23440 Federal Register / Vol. 74, No. 95 / Tuesday, May 19, 2009 / Notices 

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

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

Day Event 

205 ......................................................... Deadline for petitioner to seek reversal of a final adverse NRC staff determination either before the 

presiding officer or another designated officer. 

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

A+3 ......................................................... If access granted: Issuance of presiding officer or other designated officer decision on motion for protective order for access to sensitive information (including schedule for providing access and submission of contentions) or decision reversing a final adverse determination by the NRC staff. 

A+28 ....................................................... Deadline for submission of contentions whose development depends upon access to SUNSI. However, if more than 25 days remain between the petitioner’s receipt of (or access to) the information 

and the deadline for filing all other contentions (as established in the notice of hearing or opportunity 

for hearing), the petitioner may file its SUNSI contentions by that later deadline. 

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

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

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

[FR Doc. E9–11604 Filed 5–18–09; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2009–0204] 

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 April 23, 

2009, to May 6, 2009. The last biweekly 

notice was published on May 5, 2009 

(74 FR 20741). 

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 

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Federal Register / Vol. 74, No. 95 / Tuesday, May 19, 2009 / Notices 23441 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/ 

requestor to relief. A petitioner/ 

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

All documents filed in NRC 

adjudicatory proceedings, including a 

request for hearing, a petition for leave 

to intervene, any motion or other 

document filed in the proceeding prior 

to the submission of a request for 

hearing or petition to intervene, and 

documents filed by interested 

governmental entities participating 

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

accordance with the NRC E-Filing rule, 

which the NRC promulgated in August 

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

process requires participants to submit 

and serve all adjudicatory documents 

over the Internet or in some cases to 

mail copies on electronic storage media. 

Participants may not submit paper 

copies of their filings unless they seek 

a waiver in accordance with the 

procedures described below. 

To comply with the procedural 

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

days prior to the filing deadline, the 

petitioner/requestor must contact the 

Office of the Secretary by e-mail at 

hearingdocket@nrc.gov, or by calling 

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

ID certificate, which allows the 

participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and/or (2) creation of an 

electronic docket for the proceeding 

(even in instances in which the 

petitioner/requestor (or its counsel or 

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

petitioner/requestor will need to 

download the Workplace Forms 

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 

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: 

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23442 Federal Register / Vol. 74, No. 95 / Tuesday, May 19, 2009 / Notices 

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– 

4209, (301) 415–4737 or by e-mail to 

pdr.resource@nrc.gov. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units 1, 2, and 3, 

Maricopa County, Arizona 

Date of amendment request: February 

19, 2009. 

Description of amendment request: 

The amendments would relocate the 

reactor coolant system pressure and 

temperature (P/T) limits and the low 

temperature overpressure protection 

(LTOP) enable temperatures to a 

licensee-controlled document outside of 

the Technical Specifications (TSs). The 

P/T limits and LTOP enable 

temperatures would be specified in a 

Pressure and Temperature Limits Report 

(PTLR) that would be located in the Palo 

Verde Nuclear Generating Station 

(PVNGS) Technical Requirements 

Manual and administratively controlled 

by a new TS 5.6.9. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This proposed change revises the 

Technical Specifications by relocating the 

reactor coolant system (RCS) pressure and 

temperature limits, heatup and cooldown 

curves and low temperature overpressure 

protection (LTOP) enable temperatures from 

the Technical Specifications to an [Arizona 

Public Service] APS-controlled RCS Pressure 

and Temperature Limits Report (PTLR), and 

requiring that the limits in the PTLR be 

determined using the analytical methods 

described in the NRC-approved Topical 

Report CE NPSD–683–A. Relocation of this 

information and updating it using NRCapproved methodology will not alter the 

requirement to update the RCS pressure and 

temperature curves and limits in accordance 

with 10 CFR 50 Appendices G and H. 

Updating the P/T curves and LTOP limits 

ensures the reactor coolant system’s pressure 

boundary integrity is protected throughout 

plant life. Consequently, this proposed 

change is determined to not contribute to an 

increase in the probability of, or the initiation 

of, a design basis accident. Similarly, the 

safety analysis information presented in the 

Updated Final Safety Analysis Report 

remains unchanged. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change revises the Technical 

Specifications by relocating the RCS pressure 

and temperature limits, heatup and 

cooldown curves and LTOP enable 

temperatures from the Technical 

Specifications to a PVNGS PTLR, and 

requiring that the limits in the PTLR be 

determined using the analytical methods 

described in the NRC-approved Topical 

Report CE NPSD–683–A. The PTLR 

documents removal, testing and analyzing 

the surveillance capsules, and will be 

updated by APS to reflect the results of 

testing and analysis of surveillance 

specimens withdrawn in the future. Removal, 

testing and analysis of surveillance 

specimens may result in a need to implement 

changes to the RCS pressure and temperature 

limits. Such changes are implemented to 

ensure the integrity of the RCS pressure 

boundary throughout plant lifetime. Updates 

to the RCS pressure and temperature curves 

and limits will not create a new or different 

kind of accident. Relocating the P/T curves, 

heatup and cooldown rates and LTOP limits 

to the PTLR has no impact on any safety 

analyses. 

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. 

Pressure and temperature curves and limits 

are provided as limits to plant operation to 

ensure RCS pressure boundary integrity is 

maintained throughout the plant’s lifetime. 

Changes to the RCS pressure and temperature 

curves and limits, resulting from the removal, 

testing and analysis of surveillance capsules, 

are only made within the acceptable margin 

limits thereby maintaining the required 

margin of safety. There is no change to the 

safety analysis. 

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 that 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the request 

for amendments involves no significant 

hazards consideration. 

Attorney for licensee: Michael G. 

Green, Senior Regulatory Counsel, 

Pinnacle West Capital Corporation, P.O. 

Box 52034, Mail Station 8695, Phoenix, 

Arizona 85072–2034. 

NRC Branch Chief: Michael T. 

Markley. 

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: 

December 1, 2008. 

Description of amendment request: 

The proposed amendments would 

correct a non-conservative Technical 

Specification (TS) Surveillance 

Requirement by revising McGuire TS 

3.8.1.4 to increase the minimum 

required amount of fuel oil for the 

Emergency Diesel Generators fuel oil 

day tank as read on the local fuel gauge 

used to perform the surveillance. 

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Federal Register / Vol. 74, No. 95 / Tuesday, May 19, 2009 / Notices 23443 

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. 

Implementation of the proposed 

amendment does not significantly increase 

the probability or the consequences of an 

accident previously evaluated. The 

Emergency Diesel Generators (EDGs) and 

their associated emergency buses function as 

accident mitigators. The proposed changes 

do not involve a change in the operational 

limits or the design of the electrical power 

systems (particularly the emergency power 

systems) or change the function or operation 

of plant equipment or affect the response of 

that equipment when called upon to operate. 

