No physical changes to the facility are 

initiated by the proposed change. In addition, 

the proposed change has no affect on plant 

configuration, or method of operation of 

plant structures, systems, or components. 

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. 

Installed automatic control systems will 

continue to maintain Pressurizer level at a 

predetermined setpoint and are independent 

of a prescribed minimum TS level limit. The 

deletion of the current TS limit has no 

impact on guidance or operational response 

to pressurizer level deviations. Furthermore, 

the minimum Pressure level limit is not an 

assumed value for accident prevention or 

mitigation in the [Arkansas Nuclear One, 

Unit 1] [Safety Analysis Report]. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Terence A. 

Burke, Associate General Council— 

Nuclear Entergy Services, Inc., 1340 

Echelon Parkway, Jackson, Mississippi 

39213. 

NRC Branch Chief: Michael T. 

Markley. 

Exelon Generation Company, LLC, 

Docket Nos. 50–352 and 50–353, 

Limerick Generating Station, Units 1 

and 2, Montgomery County, 

Pennsylvania 

Date of amendment request: February 

25, 2009. 

Description of amendment request: 

The proposed change removes the 

reactor coolant system (RCS) structural 

integrity requirements contained in 

Technical Specification (TS) 3/4.4.8, 

which specifies requirements relating to 

the structural integrity of American 

Society of Mechanical Engineers 

(ASME) Code Class 1, 2 and 3 

components. This specification is 

redundant to the requirements 

contained within Title 10 of the Code of 

Federal Regulations (10 CFR) section 

50.55a, ‘‘Codes and standards.’’ With 

this proposed change, RCS pressure 

boundary structural integrity will 

continue to be maintained by 

compliance with 10 CFR 50.55a, as 

implemented through the Limerick 

Generating Station, Units 1 and 2, 

Inservice Inspection Program. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below, with NRC edits in brackets: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change to remove the RCS 

structural integrity controls from the TSs 

does not impact any mitigation equipment or 

the ability of the RCS pressure boundary to 

fulfill any required safety function. Since no 

accident mitigation [equipment] or initiators 

are impacted by this change, no design basis 

accidents are affected. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of any 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 will not alter the 

plant configuration or change the manner in 

which the plant is operated. No new failure 

modes are being introduced by the proposed 

change. 

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. 

Removal of TS 3/4.4.8 from the TSs does 

not reduce the controls that are required to 

maintain the RCS pressure boundary for 

ASME Code Class 1, 2, or 3 components. 

No equipment or RCS safety margins are 

impacted due to the proposed change. 

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: J. Bradley 

Fewell, Esquire, Associate General 

Counsel, Exelon Generation Company, 

LLC, 4300 Winfield Road, Warrenville, 

IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

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Federal Register / Vol. 74, No. 75 / Tuesday, April 21, 2009 / Notices 18255 

Florida Power and Light Company, 

Docket Nos. 50–250 and 50–251, Turkey 

Point Plant, Units 3 and 4, Miami-Dade 

County, Florida 

Date of amendment request: February 

16, 2009. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specifications (TSs) by 

removing the structural integrity 

requirements contained in TS 3/4.4.10 

and the associated TS bases from the 

TSs. Removal of TS 3/4.4.10 is 

consistent with NUREG–1431, Revision 

3.0, ‘‘Standard Technical Specifications 

Westinghouse Plants,’’ in that it does 

not meet the criteria of Title 10 of the 

Code of Federal Regulations (10 CFR), 

Part 50, Section 50.36, ‘‘Technical 

Specifications,’’ for inclusion in the 

TSs. The proposed amendment would 

also relocate the reactor coolant pump 

(RCP) flywheel inspection requirements 

in Surveillance Requirement (SR) 4.4.10 

to SR 4.0.5, and would revise the RCP 

flywheel inspection interval from 10 

years to 20 years. The RCP flywheel 

inspection interval change is consistent 

with Nuclear Regulatory Commission 

approved Industry/Technical 

Specification Task Force (TSTF) 

Standard Technical Specification 

Change Traveler, TSTF–421, ‘‘Revision 

to RCP Flywheel Inspection Program 

(WCAP–15666).’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change to remove structural 

integrity controls from the TSs does not 

impact any mitigation equipment or the 

ability of the RCS [reactor coolant system] 

pressure boundary to fulfill any required 

safety function. The proposed change will 

continue to ensure the requirements of 10 

CFR 50.55a [‘‘Codes and standards’’] are 

maintained as specified in TS 4.0.5. Since no 

accident mitigation or initiators are impacted 

by this change, no design basis accidents are 

affected. 

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 previously evaluated? 

Response: No. 

The proposed change will not alter the 

plant configuration or change the manner in 

which the plant is operated. Structural 

integrity will continue to be maintained as 

required by 10 CFR 50.55a and specified in 

TS 4.0.5. No new failure modes are being 

introduced by the proposed change. 

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 the margin of safety? 

Response: No. 

Removal of TS 3/4.4.10 from the TSs does 

not reduce the controls that are required to 

maintain the structural integrity of ASME 

[American Society of Mechanical Engineers] 

Code Class 1, 2, or 3 components. No safety 

margins are impacted due to the proposed 

change. 

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 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: M.S. Ross, 

Attorney, Florida Power & Light, P.O. 

