Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 15765 

Signed at Washington, DC, this 20th day of 

March 2009. 

Linda G. Poole, 

Certifying Officer, Division of Trade 

Adjustment Assistance. 

[FR Doc. E9–7747 Filed 4–6–09; 8:45 am] 

BILLING CODE 4510–FN–P 

MORRIS K. UDALL SCHOLARSHIP 

AND EXCELLENCE IN NATIONAL 

ENVIRONMENTAL POLICY 

FOUNDATION 

Sunshine Act Meetings 

TIME AND DATE: 9 a.m. to 12 p.m., Friday, 

April 17, 2009. 

PLACE: The University of Arizona 

Foundation’s Vine Street Annex, Room 

102, 1125 N. Vine Street, Tucson, AZ 

85719. 

STATUS: This meeting will be open to the 

public, unless it is necessary for the 

Board to consider items in executive 

session. 

MATTERS TO BE CONSIDERED: (1) A report 

on the U.S. Institute for Environmental 

Conflict Resolution; (2) A report from 

the Udall Center for Studies in Public 

Policy; (3) A report on the Native 

Nations Institute; (4) Program Reports; 

and (5) A Report from the Management 

Committee. 

PORTIONS OPEN TO THE PUBLIC: All 

sessions with the exception of the 

session listed below. 

PORTIONS CLOSED TO THE PUBLIC: 

Executive session. 

CONTACT PERSON FOR MORE INFORMATION: 

Ellen K. Wheeler, Executive Director, 

130 South Scott Avenue, Tucson, AZ 

85701, (520) 901–8500. 

Dated: March 31, 2009. 

Ellen K. Wheeler, 

Executive Director, Morris K. Udall 

Scholarship and Excellence in National 

Environmental Policy Foundation, and 

Federal Register Liaison Officer. 

[FR Doc. E9–7602 Filed 4–6–09; 8:45 am] 

BILLING CODE 6820–FN–M 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2009–0148] 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a. (2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC 

staff) is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued from March 12, 

2009 to March 25, 2009. The last 

biweekly notice was published on 

March 24, 2009 (74 FR 12390). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

10 CFR 50.92, this means that operation 

of the facility in accordance with the 

proposed amendment would not (1) 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated; or (2) 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated; or (3) 

involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rulemaking and 

Directives Branch, TWB–05–B01M, 

Division of Administrative Services, 

Office of Administration, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001, and should cite the 

publication date and page number of 

this Federal Register notice. Copies of 

written comments received may be 

examined at the Commission’s Public 

Document Room (PDR), located at One 

White Flint North, Public File Area 

O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. 

Within 60 days after the date of 

publication of this notice, any person(s) 

whose interest may be affected by this 

action may file a request for a hearing 

and a petition to intervene with respect 

to issuance of the amendment to the 

subject facility operating license. 

Requests for a hearing and a petition for 

leave to intervene shall be filed in 

accordance with the Commission’s 

‘‘Rules of Practice for Domestic 

Licensing Proceedings’’ in 10 CFR Part 

2. Interested person(s) should consult a 

current copy of 10 CFR 2.309, which is 

available at the Commission’s PDR, 

located at One White Flint North, Public 

File Area 01F21, 11555 Rockville Pike 

(first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed within 60 

days, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

15766 Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also set forth the specific 

contentions which the petitioner/ 

requestor seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the petitioner/requestor shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the petitioner/ 

requestor to relief. A petitioner/ 

requestor who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

All documents filed in NRC 

adjudicatory proceedings, including a 

request for hearing, a petition for leave 

to intervene, any motion or other 

document filed in the proceeding prior 

to the submission of a request for 

hearing or petition to intervene, and 

documents filed by interested 

governmental entities participating 

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

accordance with the NRC E-Filing rule, 

which the NRC promulgated in August 

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

process requires participants to submit 

and serve all adjudicatory documents 

over the internet or in some cases to 

mail copies on electronic storage media. 

Participants may not submit paper 

copies of their filings unless they seek 

a waiver in accordance with the 

procedures described below. 