The proposed change to TS SR 3.8.1.4 

confirms the minimum supply of fuel oil in 

the emergency diesel generators (EDG) fuel 

oil day tank. The minimum value for the 

affected parameter is being increased in the 

conservative direction and further ensures 

the EDGs ability to fulfill their safety related 

function. Thus, based on the above, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not involve a 

change in the operational limits or the design 

capabilities of the emergency electrical 

power systems. The proposed changes do not 

change the function or operation of plant 

equipment or introduce any new failure 

mechanisms. The evaluation that supports 

this LAR included a review of the EDG fuel 

oil system to which this parameter applies. 

The proposed changes do not introduce any 

new or different types of failure mechanisms; 

plant equipment will continue to respond as 

designed and analyzed. 

3. Does the proposed amendment involve 

a significant reduction in the margin of 

safety? 

Response: No. 

Margin of safety is related to the 

confidence in the ability of the fission 

product barriers to perform their design 

functions during and following an accident 

situation. These barriers include the fuel 

cladding, the reactor coolant system, and the 

containment system. The performance of the 

fuel cladding, the reactor coolant system and 

the containment system will not be adversely 

impacted by the proposed changes. Thus, it 

is concluded that the proposed TS and TS 

Basis changes do not involve a significant 

reduction in the margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, 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 Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit No. 2, 

Westchester County, New York 

Date of amendment request: March 5, 

2009. 

Description of amendment request: 

The proposed amendment will revise 

the Reactor Vessel Heatup, Cooldown, 

and Low Temperature Overpressure 

Protection curves in Technical 

Specifications (TSs) 3.4.3 and 3.4.12 to 

incorporate the most recent estimates of 

lifetime neutron fluence and the effects 

of the Stretch Power Uprate 

(Amendment No. 241). 

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. Operation of the facility in accordance 

with the proposed amendment would not 

involve a significant increase in the 

probability of occurrence or consequences of 

an accident previously evaluated. 

The proposed TS changes do not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. There are no physical changes to 

the plant being introduced by the proposed 

changes to the heatup and cooldown 

limitation curves. The proposed changes do 

not modify the RCS [Reactor Coolant System] 

pressure boundary. That is, there are no 

changes in operating pressure, materials, or 

seismic loading. The proposed changes do 

not adversely affect the integrity of the RCS 

pressure boundary such that its function in 

the control of radiological consequences is 

affected. The proposed heatup and cooldown 

limitation curves were generated in 

accordance with the fracture toughness 

requirements of 10 CFR 50 [Title 10 of the 

Code of Federal Regulations Part 50] 

Appendix G, and ASME B&PV code 

[American Society of Mechanical Engineers 

Boiler and Pressure Vessel Code], Section XI, 

Appendix G edition with 2000 Addenda. The 

proposed heatup and cooldown limitation 

curves were established in compliance with 

the methodology used to calculate and 

predict effects of radiation on embrittlement 

of RPV [Reactor Pressure Vessel] beltline 

materials. Use of this methodology provides 

compliance with the intent of 10 CFR 50 

Appendix G and provides margins of safety 

that ensure non-ductile failure of the RPV 

will not occur. The proposed heatup and 

cooldown limitation curves prohibit 

operation in regions where it is possible for 

non-ductile failure of carbon and low alloy 

RCS materials to occur. Hence, the primary 

coolant pressure boundary integrity will be 

maintained throughout the limit of 

applicability of the curves, 29.2 EFPY 

[Effective Full-Power Years]. 

Operation within the proposed LTOPS 

[Low Temperature Overpressure Protection 

System] limits ensures that 

overpressurization of the RCS at low 

temperatures will not result in component 

stresses in excess of those allowed by the 

ASME B&PV Code Section XI Appendix G. 

Consequently, the proposed changes do not 

involve a significant increase in the 

probability or the consequences of an 

accident previously evaluated. 

2. Operation of the facility in accordance 

with the proposed amendment would not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

The proposed TS changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. No new modes of operation are 

introduced by the proposed changes. The 

proposed changes will not create any failure 

mode not bounded by previously evaluated 

accidents. Further, the proposed changes to 

the heatup and cooldown limitation curves 

and the LTOPS limits do not affect any 

activities or equipment other than the RCS 

pressure boundary and do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Consequently, the proposed changes do not 

involve a significant increase in the 

probability or consequence of a new or 

different kind of accident, from any accident 

previously evaluated. 

3. Operation of the facility in accordance 

with the proposed amendment would not 

involve a significant reduction in the margin 

of safety. 

The proposed TS changes do not involve 

a significant reduction in the margin of 

safety. 

The revised heatup and cooldown 

limitation curves and LTOPS limits are 

established in accordance with current 

regulations and the ASME B&PV Code 1998 

edition with 2000 Addenda. These proposed 

changes are acceptable because the ASME 

B&PV Code maintains the margin of safety 

required by 10 CFR 50.55(a). Because 

operation will be within these limits, the RCS 

materials will continue to behave in a nonbrittle manner consistent with the original 

design bases. 

The proposed changes to the allowable 

operation of charging and safety injection 

pumps when LTOPS is required to be 

operable is consistent with the IP2 licensing 

bases as established in TS Amendment 224. 

Therefore, Entergy has concluded that the 

proposed changes do not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

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23444 Federal Register / Vol. 74, No. 95 / Tuesday, May 19, 2009 / Notices 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Acting Branch Chief: Richard V. 

Guzman. 

Entergy Nuclear Operations, Inc., 

Docket Nos. 50–247, Indian Point 

Nuclear Generating Unit No. 2, 

Westchester County, New York 

Date of amendment request: March 

25, 2009. 

Description of amendment request: 

The proposed amendment would add 

two Emergency Core Cooling System 

(ECCS) valves to Surveillance 

Requirement (SR) 3.5.2.1. The SR is 

designed to verify that ECCS valves 

whose single failure could cause loss of 

the ECCS function are in the required 

position with ac power removed so that 

misalignment or single failure cannot 

prevent completion of the ECCS 

function. Entergy plans to install an 

alternate source of power during the 

spring 2010 refueling outage to provide 

the required position indication. The 

proposed changes support Entergy’s 

resolution to Generic Letter (GL) 2004– 

02 by establishing a licensing basis that 

supports meeting the regulatory 

requirements of the GL. 

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 adds two ECCS 

valves to SR 3.5.2.1. The purpose of the 

surveillance is to assure that the valves are 

in their required position with normal ac 

power removed from the valve operator so 

that misalignment or single failure cannot 

prevent completion of the ECCS function. 

The performance of the SR does not involve 

any actions related to the initiation of an 

accident and therefore the proposed changes 

cannot increase the probability of an 

accident. Misalignment or single failure of 

one of the two valves being added to TS 

[Technical Specifications] could cause a loss 

of the ECCS function based on GSI [Generic 

Safety Issue]-191 evaluations, so the change 

will not increase the consequences of an 

accident but rather provide assurance that no 

such increase can occur. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response—No. 