Box 14000, Juno Beach, Florida 33408– 

0420. 

NRC Branch Chief: Thomas H. Boyce. 

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: February 

11, 2009. 

Description of amendment request: 

The proposed amendment would delete 

those portions of the Technical 

Specifications (TSs) superseded by 10 

CFR Part 26, Subpart I. The proposed 

change is consistent with Nuclear 

Regulatory Commission (NRC)-approved 

Revision 0 to TS Task Force (TSTF) 

Change Traveler, TSTF–511–A, 

‘‘Eliminate Working Hour Restrictions 

from TS 5.2.2 to Support Compliance 

with 10 CFR Part 26.’’ The availability 

of the TS improvement was announced 

in the Federal Register (FR) on 

December 30, 2008 (73 FR 79923) as 

part of the consolidated line item 

improvement process. The licensee 

concluded that the no significant 

hazards consideration determination as 

presented in the FR notice is applicable 

to NMP 1 and 2. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change removes Technical 

Specification restrictions on working hours 

for personnel who perform safety related 

functions. The Technical Specification 

restrictions are superseded by the worker 

fatigue requirements in 10 CFR Part 26. 

Removal of the Technical Specification 

requirements will be performed concurrently 

with the implementation of the 10 CFR Part 

26, Subpart I, requirements. The proposed 

change does not impact the physical 

configuration or function of plant structures, 

systems, or components (SSCs) or the manner 

in which the SSCs are operated, maintained, 

modified, tested, and inspected. Worker 

fatigue is not an initiator of any accident 

previously evaluated. Worker fatigue is not 

an assumption in the consequence mitigation 

of any accident previously evaluated. 

Therefore, it is concluded that 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 removes Technical 

Specification restrictions on working hours 

for personnel who perform safety related 

functions. The Technical Specification 

restrictions are superseded by the worker 

fatigue requirements in 10 CFR Part 26. 

Working hours will continue to be controlled 

in accordance with NRC requirements. The 

new rule allows for deviations from controls 

to mitigate or prevent a condition adverse to 

safety or as necessary to maintain the 

security of the facility. This ensures that the 

new rule will not unnecessarily restrict 

working hours and thereby create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. The proposed change does not 

alter the plant configuration, require new 

plant equipment to be installed, alter 

accident analysis assumptions, add any 

initiators, or affect the function of plant 

systems or the manner in which systems are 

operated, maintained, modified, tested, or 

inspected. 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 

The proposed change removes Technical 

Specification restrictions on working hours 

for personnel who perform safety related 

functions. The Technical Specification 

restrictions are superseded by the worker 

fatigue requirements in 10 CFR Part 26. 

Working hours will continue to be controlled 

in accordance with NRC requirements. The 

proposed change does not involve any 

physical changes to the plants or alter the 

manner in which plant systems are operated, 

maintained, modified, tested, and inspected. 

The proposed change does not alter the 

manner in which safety limits, limiting safety 

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18256 Federal Register / Vol. 74, No. 75 / Tuesday, April 21, 2009 / Notices 

system settings or limiting conditions for 

operation are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed change will not result 

in plant operation in a configuration outside 

the design basis. The proposed change does 

not adversely affect systems that respond to 

safely shutdown the plants and to maintain 

the plants in a safe shutdown condition. 

Removal of plant-specific Technical 

Specification administrative requirements 

will not reduce a margin of safety because the 

requirements in 10 CFR Part 26 are adequate 

to ensure that worker fatigue is managed. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: 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: February 

20, 2009. 

Description of amendment request: 

The proposed amendment would 

modify Technical Specifications (TS) 

requirements related to control room 

envelope habitability in TS 3.7.3, ‘‘Plant 

Systems Control Room Emergency 

Outside Air Supply (CREOAS) System,’’ 

and TS Section 5.5, ‘‘Administrative 

Controls Programs and Manuals.’’ 

The NRC staff issued a notice of 

opportunity for comment in the Federal 

Register on October 17, 2006 (71 FR 

61075), on possible amendments to 

revise the plant specific TS, to 

strengthen TS requirements regarding 

control room envelope (CRE) 

habitability by changing the action and 

surveillance requirements associated 

with the limiting condition for 

operation operability requirements for 

the CRE emergency ventilation system. 

A new TS administrative controls 

program on CRE habitability is being 

added, including a model safety 

evaluation and model no significant 

hazards consideration determination, 

using the consolidated line-item 

improvement process. The NRC staff 

subsequently issued a notice of 

availability of the models for referencing 

in license amendment applications in 

the Federal Register on January 17, 

2007 (72 FR 2022). The licensee 

affirmed the applicability of the model 

NSHC determination in its application 

dated February 20, 2009. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below: 

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change does not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, or 

configuration of the facility. The proposed 

change does not alter or prevent the ability 

of structures, systems, and components 

(SSCs) to perform their intended function to 

mitigate the consequences of an initiating 

event within the assumed acceptance limits. 