To comply with the procedural 

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

days prior to the filing deadline, the 

petitioner/requestor must contact the 

Office of the Secretary by e-mail at 

hearingdocket@nrc.gov, or by calling 

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

ID certificate, which allows the 

participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and/or (2) creation of an 

electronic docket for the proceeding 

(even in instances in which the 

petitioner/requestor (or its counsel or 

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

petitioner/requestor will need to 

download the Workplace Forms 

Viewer TM to access the Electronic 

Information Exchange (EIE), a 

component of the E-Filing system. The 

Workplace Forms Viewer TM is free and 

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

Information about applying for a digital 

ID certificate is available on NRC’s 

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

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

Once a petitioner/requestor has 

obtained a digital ID certificate, had a 

docket created, and downloaded the EIE 

viewer, it can then submit a request for 

hearing or petition for leave to 

intervene. Submissions should be in 

Portable Document Format (PDF) in 

accordance with NRC guidance 

available on the NRC public Web site at 

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

complete at the time the filer submits its 

documents through EIE. To be timely, 

an electronic filing must be submitted to 

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

Eastern Time on the due date. Upon 

receipt of a transmission, the E-Filing 

system time-stamps the document and 

sends the submitter an e-mail notice 

confirming receipt of the document. The 

EIE system also distributes an e-mail 

notice that provides access to the 

document to the NRC Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/ 

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically may 

seek assistance through the ‘‘Contact 

Us’’ link located on the NRC Web site 

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

electronic filing Help Desk, which is 

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

Eastern Time, Monday through Friday, 

excluding government holidays. The 

help electronic filing Help Desk can be 

contacted by telephone at 1–866–672– 

7640 or by e-mail at 

MSHD.Resource@nrc.gov. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file a 

motion, in accordance with 10 CFR 

2.302(g), with their initial paper filing 

requesting authorization to continue to 

submit documents in paper format. 

Such filings must be submitted by: (1) 

First class mail addressed to the Office 

of the Secretary of the Commission, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, Attention: 

Rulemaking and Adjudications Staff; or 

(2) courier, express mail, or expedited 

delivery service to the Office of the 

Secretary, Sixteenth Floor, One White 

Flint North, 11555 Rockville Pike, 

Rockville, Maryland 20852, Attention: 

Rulemaking and Adjudications Staff. 

Participants filing a document in this 

manner are responsible for serving the 

document on all other participants. 

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

the mail, or by courier, express mail, or 

expedited delivery service upon 

depositing the document with the 

provider of the service. 

Non-timely requests and/or petitions 

and contentions will not be entertained 

Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 15767 

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. 

Duke Energy Carolinas, LLC, et al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina 

Date of amendment request: 

September 2, 2008. 

Description of amendment request: 

The amendments would revise the 

technical specifications to allow manual 

operation of the containment spray 

system and to revise the upper and 

lower limits of the refueling water 

storage tank. 

Basis for proposed no significant 

hazards consideration determination: 

As required by Title 10 of the Code of 

Federal Regulations (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 Containment Spray System and 

RWST [refueling water storage tank] are 

accident mitigation equipment. As such, 

changes in operation of these systems 

cannot have an impact on the 

probability of an accident. 

The RWST will continue to comply 

with all applicable regulatory 

requirements and design criteria 

following approval of the proposed 

changes (e.g., train separation, 

redundancy, and single failure). The 

water level on the containment floor 

will be higher at the start of transfer to 

the containment sump but will remain 

below the maximum design level 

analyzed for equipment submergence. 

The change in the sump pH will not 

result in a significant increase in 

radiological consequences of a LOCA 

[loss of coolant accident]. Therefore, the 

design functions performed by the 

equipment are not changed. 

The delay in containment spray 

operation will result in an increase in 

containment temperature, containment 

pressure, offsite dose, and control room 

dose during a LOCA or high energy line 

break inside containment. Containment 

analyses have been performed to 

demonstrate that containment pressure 

and temperature remain within the 

design limits and there is no significant 

impact on the environmental 

qualification for equipment inside 

containment. The impact on piping and 

supports is acceptable without 

modification. The reduction in fission 

product removal due to delayed 

containment spray operation does not 

result in exceeding the offsite dose and 

control room dose limits in 10 CFR 

50.67 and 10 CFR Part 50, Appendix A, 

GDC 19. The analysis of the change in 

containment conditions due to a single 

failure of an operating spray pump and 

the suspension of containment spray 

determined that the pressure remained 

below the design limits. 

Regarding the proposed change to 

adopt TSTF–493, Rev. 3 on a limited 

basis, the change clarifies the 

requirements for instrumentation to 

ensure the instrumentation will actuate 

as assumed in the safety analysis. 