The proposed change adds two ECCS 

valves to SR 3.5.2.1. The purpose of the 

surveillance is to assure that the valves are 

in their required position with normal ac 

power removed from the valve operators so 

that misalignment or single failure cannot 

prevent completion of the ECCS function. 

The provision of alternate power to the 

existing valve position indication during the 

upcoming spring 2010 outage (2R19), will 

allow the valve operators to be normally 

deenergized. The change assures that the 

valves will be in their correct position and 

does not introduce any new failure modes or 

the possibility of a different 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 change involve a 

significant reduction in a margin of safety? 

Response—No. 

The proposed change adds two ECCS 

valves to SR 3.5.2.1. The purpose of the 

surveillance is to assure that the valves are 

in their required position with normal ac 

power removed so that misalignment or 

single failure cannot prevent completion of 

the ECCS function. The valves will be reenergized 24 hours following a DBA [designbasis accident] and therefore will be capable 

of performing their required function of 

isolating a potential passive failure at that 

time. This ensures that the ECCS function 

can be performed without a reduction in the 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Acting Branch Chief: Richard V. 

Guzman. 

Entergy Nuclear Operations, Inc., 

Docket Nos. 50–247, Indian Point 

Nuclear Generating Unit No. 2, 

Westchester County, New York 

Date of amendment request: March 

29, 2009. 

Description of amendment request: 

The proposed amendment will establish 

a more restrictive acceptance criterion 

for surveillance requirement (SR) 3.8.6.6 

regarding periodic verification of 

capacity for the affected station 

batteries. 

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? 

No. The proposed change revises the 

acceptance criterion applied to an existing 

surveillance test for the Indian Point 2 station 

batteries. Performing a technical specification 

surveillance test is not an accident initiator 

and does not increase the probability of an 

accident occurring. The proposed revision to 

the test acceptance criterion is based on the 

design calculation for battery performance at 

the minimum design temperature. The 

proposed new value for the test acceptance 

criteria is more limiting than the existing 

value which does not account for the 

minimum environmental design temperature 

assumed for the limiting battery locations. 

Establishing a test acceptance criterion that 

bounds existing or assumed conditions 

validates the equipment performance 

assumptions used in the accident mitigation 

safety analyses. 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? 

No. The proposed change revises the test 

acceptance criterion for an existing technical 

specification surveillance test conducted on 

the existing station batteries. The proposed 

change does not involve installation of new 

equipment or modification of existing 

equipment, so that no new equipment failure 

modes are introduced. Also, the proposed 

change in test acceptance criterion does not 

result in a change to the way that the 

equipment or facility is operated so that no 

new accident initiators are created. 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? 

No. The conduct of performance tests on 

safety-related plant equipment is a means of 

assuring that the equipment is capable of 

performing its intended safety function and 

therefore maintaining the margin of safety 

established in the safety analysis for the 

facility. The proposed change in the 

acceptance criterion for the battery capacity 

surveillance test is more conservative and 

more restrictive than the value currently in 

the technical specification and is based on 

the applicable design calculation for these 

components. 

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 

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proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Acting Branch Chief: Richard V. 

Guzman. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and STN 50– 

457, Braidwood Station, Units 1 and 2, 

Will County, Illinois; Docket Nos. STN 

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

Unit Nos. 1 and 2, Ogle County, Illinois 

Date of amendment request: March 

26, 2009. 

Description of amendment request: 

The proposed amendments would 

revise the fire protection program (FPP) 

to eliminate the requirement for the 

backup manual carbon dioxide (CO2) 

fire suppression system in the upper 

cable spreading rooms. 

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 the FPP to 

eliminate the requirement to maintain the 

backup CO2 fire suppression system for the 

upper cable spreading rooms. With the 

exception of the CO2 fire suppression system 

itself, the proposed change does not result in 

any physical changes to safety related 

structures, systems, or components [SSCs], or 

the manner in which they are operated, 

maintained, modified, tested, or inspected. 

The proposed change does not degrade the 

performance or increase the challenges of any 

safety related SSCs assumed to function in 

the accident analysis. The proposed change 

does not change the probability of a fire 

occurring since the fire ignition frequency is 

independent of the method of fire 

suppression. The proposed change does not 

affect the consequences of an accident 

previously evaluated since the fire safe 

shutdown analysis assumes fire damage 

throughout the affected fire area. The results 

of a fire in the upper cable spreading room 

would only affect one engineered safety 

features division. Sufficient redundancy 

exists in the engineered safety features fed 

from the other division to achieve a reactor 

shutdown and to maintain the reactor in a 

safe shutdown condition. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change revises the FPP to 

eliminate the requirement to maintain the 

backup CO2 fire suppression system for the 

upper cable spreading rooms. With the 

exception of the CO2 fire suppression system 

itself, the proposed change does not result in 

any physical changes to safety related 

structures, systems, or components, or the 

manner in which they are operated, 

maintained, modified, tested, or inspected. 

The proposed change does not degrade the 

performance or increase the challenges of any 

safety related SSCs assumed to function in 

the accident analysis. As a result, the 

proposed change does not introduce nor 

increase the number of failure mechanisms of 

a new or different type than those previously 

evaluated. The fire safe shutdown analysis 

assumes fire damage throughout the area 

consistent with a complete lack of fire 

suppression capability. Potential habitability 

hazards associated with actuation of the CO2

system are eliminated with the proposed 

change. 

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 the FPP to 

eliminate the requirement to maintain the 

backup CO2 fire suppression system for the 

upper cable spreading rooms. With the 

exception of the CO2 fire suppression system 

itself, the proposed change does not result in 

any physical changes to safety related 

structures, systems, or components, or the 

manner in which they are operated, 

maintained, modified, tested, or inspected. 

The proposed change does not degrade the 

performance or increase the challenges of any 

safety related SSCs assumed to function in 

the accident analysis. Since the backup 

manual CO2 fire suppression system is not 

credited in the safe shutdown analysis to 

protect the upper cable spreading rooms, the 

proposed change does not impact plant safety 

since the conclusions of the fire safe 

shutdown analysis remain unchanged. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The Nuclear Regulatory Commission 

(NRC) staff has reviewed the licensee’s 

analysis and, based on this review, it 

appears that the three standards of 10 

CFR 50.92(c) are satisfied. Therefore, the 

NRC staff proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Russell Gibbs. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Steam Electric Station, 

Units 1 and 2, Somervell County, Texas 

Date of amendment request: April 1, 

2009. 

Brief description of amendments: The 

proposed amendment would delete 

Technical Specification (TS) 5.2.2.d, in 

TS 5.2.2, ‘‘Unit Staff,’’ 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.d 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. In its application 

dated April 1, 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 adopted 

by the licensee, is presented below: 

Criterion 1: The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change removes TS 

restrictions on working hours for personnel 

who perform safety related functions. The TS 

restrictions are superseded by the worker 

fatigue requirements in 10 CFR 26. Removal 

of the TS requirements will be performed 

concurrently with the implementation of the 

10 CFR 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 

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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 TS 

restrictions on working hours for personnel 

who perform safety related functions. The TS 

restrictions are superseded by the worker 

fatigue requirements in 10 CFR 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 TS 

restrictions on working hours for personnel 

who perform safety related functions. The TS 

restrictions are superseded by the worker 

fatigue requirements in 10 CFR 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 TS 

administrative requirements will not reduce 

a margin of safety because the requirements 

in 10 CFR 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: Timothy P. 