The proposed change revises the TS for the 

CRE emergency ventilation system, which is 

a mitigation system designed to minimize 

unfiltered air leakage into the CRE and to 

filter the CRE atmosphere to protect the CRE 

occupants in the event of accidents 

previously analyzed. An important part of 

the CRE emergency ventilation system is the 

CRE boundary. The CRE emergency 

ventilation system is not an initiator or 

precursor to any accident previously 

evaluated. Therefore, the probability of any 

accident previously evaluated is not 

increased. Performing tests to verify the 

operability of the CRE boundary and 

implementing a program to assess and 

maintain CRE habitability ensure that the 

CRE emergency ventilation system is capable 

of adequately mitigating radiological 

consequences to CRE occupants during 

accident conditions, and that the CRE 

emergency ventilation system will perform as 

assumed in the consequence analyses of 

design basis accidents. Thus, the 

consequences of any accident previously 

evaluated are not increased. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

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 impact the 

accident analysis. The proposed change does 

not alter the required mitigation capability of 

the CRE emergency ventilation system, or its 

functioning during accident conditions as 

assumed in the licensing basis analyses of 

design basis accident radiological 

consequences to CRE occupants. No new or 

different accidents result from performing the 

new surveillance or following the new 

program. The proposed change does not 

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

no new or different type of equipment will 

be installed) or a significant change in the 

methods governing normal plant operation. 

The proposed change does not alter any 

safety analysis assumptions and is consistent 

with current plant operating practice. 

Therefore, this change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety 

The proposed change does not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The proposed 

change does not affect safety analysis 

acceptance criteria. The proposed change 

will not result in plant operation in a 

configuration outside the design basis for an 

unacceptable period of time without 

compensatory measures. The proposed 

change does not adversely affect systems that 

respond to safely shut down the plant and to 

maintain the plant in a safe shutdown 

condition. Therefore, the proposed change 

does not involve a significant reduction in a 

margin of safety. 

The NRC staff 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. 

R.E. Ginna Nuclear Power Plant, LLC, 

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

Power Plant, Wayne County, New York 

Date of amendment request: March 

23, 2009. 

Description of amendment request: 

The proposed amendment would delete 

those portions of the Technical 

Specifications (TSs) superseded by Part 

26, Subpart I of Title 10 of the Code of 

Federal Regulations (10 CFR). This 

change incorporates NRC approved 

Revision 0 of Technical Specification 

Task Force (TSTF) Improved Standard 

Technical Specification Change 

Traveler, TSTF–511, ‘‘Eliminate 

Working Hour Restrictions from TS 

5.2.2 to Support Compliance with 10 

CFR Part 26.’’ The availability of this TS 

improvement was announced as part of 

the consolidated line item improvement 

process (CLIIP) in the Federal Register 

on December 30, 2008 (73 FR 79923). 

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 removes technical 

specification restrictions on working hours 

for personnel who perform safety related 

functions. The technical specification 

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restrictions are superseded by the worker 

fatigue requirements in 10 CFR Part 26. 

Removal of the technical specification 

requirements will be performed concurrently 

with the implementation of 10 CFR Part 26, 

Subpart I, requirements. The proposed 

change does 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. Worker fatigue 

is not an assumption in the consequence 

mitigation of any accident previously 

evaluated. 

Therefore, it is concluded that 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 removes technical 

specification restrictions on working hours 

for personnel who perform safety related 

functions. The technical specification 

restrictions are superseded by the worker 

fatigue requirements in 10 CFR Part 26. 

Working hours will continue to be controlled 

in accordance with NRC requirements. The 

new rule allows for deviations from controls 

to mitigate or prevent a condition adverse to 

safety or as necessary to maintain the 

security of the facility. This ensures that the 

new rule will not unnecessarily restrict 

working hours and thereby create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed change does not alter the 

plant configuration, require new plant 

equipment to be installed, alter accident 

analysis assumptions, add any initiators, or 

effect the function of plant systems or the 

manner in which systems are operated, 

maintained, modified, tested, or inspected. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

Criterion 3: The Proposed Change Does Not 

Involve a Significant Reduction in a Margin 

of Safety 

The proposed change removes technical 

specification restrictions on working hours 

for personnel who perform safety related 

functions. The technical specification 

restrictions are superseded by the worker 

fatigue requirements in 10 CFR Part 26. 

The proposed change does not involve any 

physical changes to the plant or alter the 

manner in which plant systems are operated, 

maintained, modified, tested, or inspected. 

The proposed change does not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions or 

operation are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed change will not result 

in plant operation in a configuration outside 

the design basis. The proposed change does 

not adversely affect systems that respond to 

safely shutdown the plant and to maintain 

the plant in a safe shutdown condition. 

Removal of plant-specific technical 

specification administrative requirements 

will not reduce a margin of safety because the 

requirements in 10 CFR Part 26 are adequate 

to ensure that worker fatigue is managed. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Carey Fleming, 

Sr. Counsel—Nuclear Generation, 

Constellation Group, LLC, 750 East Pratt 

Street, 17 Floor, Baltimore, MD 21202. 

NRC Branch Chief: Mark G. Kowal. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: 

December 29, 2008. 

Description of amendment request: 

The amendment would revise Technical 

Specification (TS) 3.8.4, ‘‘DC [Direct 

Current] Sources—Operating,’’ and TS 

3.8.5, ‘‘DC Sources—Shutdown.’’ 