Instruments are not an assumed initiator 

of any accident previously evaluated. As 

a result, the proposed change will not 

increase the probability of an accident 

previously evaluated. The proposed 

change will ensure that the instruments 

actuate as assumed to mitigate the 

accidents previously evaluated. As a 

result, the proposed change will not 

increase the consequences of an 

accident previously evaluated. 

Based on this discussion, the 

proposed amendment does not 

significantly increase 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 modifications to install RWST 

narrow range level indication will be 

seismically qualified and isolated from 

the safety related portion of the RWST 

level indication system. As such, the 

new level indication will not create the 

possibility of a new or different kind of 

accident. 

The modification to the low level 

setpoint will not install any new plant 

equipment. The setpoint will continue 

to be included within the engineered 

safeguards features instrumentation and 

monitored according to the applicable 

surveillance requirements. The 

evaluation of the new level setpoint and 

the change in the swapover sequence 

concluded that the equipment aligned to 

the sump will continue to have 

sufficient suction pressure prior to 

containment sump suction swapover. 

The design of the RWST low level 

instrumentation-complies with all 

applicable regulatory requirements and 

design criteria. 

The overall function of the 

Containment Spray System is not 

changed by this proposed amendment. 

The proposed change alters the method 

of controlling the safety system 

following a design basis event so that 

manual actions are substituted for 

automatic actions. Calculations confirm 

that these actions will be taken within 

the appropriate scenario sequence 

timing to provide containment cooling 

and source term reduction with no 

significant increase in radiological 

consequences and without exceeding 

containment design limits. 

Regarding the proposed change to 

adopt TSTF–493, Rev. 3 on a limited 

basis, the change does not involve a 

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

new or different type of equipment will 

be installed) or a change in the methods 

governing normal plant operation. The 

change does not alter assumptions made 

in the safety analysis but ensures that 

the instruments behave as assumed in 

the accident analysis. The proposed 

change is consistent with the safety 

analysis assumptions. 

Therefore, the proposed change does 

not create the possibility of a new or 

15768 Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 

different kind of accident from any 

previously evaluated. 

3. Does the proposed amendment 

involve a significant reduction in the 

margin of safety? 

Response: No. 

The proposed change has the 

potential to increase the radiological 

dose at the site boundary and in the 

control room. However, the calculations 

demonstrate that the dose consequences 

at the site boundary, low population 

zone, and control room remain within 

regulatory acceptance limits. Additional 

analysis concluded: 

• Peak containment pressure for 

analyzed design basis accidents will not 

be significantly increased and 

containment design limits will not be 

exceeded. 

• Assumptions used in the 

environmental qualification of 

equipment exposed to the containment 

atmosphere remain bounding. 

• Pumps aligned to the RWST and to 

the containment sump will have 

adequate suction pressure. 

Regarding the proposed change to 

adopt TSTF–493, Rev. 3 on a limited 

basis, the change clarifies the 

requirements for instrumentation to 

ensure the instrumentation will actuate 

as assumed in the accident analysis. No 

change is made to the accident analysis 

assumptions and no margin of safety is 

reduced as part of this change. 

Therefore, the proposed change does 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ms. Lisa F. 

Vaughn, Associate General Counsel and 

Managing Attorney, Duke Energy 

Carolinas, LLC, 526 South Church 

Street, EC07H, Charlotte, NC 28202. 

NRC Branch Chief: Melanie C. Wong. 

Duke Energy Carolinas, LLC, et al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina 

Date of amendment request: October 

2, 2008. 

Description of amendment request: 

The amendments would revise 

Technical Specifications (TS) associated 

with the verification of ice condenser 

door operability. The proposed 

amendment affects the current TS 

surveillance requirements 3.6.13.5 and 

3.6.13.6. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated? 

Response: No. 

The only analyzed accidents of 

possible consideration in regards to 

changes potentially affecting the ice 

condenser are a loss of coolant accident 

(LOCA) and a high energy line break 

(HELB) inside Containment. However, 

the ice condenser is not postulated as 

being the initiator of any LOCA or 

HELB. This is because it is designed to 

remain functional following a design 

basis earthquake, and the ice condenser 

does not interconnect or interact with 

any systems that interconnect or interact 

with the Reactor Coolant or Main Steam 

Systems. Since these proposed changes 

do not result in, or require, any physical 

change to the ice condenser that could 

introduce an interaction with the 

Reactor Coolant or Main Steam Systems, 

then there can be no change in the 

probability of an accident previously 

evaluated. Regarding consequences of 

analyzed accidents, the ice condenser is 

an engineered safety feature designed, 

in part, to limit the Containment subcompartment and Containment vessel 

pressure immediately following the 

initiation of a LOCA or HELB. 