Matthews, Esq., Morgan, Lewis and 

Bockius, 1800 M Street, NW., 

Washington, DC 20036. 

NRC Branch Chief: Michael T. 

Markley. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Steam Electric Station, 

Units 1 and 2, Somervell County, Texas 

Date of amendment request: April 2, 

2009. 

Brief description of amendments: The 

amendment revises Technical 

Specification (TS) 3.3.1, ‘‘Reactor Trip 

System (RTS) Instrumentation,’’ to add 

Surveillance Requirement (SR) 3.3.1.16 

to Function 3 of TS Table 3.3.1–1. SR 

3.3.1.16 requires that RTS RESPONSE 

TIMES be verified to be within limits 

every 18 months on a STAGGERED 

TEST BASIS. Function 3 is the power 

range neutron flux—high positive rate 

reactor trip function (hereafter referred 

to as the positive flux rate trip (PFRT) 

function). This change is based on a 

reanalysis of the Rod Cluster Control 

Assembly Bank Withdrawal at Power 

event. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change imposes additional 

surveillance requirements to assure safety 

related structures, systems, and components 

are verified to be consistent with the safety 

analysis and licensing basis. In this specific 

case, a response time verification 

requirement will be added to the positive 

flux rate trip (PFRT) function. 

Overall protection system performance will 

remain within the bounds of the accident 

analysis since there are no hardware changes. 

The design of the Reactor Trip System (RTS) 

instrumentation, specifically the positive flux 

rate trip (PFRT) function, will be unaffected. 

The reactor protection system will continue 

to function in a manner consistent with the 

plant design basis. All design, material, and 

construction standards that were applicable 

prior to the request are maintained. 

The proposed changes will not modify any 

system interface. The proposed changes will 

not affect the probability of any event 

initiators. There will be no degradation in the 

performance of or an increase in the number 

of challenges imposed on safety-related 

equipment assumed to function during an 

accident situation. There will be no change 

to normal plant operating parameters or 

accident mitigation performance. The 

proposed changes will not alter any 

assumptions or change any mitigation actions 

in the radiological consequences evaluations 

in the updated Final Safety Analysis Report 

(FSAR). 

The proposed changes do not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, or 

configuration of the facility or the manner in 

which the plant is operated and maintained. 

The proposed changes do 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 changes do 

not affect the source term, containment 

isolation, or radiological release assumptions 

used in evaluating the radiological 

consequences of an accident previously 

evaluated. The proposed changes are 

consistent with safety analysis assumptions 

and resultant consequences. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change imposes additional 

surveillance requirements to assure safety 

related structures, systems, and components 

are verified to be consistent with the safety 

analysis and licensing basis. 

There are no hardware changes nor are 

there any changes in the method by which 

any safety related plant system performs its 

safety function. This change will not affect 

the normal method of plant operation or 

change any operating parameters. No 

performance requirements will be affected; 

however, the proposed change does impose 

additional surveillance requirements. The 

additional requirements are consistent with 

assumptions made in the safety analysis and 

licensing basis. 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures are introduced as a result of 

these changes. There will be no adverse effect 

or challenges imposed on any safety-related 

system as a result of these changes. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change imposes additional 

surveillance requirements to assure safety 

related structures, systems, and components 

are verified to be consistent with the safety 

analysis and licensing basis. 

The proposed changes do not affect the 

acceptance criteria for any analyzed event. 

The margin of safety is affected in that in the 

new analyses of the Rod (Bank) Withdrawal 

at Power analyses, it is necessary to credit a 

previously uncredited reactor trip function in 

an analysis. However, that reactor trip 

function is described in the plant Technical 

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Specifications with well-defined operability 

requirements. An additional attribute, 

specifically the channel response time 

verification on, a periodic frequency, 

provides additional assurance that the trip 

function performs as credited in the accident 

analysis. With the credit for this reactor trip 

function, all relevant event acceptance 

criteria continue to be met. None of the event 

acceptance limits are exceeded, and none of 

the event acceptance limits are revised by the 

proposed activity. There is no effect on the 

manner in which safety limits, limiting safety 

system settings, or limiting conditions for 

operation are determined nor is there any 

effect on those plant systems necessary to 

assure the accomplishment of protection 

functions. There is no impact on the 

overpower limit, the minimum departure 

from nucleate boiling ratio limit, the radial 

and axial peaking factor limits, the loss of 

coolant accident (LOCA) peak clad 

temperature limit, nor any other limit which, 

in whole or in part, defines a margin of 

safety. The radiological dose consequence 

acceptance criteria listed in the Standard 

Review Plan will continue to be met. 

Therefore the proposed change does not 

involve a 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: Timothy P. 

Matthews, Esq., Morgan, Lewis and 

Bockius, 1800 M Street, NW., 

Washington, DC 20036. 

NRC Branch Chief: Michael T. 

Markley. 

Nine Mile Point Nuclear Station, LLC, 

(NMPNS) Docket No. 50–220, Nine Mile 

Point Nuclear Station Unit No. 1 (NMP 

1), Oswego County, New York 

Date of amendment request: March 3, 

2009. 

Description of amendment request: 

The proposed amendment would 

modify Technical Specification (TS) 

Section 3.2.1, ‘‘Reactor Vessel Heatup 

and Cooldown Rates,’’ and Section 

3.2.2, ‘‘Minimum Reactor Vessel 

Temperature for Pressurization,’’ by 

replacing the existing reactor vessel 

heatup and cooldown rate limits and the 

pressure and temperature limit curves 

with references to the Pressure and 

Temperature Limits Report (PTLR). In 

addition, a new definition for the PTLR 

would be added to TS Section 1.0, 

‘‘Definitions,’’ and a new section 

addressing administrative requirements 

for the PTLR would be added to TS 

Section 6.0, ‘‘Administrative Controls.’’ 

The proposed changes are consistent 

with the guidance in Generic Letter 96– 

03, ‘‘Relocation of the Pressure 

Temperature Limit Curves and Low 

Temperature Overpressure Protection 

System Limits,’’ as supplemented by TS 

Task Force (TSTF) traveler TSTF–419– 

A, ‘‘Revise PTLR Definition and 

References in ISTS 5.6.6, RCS [Reactor 

Coolant System] PTLR.’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes modify the TS by 

replacing references to existing reactor vessel 

heatup and cooldown rate limits and P–T 

[pressure-temperature] limit curves with 

references to the PTLR. The proposed 

amendment also adopts the NRC-approved 

methodology of SIR–05–044–A for the 

preparation of NMP1 P–T limit curves. In 10 

CFR 50 Appendix G, requirements are 

established to protect the integrity of the 

reactor coolant pressure boundary (RCPB) in 

nuclear power plants. Implementing the 

NRC-approved methodology for calculating 

P–T limit curves and relocating those curves 

to the PTLR provide an equivalent level of 

assurance that RCPB integrity will be 

maintained, as specified in 10 CFR 50 

Appendix G. 