Specifically, this amendment would 

revise the battery connection resistance 

limits in Surveillance Requirement (SR) 

3.8.4.2 and SR 3.8.4.5 from 150 microohms (150E–6 ohm) to 69 micro-ohms 

(69E–6 ohm). TS 3.8.5 is affected by 

virtue of SR 3.8.5.1 invoking both SR 

3.8.4.2 and SR 3.8.4.5 for DC sources 

that are required to be operable in 

Modes 5 and 6. 

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 replaces a battery 

surveillance limit with a value based on 

voltage drop calculations for each of the four 

battery subsystems at Callaway under both 

normal operating and accident load profiles. 

The new value is more conservative, as well 

as being more appropriate, as an acceptance 

criterion for verifying battery operability 

pursuant to SR 3.8.4.2 and SR 3.8.4.5, thus 

providing greater assurance that the batteries 

can perform their specified safety functions 

with regard to accident mitigation. 

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 change 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 change 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 change does not physically 

alter safety-related systems nor affect the way 

in which safety-related systems perform their 

functions. 

All accident analysis acceptance criteria 

will continue to be met with the proposed 

change. The proposed change will not affect 

the source term, containment isolation, or 

radiological release assumptions used in 

evaluating the radiological consequences of 

any accident previously evaluated. The 

applicable radiological dose criteria will 

continue to be met. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

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 change 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 change does not, therefore, 

create the possibility of a new or different 

accident from any accident previously 

evaluated. 

3. Does the proposed change does not 

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 

(FDH), 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. 

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18258 Federal Register / Vol. 74, No. 75 / Tuesday, April 21, 2009 / Notices 

The proposed change does not eliminate 

any surveillances or alter the frequency of 

surveillances required by the Technical 

Specifications; however, the acceptance 

criterion for the specified battery resistance 

surveillances will be more restrictive. None 

of the acceptance criteria for any accident 

analysis will be changed. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: 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: March 4, 

2009. 

Description of amendment request: 

The proposed amendment consists of 

changes to the approved fire protection 

program as described in Wolf Creek 

Generating Station (WCGS) Updated 

Safety Analysis Report (USAR). 

Specifically, a deviation from certain 

technical requirements of Title 10 of the 

Code of Federal Regulations (10 CFR), 

Part 50, Appendix R, Section III.G.2, as 

documented in Appendix 9.5E of the 

WCGS USAR, is requested regarding the 

use of operator manual actions in lieu 

of meeting circuit separation protection 

criteria. Table 3–1 of the submittal dated 

March 4, 2009 (Agencywide Documents 

Access and Management System 

(ADAMS) Accession No. 

ML090771269), identifies the proposed 

feasible and reliable operator manual 

actions requested for permanent 

approval and Table 3–2 of the submittal 

identifies the proposed feasible operator 

manual actions requested for approval 

on an interim basis. The interim 

operator actions will be eliminated with 

the implementation of associated design 

change package. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The design function of structures, systems 

and components are not impacted by the 

proposed change. The proposed change 

involves the performance of operator manual 

actions to achieve and maintain safe 

shutdown in the event of a fire outside of the 

control room and will not initiate an event. 

The proposed change does not increase the 

probability of occurrence of a fire or any 

other accident previously evaluated. 

The proposed operator manual actions are 

feasible and reliable and demonstrate that the 

plant can be safely shutdown in the event of 

a fire. No significant consequences result 

from the performance of the proposed 

change. 

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 design function of structures, systems 

and components are not impacted by the 

proposed change. The proposed change 

involves the performance of operator manual 

actions to achieve and maintain safe 

shutdown in response to a fire outside of the 

control room. The operator manual actions 

do not involve new failure mechanisms or 

malfunctions that can initiate a new accident. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

For the permanent operator manual 

actions, adequate time is available to perform 

the proposed operator manual actions to 

account for uncertainties in estimates of the 

time available and in estimates of how long 

it takes to diagnose and execute the actions. 

The actions have been verified that they can 

be performed through demonstration and the 

actions are proceduralized. The proposed 

actions are feasible and reliable and 

demonstrate that the plant can be safely 

shutdown in the event of a fire. 

For the interim operator manual actions 

adequate time is available to feasibly perform 

the proposed operator manual actions and a 

compensatory measure fire watch is provided 

for the affected area as an added defense in 

depth fire protection measure. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: 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: March 6, 

2009. 

Description of amendment request: 

The proposed amendment would delete 

Technical Specification (TS) 5.2.2.d 

regarding the requirement to develop 

and implement administrative 

procedures to limit the working hours of 

personnel who perform safety-related 

functions. The requirements of TS 5.2.2 

have been superseded by Title 10 of the 

Code of Federal Regulations (10 CFR) 

Part 26, Subpart I. The change is 

consistent with U.S. Nuclear Regulatory 

Commission (NRC)-approved Revision 0 

to Technical Specification Task Force 

(TSTF) Improved Technical 

Specification Change Traveler, TSTF– 

511, ‘‘Eliminate Working Hour 

Restrictions from TS 5.2.2 to Support 

Compliance with 10 CFR Part 26.’’ 

The NRC staff issued a ‘‘Notice of 

Availability of Model Safety Evaluation, 

Model No Significant Hazards 

Determination, and Model Application 

for Licensees That Wish To Adopt 

TSTF–511, Revision 0, ‘Eliminate 

Working Hour Restrictions From TS 

5.2.2 To Support Compliance With 10 

CFR Part 26,’ ’’ in the Federal Register 

on December 30, 2008 (73 FR 79923). 

The notice included a model safety 

evaluation, a model no significant 

hazards consideration (NSHC) 

determination, and a model license 

amendment request, using the 

consolidated line item improvement 

process. In its application dated March 

6, 2009, the licensee affirmed the 

applicability of the model NSHC 

determination, which is presented 

below. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 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 

The proposed change removes Technical 

Specification restrictions on working hours 

for personnel who perform safety related 

functions. The Technical Specification 

restrictions are superseded by the worker 

fatigue requirements in 10 CFR Part 26. 

Removal of the Technical Specification 

requirements will be performed concurrently 

with the implementation of the 10 CFR Part 

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Federal Register / Vol. 74, No. 75 / Tuesday, April 21, 2009 / Notices 18259 

26, Subpart I, requirements. The proposed 

change does 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. Worker fatigue 

is not an initiator of any accident previously 

evaluated. Worker fatigue is not an 

assumption in the consequence mitigation of 

any accident previously evaluated. Therefore, 

it is concluded that 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 removes Technical 

Specification restrictions on working hours 

for personnel who perform safety related 

functions. The Technical Specification 

restrictions are superseded by the worker 

fatigue requirements in 10 CFR Part 26. 

Working hours will continue to be controlled 

in accordance with NRC requirements. The 

new rule allows for deviations from controls 

to mitigate or prevent a condition adverse to 

safety or as necessary to maintain the 

security of the facility. This ensures that the 

new rule will not unnecessarily restrict 

working hours and thereby create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. The proposed change does not 

alter the plant configuration, require new 

plant equipment to be installed, alter 

accident analysis assumptions, add any 

initiators, or [a]ffect the function of plant 

systems or the manner in which systems are 

operated, maintained, modified, tested, or 

inspected. Therefore, the proposed change 

does not create the possibility of a new or 

different kind of accident from any 

previously evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in a Margin 

of Safety 

The proposed change removes Technical 

Specification restrictions on working hours 

for personnel who perform safety related 

functions. The Technical Specification 

restrictions are superseded by the worker 

fatigue requirements in 10 CFR Part 26. The 

proposed change does not involve any 

physical changes to plant or alter the manner 

in which plant systems are operated, 

maintained, modified, tested, or inspected. 

The proposed change does not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed change will not result 

in plant operation in a configuration outside 

the design basis. The proposed change does 

not adversely affect systems that respond to 

safely shutdown the plant and to maintain 

the plant in a safe shutdown condition. 

Removal of plant-specific Technical 

Specification administrative requirements 

will not reduce a margin of safety because the 

requirements in 10 CFR Part 26 are adequate 

to ensure that worker fatigue is managed. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

analysis adopted by the licensee and, 

based on this review, it appears that the 

three standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jay Silberg, Esq., 

Pillsbury Winthrop Shaw Pittman LLP, 

2300 N Street, NW., Washington, DC 

20037. 

NRC Branch Chief: Michael T. 

Markley. 

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. 

Duke Energy Carolinas, LLC, et al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina 

Date of application for amendments: 

June 23, 2008. 

Brief description of amendments: This 

request modifies the subject Technical 

Specifications (TSs) and Bases by 

changing the logic configuration of TS 

Table 3.3.2–1, ‘‘Engineered Safety 

Feature Actuation System 

Instrumentation,’’ Function 5.b.(5), 

‘‘Turbine Trip and Feedwater Isolation, 

Feedwater Isolation, Doghouse Water 

Level—High High.’’ The existing oneout-of-one (1/1) logic per train per 

doghouse is being modified to a twoout-of-three (2/3) logic per train per 

doghouse. The proposed change will 

improve the overall reliability of this 

function and will reduce the potential 

for spurious actuations. 

Date of issuance: April 2, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 249/243. 

Facility Operating License Nos. NPF– 

35 and NPF–52: Amendments revised 

the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: February 24, 2009 (74 FR 

8276). 

The Commission’s related evaluation, 

state consultation, and final no 

significant hazards consideration 

determination of the amendments are 

contained in a Safety Evaluation dated 

April 2, 2009. 

No significant hazards consideration 

comments received: No. 

Duke Power Company LLC, Docket Nos. 

50–369 and 50–370, McGuire Nuclear 

Station, Units 1 and 2, Mecklenburg 

County, North Carolina 

Date of application for amendments: 

March 20, 2008. 

Brief description of amendments: The 

proposed amendments would revise the 

McGuire licensing basis by adopting the 

Alternative Source Term (AST) 

radiological analysis methodology as 

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18260 Federal Register / Vol. 74, No. 75 / Tuesday, April 21, 2009 / Notices 

allowed by 10 CFR 50.67, ‘‘Accident 

source term,’’ for the Loss of Coolant 

Accident. This amendment request 

represents full scope implementation of 

the AST as described in Nuclear 

Regulatory Commission (NRC) 

Regulatory Guide 1.183, ‘‘Alternative 

Radiological Source Terms for 

Evaluating Design Basis Accidents at 

Nuclear Power Reactors, Revision 0.’’ 