The proposed changes do not adversely 

affect accident initiators or precursors, and 

do not alter the design assumptions, 

conditions, or configuration of the plant or 

the manner in which the plant is operated 

and maintained. The ability of structures, 

systems, and components to perform their 

intended safety function is not altered or 

prevented by the proposed changes, and the 

assumptions used in determining the 

radiological consequences of previously 

evaluated accidents are not affected. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The change in methodology for calculating 

P–T limits and the relocation of those limits 

to the PTLR do not alter or involve any 

design basis accident initiators. RCPB 

integrity will continue to be maintained in 

accordance with 10 CFR 50 Appendix G, and 

the assumed accident performance of plant 

structures, systems and components will not 

be affected. These changes do not involve 

any physical alteration of the plant (i.e., no 

new or different type of equipment will be 

installed), and installed equipment is not 

being operated in a new or different manner. 

Thus, no new failure modes are introduced. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes do not affect the 

function of the RCPB or its response during 

plant transients. By calculating the P–T 

limits using NRC-approved methodology, 

adequate margins of safety relating to RCPB 

integrity are maintained. The proposed 

changes do not alter the manner in which 

safety limits, limiting safety system settings, 

or limiting conditions for operation are 

determined, there are no changes to the 

setpoints at which actions are initiated, and 

the operability requirements for equipment 

assumed to operate for accident mitigation 

are not affected. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mark J. 

Wetterhahn, Esquire, Winston & Strawn, 

1700 K Street, NW., Washington, DC 

20006. 

NRC Acting Branch Chief: John P. 

Boska. 

Nine Mile Point Nuclear Station, LLC, 

(NMPNS) Docket No. 50–410, Nine Mile 

Point Nuclear Station Unit No. 2 (NMP 

2), Oswego County, New York 

Date of amendment request: March 9, 

2009. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specification (TS) testing 

frequency for the surveillance 

requirement (SR) in TS 3.1.4, ‘‘Control 

Rod Scram Times.’’ Specifically, the 

proposed change is based on TS Task 

Force (TSTF) change traveler TSTF– 

460–A, Revision 0, and extends the 

frequency for testing control rod scram 

time testing in SR 3.1.4.2 from every 120 

days of cumulative Mode 1 operation to 

200 days of cumulative Mode 1 

operation. A notice of availability of this 

proposed TS change using the 

consolidated line item improvement 

process was published in the Federal 

Register on August 23, 2004 (69 FR 

51864). The licensee affirmed the 

applicability of the model no significant 

hazards consideration determination in 

its application. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

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23448 Federal Register / Vol. 74, No. 95 / Tuesday, May 19, 2009 / Notices 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change extends the 

frequency for testing control rod scram time 

testing from every 120 days of cumulative 

Mode 1 operation to 200 days of cumulative 

Mode 1 operation. The frequency of 

surveillance testing is not an initiator of any 

accident previously evaluated. The frequency 

of surveillance testing does not affect the 

ability to mitigate any accident previously 

evaluated, as the tested component is still 

required to be operable. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change extends the 

frequency for testing control rod scram time 

testing from every 120 days of cumulative 

Mode 1 operation to 200 days of cumulative 

Mode 1 operation. The proposed change does 

not result in any new or different modes of 

plant operation. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

4. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change extends the 

frequency for testing control rod scram time 

testing from every 120 days of cumulative 

Mode 1 operation to 200 days of cumulative 

Mode 1 operation. The proposed change 

continues to test the control rod scram time 

to ensure the assumptions in the safety 

analysis are protected. 

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 Acting Branch Chief: John P. 

Boska. 

Northern States Power Company, 

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

Island Nuclear Generating Plant, Units 

1 and 2 (PINGP), Goodhue County, 

Minnesota 

Date of amendment request: March 5, 

2009, as supplemented by letter dated 

April 13, 2009. 

Description of amendment request: 

The proposed amendments would make 

changes to the PINGP Technical 

Specifications (TSs) to revise TS 3.8.1, 

‘‘AC Sources—Operating,’’ Surveillance 

Requirement (SR) 3.8.1.8 Frequency to 

allow use of the SR 3.0.2 interval 

extension (1.25 times the specified 24 

month Frequency). This would be an 

exception to the SR 3.0.2 limitations in 

the PINGP TS, which do not allow use 

of the interval extension for SRs with a 

24 month Frequency. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This license amendment request proposes 

to add a Frequency Note to Surveillance 

Requirement 3.8.1.8 which will allow 

application of the Surveillance Requirement 

3.0.2 interval extension (1.25 times the 

specified 24 month Frequency) for 

performance of this surveillance. This would 

be an exception to the limitations specified 

in the Prairie Island Nuclear Generating Plant 

Technical Specification Surveillance 

Requirement 3.0.2 for Surveillance 

Requirements with a 24 month Frequency 

and would allow an interval up to 30 months 

for performance of the surveillance. 

The emergency diesel generators are not 

accident initiators and therefore, these 

changes do not involve a significant increase 

[in] the probability of an accident. 

Failure of the bypass relay, by itself, does 

not prevent an emergency diesel generator 

from performing its safety related functions. 

Since the accident analyses only require one 

of the two trains of onsite emergency AC to 

be operable, the changes proposed in the 

license amendment request do not involve a 

significant increase in the consequences of an 

accident. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

This license amendment request proposes 

to add a Frequency Note to Surveillance 

Requirement 3.8.1.8 which will allow 

application of the Surveillance Requirement 

3.0.2 interval extension (1.25 times the 

specified 24 month Frequency) for 

performance of this surveillance. This would 

be an exception to the limitations specified 

in the Prairie Island Nuclear Generating Plant 

Technical Specification Surveillance 

Requirement 3.0.2 for Surveillance 

Requirements with a 24 month Frequency 

and would allow an interval up to 30 months 

for performance of the surveillance. 

The changes proposed for the emergency 

diesel generators do not change any system 

operations or maintenance activities. Testing 

requirements will be revised and will 

continue to demonstrate that the Limiting 

Conditions for Operation are met and the 

system components are functional. The 

revised test Frequency does not create new 

failure modes or mechanisms and no new 

accident precursors are generated. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

This license amendment request proposes 

to add a Frequency Note to Surveillance 

Requirement 3.8.1.8 which will allow 

application of the Surveillance Requirement 

3.0.2 interval extension (1.25 times the 

specified 24 month Frequency) for 

performance of this surveillance. This would 

be an exception to the limitations specified 

in the Prairie Island Nuclear Generating Plant 

Technical Specification Surveillance 

Requirement 3.0.2 for Surveillance 

Requirements with a 24 month Frequency 

and would allow an interval up to 30 months 

for performance of the surveillance. 