Selective implementation of AST for the 

McGuire Fuel Handling Accidents was 

approved by the NRC on December 22, 

2006. There are no changes proposed to 

the McGuire Technical Specifications 

within this amendment request. The 

application of the AST methodology to 

the Loss of Coolant Accident (LOCA) 

radiological analysis will allow McGuire 

to resolve the Control Room envelope 

degraded boundary condition as 

discussed in McGuire’s response to NRC 

Generic Letter 2003–01, ‘‘Control Room 

Habitability,’’ dated February 19, 2004. 

By separate amendment request dated 

January 22, 2008, Duke proposed to 

revise the McGuire Technical 

Specification (TS) requirements related 

to control room envelope habitability in 

TS 3.7.9, ‘‘Control Room Area 

Ventilation System.’’ The proposed 

changes are consistent with the Industry 

and NRC-approved Technical 

Specification Task Force (TSTF) change 

TSTF–448, Control Room Habitability, 

Revision 3 and the NRC Consolidated 

Line Item Improvement Process (CLIIP). 

Duke has performed a review of all 

McGuire License Amendment Requests 

(LAR) currently under review by the 

NRC for impacts to this AST LAR. None 

of these LARs impact any assumptions 

or results of the LOCA AST radiological 

analysis. 

Date of issuance: March 31, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: 251 and 231. 

Renewed Facility Operating License 

Nos. NPF–9 and NPF–17: The 

amendments revised the license. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): The notice 

provided an opportunity to submit 

comments on the Commission’s 

proposed NSHC determination by 

March 30, 2009. No comments have 

been received to date. However, the 

notice also provided an opportunity to 

request a hearing by April 28, 2009, but 

indicated that if the Commission make 

a final NSHC determination, any such 

hearing would take place after issuance 

of the amendment. 

Date of initial notice in Federal 

Register: February 27, 2009 (74 FR 

9009). 

The supplements dated May 28, 2008, 

October 6, 2008, December 17, 2008 and 

February 12, 2009, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 31, 2009. 

No significant hazards consideration 

comments received: No. 

Entergy Gulf States Louisiana, LLC, and 

Entergy Operations, Inc., Docket No. 50– 

458, River Bend Station, Unit 1, West 

Feliciana Parish, Louisiana 

Date of amendment request: 

December 8, 2008. 

Brief description of amendment: The 

amendment added a license condition 

to allow a one-time extension of 

surveillance requirements involving the 

18-month channel calibration and logic 

system functional tests for one channel 

of the reactor water level 

instrumentation system. The extension 

is to account for the effects of 

rescheduling the next refueling outage 

from early to late 2009. 

Date of issuance: April 1, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 15 days from the date of 

issuance. 

Amendment No.: 162. 

Facility Operating License No. NPF– 

47: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: January 27, 2008 (74 FR 4770). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated April 1, 2009. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50– 

313, Arkansas Nuclear One, Unit No. 1, 

Pope County, Arkansas 

Date of amendment request: July 21, 

2008. 

Brief description of amendment: The 

amendment deleted the exception to 

Limiting Condition for Operation (LCO) 

3.0.4 to the 30-day allowable outage 

time of the Startup No. 2 Transformer 

and corrected a spelling error in 

Technical Specification (TS) 3.8.1. The 

NRC approved the adoption of Industry/ 

TS Task Force (TSTF) change TSTF– 

359, ‘‘Increased Flexibility in Mode 

Restraints,’’ for ANO–1 in TS 

Amendment 232 dated April 2, 2008 

(Agencywide Documents Access and 

Management System (ADAMS) 

Accession No. ML080600006). The 

intent of TSTF–359 was to eliminate 

exceptions to LCO 3.0.4 within 

individual specifications and provide 

requirements within LCO 3.0.4 to 

control mode changes when TS-required 

equipment is inoperable. The licensee 

omitted deleting this LCO 3.0.4 

exception in its October 22, 2007 

(ADAMS Accession No. ML073030542), 

amendment request to adopt TSTF–359. 

Date of issuance: March 30, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 236. 

Renewed Facility Operating License 

No. DPR–51: Amendment revised the 

Technical Specifications/license. 

Date of initial notice in Federal 

Register: October 21, 2008 (73 FR 

62563). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 30, 2009. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of application for amendment: 

December 16, 2008, as supplemented by 

letter dated February 19, 2009. 

Brief description of amendment: This 

amendment request would revise the 

Technical Specifications Section 2.1.2, 

Safety Limit Minimum Critical Power 

Ratio (SLMCPR) for two-loop and 

single-loop operation. 

Date of issuance: March 26, 2009. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 232. 

Facility Operating License No. DPR– 

35: The amendment revised the License 

and Technical Specifications. 

Date of initial notice in Federal 

Register: January 23, 2009 (74 FR 4250). 

The supplemental letter dated 

February 19, 2009, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. The 

Commission’s related evaluation of this 

amendment is contained in a Safety 

Evaluation dated March 26, 2009. 

No significant hazards consideration 

comments received: No. 

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Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and STN 50– 

457, Braidwood Station, Units 1 and 2, 

Will County, Illinois; Docket Nos. STN 

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

Unit Nos. 1 and 2, Ogle County, Illinois 

Date of application for amendment: 

April 9, 2008, as supplemented by letter 

dated October 1, 2008. 

Brief description of amendment: The 

amendments revise Technical 

Specifications (TSs) 5.5.6, Pre-Stressed 

Concrete Containment Tendon 

Surveillance Program, and 5.6.8, 

Tendon Surveillance Report, for 

consistency with the requirements of 

Title 10 Code of Federal Regulations (10 

CFR) Section 50.55a, Codes and 

standards, paragraph (g)(4) for 

components classified as American 

Society of Mechanical Engineers Boiler 

and Pressure Vessel Code (ASME Code) 

Class CC, by replacing the reference to 

the specific ASME Code year for the 

tendon surveillance program with a 

requirement to use the applicable ASME 

Code and addenda as required by 10 

CFR 50.55a. 

Date of issuance: March 26, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: Braidwood Unit 1– 

158; Braidwood Unit 2–158; Byron Unit 

No. 1–163; and Byron Unit No. 2–163. 

Facility Operating License Nos. NPF– 

72, NPF–77, NPF–37, and NPF–66: The 

amendments revise the TSs and 

Licenses. 

Date of initial notice in Federal 

Register: July 1, 2008 (73 FR 37504). 

The October 1, 2008, supplemental 

letter provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the NRC staff’s original proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 26, 2009. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station (CPS), Unit No. 1, DeWitt 

County, Illinois 

Date of application for amendment: 

September 2, 2008. 

Brief description of amendment: The 

amendment requested to amend the CPS 

Unit No. 1 Technical Specifications (TS) 

to relocate the TS surveillance 

requirement (SR) 3.8.3.6 from the TS to 

a licensee-controlled document. SR 

3.8.3.6 requires the emergency diesel 

generator fuel oil storage tanks to be 

drained, sediment removed, and 

cleaned on a 10-year interval. The 

request is submitted consistent with the 

guidance contained in Nuclear 

Regulatory Commission (NRC)-approved 

Technical Specifications Task Force 

Report 2 (TSTF–2). 

Date of issuance: April 2, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 186. 

Facility Operating License No. NPF– 

62: The amendment revised the 

Technical Specifications and License. 

Date of initial notice in Federal 

Register: November 4, 2008 (73 FR 

65687) and January 27, 2009 (74 FR 

4771). The notice on January 27, 2009, 

was inadvertently placed in the Federal 

Register a second time 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 April 2, 2009. 

No significant hazards consideration 

comments received: No. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–412, 

Beaver Valley Power Station, Unit No. 2 

(BVPS–2), Beaver County, Pennsylvania 

Date of application for amendment: 

November 7, 2008. 

Brief description of amendment: The 

amendment modifies the method used 

to calculate the available net positive 

suction head (NPSH) for the BVPS–2 

recirculation spray (RS) pumps as 

described in the BVPS–2 Updated Final 

Safety Analysis Report (UFSAR). The 

BVPS–2 UFSAR takes credit for 

containment overpressure by allowing 

for the difference between containment 

total pressure and the vapor pressure of 

the water in the containment sump in 

the available NPSH calculation. 

Date of issuance: March 26, 2009. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 30 days. 

Amendment No.: 167. 

Facility Operating License No. NPF– 

73: The amendment revised the License 

and the Updated Final Safety Analysis 

Report. 

Date of initial notice in Federal 

Register: December 16, 2008 (73 FR 

76411). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 26, 2009. 

No significant hazards consideration 

comments received: No. 

Indiana Michigan Power Company, 

Docket Nos. 50–315 and 50–316, Donald 

C. Cook Nuclear Plant, Units 1 and 2 

(CNP–1 and CNP–2), Berrien County, 

Michigan 

Date of application for amendment: 

October 21, 2008. 

Brief description of amendment: The 

amendment modifies Technical 

Specification 5.6.3, ‘‘Radioactive 

Effluent Release Report,’’ by changing 

the required annual submittal date for 

the report from ‘‘within 90 days of 

January 1 of each year’’ (i.e., prior to 

April 1), to ‘‘prior to May 1 of each 

year.’’ 

Date of issuance: March 30, 2009. 

Effective date: As of the date of 

issuance. 

Amendment Nos.: 308 (CNP–1), 290 

(CNP–2). 

Facility Operating License Nos. DPR– 

58 and DPR–74: Amendments revised 

the Renewed Operating Licenses and 

Technical Specifications. 

Date of initial notice in Federal 

Register: December 16, 2008 (73 FR 

76412). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 30, 2009. 

No significant hazards consideration 

comments received: No. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

No. 1, Washington County, Nebraska 

Date of amendment request: April 22, 

2008, as supplemented by letter dated 

March 6, 2009. 

Brief description of amendment: The 

amendment modifies the Technical 

Specification (TS) 2.7, ‘‘Electrical 

Systems,’’ Limiting Condition for 

Operation (LCO) 2.7(2)j related to the 

allowed outage time for the Emergency 

Diesel Generators (EDGs). The change 

clarifies LCO 2.7(2)j such that a single 

period of inoperability for one EDG is 

limited to 7 consecutive days and that 

the cumulative total time of 

inoperability for both EDGs during any 

calendar month cannot exceed 7 days. 

Date of issuance: March 27, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment No.: 258. 