The proposed change will continue to 

ensure that the DG trips bypass function 

operates as designed. The functionality and 

operability of the emergency power system is 

not being changed. Since the requested 

change only allows extension of the relay 

testing interval and failure of the relay by 

itself does not prevent the diesel from 

performing its safety function, this change 

does not involve a significant reduction in a 

margin of safety. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment requests involve no 

significant hazards consideration. 

Attorney for licensee: Peter M. Glass, 

Assistant General Counsel, Xcel Energy 

Services, Inc., 414 Nicollet Mall, 

Minneapolis, MN 55401. 

NRC Branch Chief: Lois M. James. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–348 and 50–364, 

Joseph M. Farley Nuclear Plant, Units 1 

and 2, Houston County, Alabama 

Date of amendment request: March 

30, 2009. 

Description of amendment request: 

The proposed amendment revises the 

Technical Specifications (TS), 

Appendix A to Facility Operating 

License Nos. NPF–2 and NPF–8 for the 

Joseph M. Farley Nuclear Plant, Units 1 

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Federal Register / Vol. 74, No. 95 / Tuesday, May 19, 2009 / Notices 23449 

and 2, respectively. The changes would 

eliminate the Reactor Coolant Pump 

(RCP) Breaker Position reactor trip. The 

changes will allow the elimination of a 

trip circuitry that is susceptible to single 

failure vulnerabilities which can result 

in unwarranted reactor trips. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes do not significantly 

increase the probability or consequences of 

an accident previously evaluated in the Final 

Safety Analysis Report (FSAR). All of the 

safety analyses have been evaluated for 

impact. The elimination of Reactor Coolant 

Pump Breaker Position reactor trip will not 

initiate any accident; therefore, the 

probability of an accident has not been 

increased. An evaluation of dose 

consequences, with respect to the proposed 

changes, indicates there is no impact due to 

the proposed changes and all acceptance 

criteria continue to be met. Therefore, these 

changes do not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

The proposed changes do not create the 

possibility of a new or different kind of 

accident than any accident already evaluated 

in the FSAR. No new accident scenarios, 

failure mechanisms or limiting single failures 

are introduced as a result of the proposed 

changes. The changes have no adverse effects 

on any safety-related system. Therefore, all 

accident analyses criteria continue to be met 

and these changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes do not involve a 

significant reduction in a margin of safety. 

All analyses that credit the Reactor Coolant 

System Low Flow reactor trip function have 

been reviewed and no changes to any inputs 

are required. The evaluation demonstrated 

that all applicable acceptance criteria are 

met. Therefore, the proposed changes do not 

involve a significant reduction in the margin 

of safety. 

Based on the preceding evaluation, SNC 

has determined that the proposed changes 

meet the requirements of 10 CFR 50.92(c) and 

do not involve a significant hazards 

consideration. 

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: M. Stanford 

Blanton, Esq., Balch and Bingham, Post 

Office Box 306, 1710 Sixth Avenue 

North, Birmingham, Alabama 35201. 

NRC Branch Chief: Melanie C. Wong. 

Tennessee Valley Authority (TVA), 

Docket No. 50 390, Watts Bar Nuclear 

Plant, Unit 1, Rhea County, Tennessee 

Date of amendment request: April 30, 

2009. 

Description of amendment request: 

The proposed amendment would revise 

technical specification (TS) Section 5.7, 

‘‘Procedures, Programs, and Manuals,’’ 

to correct typographical errors 

introduced in Amendment No. 70, dated 

October 8, 2008. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. This change is limited only 

to correcting a typographical error in a 

section number (5.7.2.20 versus 5.2.7.20) 

contained in Technical Specification Section 

5.0, which will not change the intent of the 

added section previously approved in 

License Amendment 70. Therefore, no 

increase in the probability or consequences 

of an accident previously evaluated has been 

created. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. This change is limited only 

to correcting a typographical error in a 

section number (5.7.2.20 versus 5.2.7.20) 

contained in Technical Specification Section 

5.0, which will not change the intent of the 

added section previously approved in 

License Amendment 70. Therefore, the 

possibility of a new or different kind of 

accident from those previously analyzed has 

not been created. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. This change is limited only 

to correcting a typographical error in a 

section number (5.7.2.20 versus 5.2.7.20) 

contained in Technical Specification Section 

5.0, which will not change the intent of the 

added section previously approved in 

License Amendment 70. Therefore, the 

proposed change does not involve a 

significant reduction in a margin of safety. 

Based on the above, TVA concludes that 

the proposed amendment presents no 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Branch Chief: L. Raghavan. 

Virginia Electric and Power Company, 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Unit Nos. 1 and 2, 

Louisa County, Virginia 

Date of amendment request: March 

26, 2009. 

Description of amendment request: 

The proposed amendments would 

increase each unit’s rated thermal power 

(RTP) level from 2893 megawatts 

thermal (MWt) to 2940 MWt, and make 

technical specification changes as 

necessary to support operation at the 

uprated power level. The proposed 

change is an increase in RTP of 

approximately 1.6 percent. 

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 will increase the 

North Anna Power Station (NAPS) Units 1 

and 2 rated thermal power (RTP) from 2893 

megawatts thermal (MWt) to 2940 MWt. 

Nuclear steam supply systems and balanceof-plant systems, components and analyses 

that could be affected by the proposed 

change to the RTP were evaluated using 

revised design parameters. The evaluations 

determined that these structures, systems and 

components are capable of performing their 

design function at the proposed uprated RTP 

of 2940 MWt. An evaluation of the accident 

analyses demonstrates that the applicable 

analysis acceptance criteria are still met with 

the proposed changes. Power level is an 

input assumption to equipment design and 

accident analyses, but it is not a transient or 

accident initiator. Accident initiators are not 

affected by the power uprate, and plant safety 

barrier challenges are not created by the 

proposed changes. 

The radiological consequences of operation 

at the uprated power conditions have been 

assessed. The proposed change to RTP does 

not affect release paths, frequency of release, 

or the analyzed source term for any accidents 

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23450 Federal Register / Vol. 74, No. 95 / Tuesday, May 19, 2009 / Notices 

previously evaluated in the NAPS Updated 

Final Safety Analysis Report. Structures, 

systems and components required to mitigate 

transients are capable of performing their 

design functions with the proposed changes, 

and are thus acceptable. Analyses performed 

to assess the effects of mass and energy 

releases remain valid. The source term used 

to assess radiological consequences was 

reviewed and determined to bound operation 

at the proposed power level. 

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 new accident scenarios, failure 

mechanisms, or single failures are introduced 

as a result of any proposed changes. The 

UFM has been analyzed, and system failures 

will not adversely affect any safety-related 

system or any structures, systems or 

components required for transient mitigation. 

Structures, systems and components 

previously required for transient mitigation 

are still capable of fulfilling their intended 

design functions. The proposed changes have 

no significant adverse affect on any safetyrelated structures, systems or components 

and do not significantly change the 

performance or integrity of any safety-related 

system. 

The proposed changes do not adversely 

affect any current system interfaces or create 

any new interfaces that could result in an 

accident or malfunction of a different kind 

than previously evaluated. Operating at RTP 

of 2940 MWt does not create any new 

accident initiators or precursors. Credible 

malfunctions are bounded by the current 

accident analyses of record or recent 

evaluations demonstrating that applicable 

criteria are still met with the proposed 

changes. 

Therefore, this change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The margins of safety associated with the 

power uprate are those pertaining to core 

thermal power. These include fuel cladding, 

reactor coolant system pressure boundary, 

and containment barriers. Core analyses 

demonstrate that power uprate 

implementation will continue to meet the 

current nuclear design basis. Impacts to 

components associated with the reactor 

coolant system pressure boundary structural 

integrity, and factors such as pressuretemperature limits, vessel fluence, and 

pressurized thermal shock were determined 

to be bounded by the current analyses. 

Systems will continue to operate within 

their design parameters and remain capable 

of performing their intended safety functions 

following implementation of the proposed 

change. The current NAPS safety analyses, 

including the design basis radiological 

accident dose calculations, bound the power 

uprate. 

Therefore, this 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: Lillian M. 

Cuoco, Esq., Senior Counsel, Dominion 

Resources Services, Inc., 120 Tredegar 

Street, RS–2, Richmond, VA 23219. 

NRC Branch Chief: Melanie C. Wong. 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

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

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

Systems (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1 (800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr.resource@nrc.gov. 

Dominion Energy Kewaunee, Inc. Docket 

No. 50–305, Kewaunee Power Station 

(KPS), Kewaunee County, Wisconsin 

Date of application for amendment: 

September 11, 2008, as supplemented 

by letter dated December 17, 2008, and 

January 20, 2009. 

Brief Description of amendment: The 

amendment revised the Technical 

Specifications, extending the 15-year 

interval between containment Type A 

tests specified by Specification 4.4.a, 

‘‘Integrated Leak Rate Test,’’ by 6 

months. The current Type A test 

interval expires at the end of April 2009. 

The amendment extends this interval, 

on a one-time basis, to October 2009 to 

coincide with completion of the next 

scheduled refueling outage. 

Date of issuance: April 27, 2009. 

Effective date: As of the date of 

issuance and should be implemented 

within 60 days. 

Amendment No.: 204. 

Facility Operating License No. DPR– 

43: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: November 4, 2008 (73 FR 

65689). The commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

April 27, 2009. 

No significant hazards consideration 

comments received: No. 

Dominion Energy Kewaunee, Inc. Docket 

No. 50–305, Kewaunee Power Station 

(KPS), Kewaunee County, Wisconsin 

Date of application for amendment: 

July 7, 2008, as supplemented on 

September 19, 2008, and March 17, 

2009. 

Brief description of amendment: The 

amendment revised the licensing basis, 

authorizing the licensee to use the 

methodology conveyed in the licensee’s 

letters cited above to determine the 

seismic loads on the recently upgraded 

Auxiliary Building crane. The 

authorization is conveyed by addition of 

a new License Condition 2.C.(11) to 

Facility Operating License DPR–43. 

Date of issuance: April 30, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 205. 

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Facility Operating License No. DPR– 

43: The amendment revised Facility 

Operating License No. DPR–43. 

Date of initial notice in Federal 

Register: August 26, 2008 (73 FR 

50358). The Commission’s related 

evaluation of the amendment is 

contained in a safety evaluation dated 

April 30, 2009. 

No Significant hazards consideration 

comments received: No. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of application for amendment: 

July 16, 2008, as supplemented by 

letters dated January 2 and March 19, 

2009. 

Brief description of amendment: The 

amendment revised Technical 

Specifications 3.1.4, ‘‘Control Rod 

Scram Times,’’ 3.2.2, ‘‘Minimum 

Critical Power Ratio (MCPR),’’ and 5.6.3, 

‘‘Core Operating Limits Report (COLR),’’ 

to allow incorporation of the analytical 

methodologies associated with 

operation of Global Nuclear FuelAmericas (GNF) fuel into the licensing 

basis to support transition to GNF GE14 

fuel. 

Date of issuance: May 5, 2009. 

Effective date: As of its date of 

issuance and shall be implemented 

prior to beginning operating cycle 20. 

Amendment No.: 211. 

Facility Operating License No. NPF– 

21: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: October 14, 2008 (73 FR 

60729). 

The supplements dated January 2 and 

March 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 as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated May 5, 2009. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50– 

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: 

September 18, 2008, as supplemented 

by letter dated February 4, 2009. 

Brief description of amendment: The 

amendment modified Technical 

Specification (TS) requirements for 

inoperable snubbers by relocating the 

current TS 3.7.8, ‘‘Snubbers,’’ to the 

Technical Requirements Manual and 

adding Limiting Condition for 

Operation (LCO) 3.0.8. The amendment 

also made conforming changes to TS 

LCO 3.0.1. The proposed amendment is 

consistent with U.S. Nuclear Regulatory 

Commission (NRC)-approved Technical 

Specification Task Force (TSTF) 

Improved Standard Technical 

Specifications Change Traveler, TSTF– 

372, Revision 4, ‘‘Addition of LCO 3.0.8, 

Inoperability of Snubbers,’’ as part of 

the consolidated line item improvement 

process. 

Date of issuance: May 1, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 60 

days from the date of issuance. 

Amendment No.: 219. 

Facility Operating License No. NPF– 

38: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: December 16, 2008 (73 FR 

76410). The supplemental letter dated 

February 4, 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 May 1, 2009. 

No significant hazards consideration 

comments received: No. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–440, 

Perry Nuclear Power Plant, Unit No. 1, 

Lake County, Ohio 

Date of application for amendment: 

November 18, 2008. 

Brief description of amendment: This 

amendment modifies Technical 

Specification 5.5.6 to incorporate 

Technical Specification Task Force 

(TSTF) Travelers TSTF–479, ‘‘Changes 

to Reflect Revision of 10 CFR [Code of 

Federal Regulations] 50.55a,’’ and 

TSTF–497, ‘‘Limit Inservice Testing 

Program SR [Surveillance Requirement] 

3.0.2 Application to Frequencies of 2 

Years or Less.’’ 

Date of issuance: May 1, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 151. 

Facility Operating License No. NPF– 

58: This amendment revised the 

Technical Specifications and License. 

Date of initial notice in Federal 

Register: January 27, 2009 (74 FR 

4772). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

May 1, 2009. 

No significant hazards consideration 

comments received: No. 