Renewed Facility Operating License 

No. DPR–40: The amendment revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: June 17, 2008 (73 FR 34342). 

The supplemental letter dated March 6, 

2009, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

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18262 Federal Register / Vol. 74, No. 75 / Tuesday, April 21, 2009 / Notices 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

safety evaluation dated March 27, 2009. 

No significant hazards consideration 

comments received: No. 

Pacific Gas and Electric Company, 

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

Canyon Nuclear Power Plant, Unit Nos. 

1 and 2, San Luis Obispo County, 

California 

Date of application for amendments: 

April 3, 2008, as supplemented by 

letters dated June 20, October 1, 

November 6, and December 16, 2008. 

Brief description of amendments: The 

amendments revised Technical 

Specification (TS) 3.7.5, ‘‘Auxiliary 

Feedwater (AFW) System,’’ to remove 

Surveillance Requirement (SR) 3.7.5.6, 

and revised TS 3.7.6, ‘‘Condensate 

Storage Tank (CST) and Fire Water 

Storage Tank (FWST),’’ to remove the 

FWST level requirements, revise the 

CST level requirements, and revise TS 

3.7.6 to be consistent with the NUREG– 

1431, ‘‘Standard Technical 

Specifications (STS).’’ Specifically, 

these changes reflect design changes 

made to the CSTs and are necessary to 

support the on-line refurbishment of the 

FWST and replacement of the 

recirculation piping for the fire water 

pumps. The design changes to the CSTs 

are intended to eliminate the reliance on 

the FWST for additional seismicallyqualified feedwater supply and thus, 

make the existing TS requirements for 

the FWST unnecessary. 

Date of issuance: March 30, 2009. 

Effective date: As of its date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: Unit 1–204; Unit 

2–205. 

Facility Operating License Nos. DPR– 

80 and DPR–82: The amendments 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in Federal 

Register: July 29, 2008 (78 FR 43956). 

The supplemental letters dated June 20, 

October 1, November 6, and December 

16, 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 March 30, 2009. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket No. 

50–390, Watts Bar Nuclear Plant (WBN), 

Unit 1, Rhea County, Tennessee 

Date of application for amendment: 

September 18, 2008. 

Brief description of amendment: The 

amendment revised WBN Unit 1 

Technical Specification 3.8.7, 

‘‘Inverters—Operating.’’ The 

amendment revised the requirement to 

two inverters for each of the four 

channels. 

Date of issuance: March 24, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 240 days of issuance. 

Amendment No.: 76. 

Facility Operating License No. NPF– 

90: Amendment revises the Technical 

Specification 3.8.7 and Updated Final 

Safety Analysis Report. 

Date of initial notice in Federal 

Register: November 4, 2008 (73 FR 

65697). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 24, 2009. 

No significant hazards consideration 

comments received: No. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: March 

14, 2007, as supplemented by letters 

dated April 18, May 9, June 15, August 

31, September 12 and 20, October 16, 

November 16, two letters dated 

December 14, and December 18, 2007; 

two letters dated January 18, January 31, 

February 26 and 28, March 14, April 26, 

May 14, June 19, and July 31, 2008; and 

January 16 and 29, and February 17 and 

27, 2009. 

Brief description of amendment: The 

amendment revised the licensing basis 

for the Main Steam and Feedwater 

Isolation System (MSFIS) controls to 

incorporate field programmable gate 

array technology. Other related changes 

requested in the March 14, 2007, 

application were previously approved 

in Amendment No. 174, dated August 

28, 2007, Amendment No. 175, dated 

March 3, 2008, Amendment No. 176, 

dated March 21, 2008, and Amendment 

No. 177, dated April 3, 2008. 

Date of issuance: March 31, 2009. 

Effective date: Effective as of date of 

issuance and shall be implemented 

before entry into Mode 3 in the restart 

from Refueling Outage 17. 

Amendment No.: 181. 

Renewed Facility Operating License 

No. NPF–42. The amendment revised 

the Operating License. 

Date of initial notice in Federal 

Register: June 19, 2007 (72 FR 33785). 

The supplemental letters dated April 18, 

May 9, June 15, August 31, September 

12 and 20, October 16, November 16, 

two letters dated December 14, and 

December 18, 2007; two letters dated 

January 18, January 31, February 26 and 

28, March 14, April 26, May 14, June 19, 

and July 31, 2008; and January 16 and 

29, and February 17 and 27, 2009, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 31, 2009. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 10th day 

of April 2009. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E9–8832 Filed 4–20–09; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2009–0176] 

Draft Regulatory Guide: Issuance, 

Availability 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Notice of Issuance and 

Availability of Draft Regulatory Guide, 

DG–1214. 

FOR FURTHER INFORMATION CONTACT: Dan 

Frumkin, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, telephone (301) 415–2280, e-mail 

Dan.Frumkin@nrc.gov, or, R. A. Jervey, 

U.S. Nuclear Regulatory Commission, 

Washington, DC 20555–0001, telephone: 

(301) 251–7407, e-mail to raj@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Introduction 

The U.S. Nuclear Regulatory 

Commission (NRC) is issuing for public 

comment a draft guide in the agency’s 

‘‘Regulatory Guide’’ series. This series 

was developed to describe and make 

available to the public such information 

mstockstill on PROD1PC66 with NOTICES