GPU Nuclear, Inc., Docket No. 50–320, 

Three Mile Island Nuclear Station, Unit 

2, Dauphin County, Pennsylvania 

Date of amendment request: June 11, 

2008, as supplemented by letters dated 

September 15, 2008, December 10, 2008, 

and March 16, 2009. 

Brief description of amendment: The 

amendment deletes Technical 

Specification 6.5, which provided the 

requirements related to review and 

audit functions. 

Date of issuance: May 1, 2009. 

Effective date: May 1, 2009. 

Amendment No.: 63. 

Possession Only License No. DPR–73: 

The amendment revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 26, 2008 (73 FR 

50356) The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation Report, 

dated May 1, 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: 

June 27, 2007, as supplemented on 

April 28, September 4, and December 

17, 2008. 

Brief description of amendment: The 

amendment revises surveillance 

requirements in Technical 

Specifications (TS) Section 3.8.1, ‘‘AC 

Sources—Operating,’’ associated with 

the diesel generator (DG) steady-state 

frequency and voltage. The amendment 

corrects non-conservative TS frequency 

and voltage values, which the licensee 

states have the potential to result in 

undesirable effects such as centrifugal 

charging pump motor brake horsepower 

exceeding its nameplate maximum 

horsepower, and subsequently 

overloading the DGs. 

Date of issuance: April 30, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 45 days from the date of issuance 

April 30, 2009. 

Amendment Nos.: 309 (CNP–1), 291 

(CNP–2). 

Facility Operating License Nos. DPR– 

58 and DPR–74: Amendment revises the 

Renewed Operating License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: November 4, 2008 (73 FR 

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65696). The April 28 and December 17, 

2008 supplements provided additional 

information that clarified the 

application, but did not expand the 

scope of the application as originally 

noticed, and did not change the staff’s 

original proposed significant hazards 

consideration published in the Federal 

Register on August 14, 2007. 

The September 4, 2008 supplement 

provided additional information which 

expanded the scope of the application 

as originally noticed. The NRC staff 

identified that the specified DG voltage 

of 3,740 volts at 10 seconds after the DG 

start was non-conservative and 

inconsistent with the 3,910 volt 

minimum steady-state voltage provided 

in other parts of TS Section 3.8.1. The 

licensee proposed additional changes to 

TS Section 3.8.1 in its September 4, 

2008 letter. The NRC staff determined 

that the proposed expanded scope of the 

amendment involved a proposed no 

significant hazards consideration as 

published in the Federal Register on 

November 4, 2008 (73 FR 65696). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated April 30, 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: 

February 24, 2009. 

Brief description of amendments: The 

amendments deleted the requirement 

for the power range neutron flux ratehigh negative rate trip (Function 3.b) in 

Technical Specification (TS) Table 

3.3.1–1, ‘‘Reactor Trip System 

Instrumentation.’’ The changes are 

consistent with the NRC-approved 

methodology presented in 

Westinghouse Topical Report, WCAP– 

11394–P–A, ‘‘Methodology for the 

Analysis of the Dropped Rod Event,’’ 

dated January 1990. The amendments 

also incorporated editorial changes to 

reflect the deletion of Function 3.b in 

TS Table 3.3.1–1. 

Date of issuance: April 29, 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—205; Unit 

2—206. 

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: March 24, 2009 (74 FR 

12394). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated April 29, 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: 

December 31, 2008 superseded the 

application dated August 1, 2008, as 

supplemented by letters dated 

November 25 and December 31, 2008. 

Brief description of amendment: The 

amendment revised WBN Unit 1 

Technical Specification (TS) 4.2.1, 

‘‘Fuel Assemblies,’’ and TS surveillance 

requirements (SRs) 3.5.1.4, 

‘‘Accumulators,’’ and 3.5.4.3, ‘‘RWST 

[Refueling Water Storage Tank],’’ to 

increase the maximum number of 

tritium producing burnable absorber 

rods from 400 to 704. 

Date of issuance: April 30, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 180 days of issuance. 

Amendment No.: 77. 

Facility Operating License No. NPF– 

90: Amendment revises the TS 4.2.1 and 

TS SRs 3.5.1.4 and 3.5.4.3. 

Date of initial notice in Federal 

Register: Originally November 12, 2008 

(73 FR 66946) was superseded by a 

notice on January 27, 2009 (74 FR 4776). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated April 30, 2009. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 7th day 

of May 2009. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E9–11268 Filed 5–18–09; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket No. 70–1113; NRC–2009–0209] 

Notice of Availability of Environmental 

Assessment and Finding of No 

Significant Impact for License Renewal 

for Global Nuclear Fuel—Americas, 

LLC, Wilmington, NC 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Notice of availability. 

FOR FURTHER INFORMATION CONTACT: 

Mary Adams, Senior Project Manager, 

Fuel Manufacturing Branch, Division of 

Fuel Cycle Safety and Safeguards, Office 

of Nuclear Material Safety and 

Safeguards, U.S. Nuclear Regulatory 

Commission, Rockville, Maryland 

20852. Telephone: (301) 492–3113; Fax: 

(301) 492–3363; e-mail: 

Mary.Adams@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Introduction 

The Nuclear Regulatory Commission 

(NRC) is considering the renewal of 

Special Nuclear Material License SNM– 

1097 for the continued operation of the 

Global Nuclear Fuel—Americas, LLC 

(GNF–A) Fuel Fabrication Facility 

located in Wilmington, North Carolina. 

This renewal authorizes the licensee to 

receive and possess nuclear materials at 

the Wilmington facility to fabricate and 

assemble nuclear fuel components 

under the provisions of 10 CFR Part 70, 

Domestic Licensing of Special Nuclear 

Material. If NRC approves the renewal 

of the license, the term would cover 40 

years. NRC has prepared an 

environmental assessment (EA) in 

support of this action in accordance 

with the requirements of 10 CFR Part 

51. Based on the EA, the NRC has 

concluded that a finding of no 

significant impact is appropriate. If 

approved, NRC will issue the renewed 

license following the publication of this 

Notice. 

II. EA Summary 

The licensee requests approval to 

renew SNM–1097 for an additional 40 

years at the Wilmington, North Carolina 

facility. Specifically, this would allow 

GNF–A to continue manufacturing and 

assembling nuclear fuel components for 

use in commercial light-water-cooled 

nuclear power reactors. GNF–A’s 

request for the renewal was previously 

noticed in the Federal Register on June 

18, 2007 (72 FR 33539), with an 

opportunity to request a hearing. No 

hearing requests were received. 

The staff has prepared the EA in 

support of the proposed license 

renewal. Staff considered direct, 

indirect, and cumulative environmental 

impacts to 12 resource areas in their 

evaluation, including: land use; 

transportation; socioeconomics; air 

quality; water quality; geology and soils; 

ecology; noise; historic and cultural; 

scenic and visual; public and 

occupational health; and waste 

management. All of the environmental 

impacts were small-to-moderate. The 

license renewal request does not require 

altering the site footprint nor does it 

mstockstill on PROD1PC66 with NOTICES