Conservative sub-compartment and 

Containment pressure analysis shows 

these criteria will be met if the total ice 

mass within the ice bed is maintained 

in accordance with the DBA analysis; 

therefore, the proposed TS [Technical 

Specification] SR [surveillance 

requirement] changes of these 

requirements will not increase the 

consequences of any accident 

previously evaluated. 

Thus, based on the above, 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. 

As previously described, the ice 

condenser is not postulated as being the 

initiator of any design basis accident. 

The proposed changes do not impact 

any plant system, structure or 

component that is an accident initiator. 

The proposed TSs and TS Bases changes 

do not involve any hardware changes to 

the ice condenser or other change that 

could create any new accident 

mechanisms. Therefore, there can be no 

new or different accidents created from 

those already identified and evaluated 

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 and the reactor coolant system 

will not be impacted by the proposed 

changes. The Application provides a 

description of additional subcompartment and Containment pressure 

response analysis that has been 

performed. This analysis demonstrates 

that Containment will remain fully 

capable of performing its design 

function with implementation of the 

proposed changes. Therefore, no safety 

margin will be significantly impacted. 

The changes proposed in this LAR 

[license amendment request] do not 

make any physical alteration to the ice 

condenser doors, nor does it affect the 

required functional capability of the 

doors in any way. The intent of the 

proposed changes to the ice condenser 

door surveillance requirements is to 

eliminate an unnecessary and overly 

restrictive Lower Inlet Door torque 

surveillance test. There will be no 

degradation in the operable status of the 

ice condenser doors and the ability to 

confirm operability for the ice 

condenser doors will be maintained, 

such that the doors will continue to 

fully perform their safety function as 

assumed in the plant’s safety analyses. 

Thus, it can be concluded that the 

proposed TS and TS Bases 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 C. Wong. 

Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 15769 

Duke Energy Carolinas, LLC, et al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina 

Date of amendment request: October 

8, 2008. 

Description of amendment request: 

The amendments would revise the 

Technical Specifications (TSs) by 

removing and updating portions of the 

TSs which are out of date or are obsolete 

including footnotes and references. 

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 are 

administrative in nature and therefore 

they do not involve any change in the 

design, configuration, or operation of 

the nuclear units. All Limiting 

Conditions for Operation, Limiting 

Safety System Settings and Safety 

Limits specified in the Technical 

Specifications remain unchanged. The 

Physical Security and related plans, 

Operator Training and Requalification 

Programs, Quality Assurance Programs, 

and the Emergency Plans will not be 

materially changed by the proposed 

license amendment due to its 

administrative nature. 

The technical qualifications of the 

operating licensee will not be reduced. 

Personnel engaged in operation, 

maintenance, engineering, assessment, 

training, and other related services will 

not be changed. Duke officers and 

executives currently responsible for the 

overall safe operation of the nuclear 

plants are expected to continue in the 

same capacity. 

Therefore, the proposed amendment 

does not involve an increase in the 

probability or consequences of an 

accident previously analyzed. 

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 are 

administrative in nature and therefore 

they do not involve any change in the 

design, configuration, or operation of 

the nuclear plant. The current plant 

safety analyses, therefore, remain 

complete and accurate in addressing the 

design basis events and in analyzing 

plant response and consequences. 

The Limiting Conditions for 

Operations, Limiting Safety System 

Settings and Safety Limits specified in 

the Technical Specifications are not 

affected by the proposed changes. As 

such, the plant conditions for which the 

design basis accident analyses were 

performed remain valid. 

The amendment does not introduce a 

new mode of plant operation or new 

accident precursors, does not involve 

any physical alterations to plant 

configurations or make changes to 

system set points that could initiate a 

new or different kind of accident. 

Therefore, the proposed amendment 

does not create the possibility of a new 

or different kind of accident from any 

accident previously evaluated. 

3. Does the proposed amendment 

involve a significant reduction in a 

margin of safety? 

Response: No. 

The proposed changes are 

administrative in nature and therefore 

they do not involve a change in the 

design, configuration, or operation of 

the nuclear plants. The change does not 

affect either the way in which the plant, 

structures, systems, and components 

perform their safety function or their 

design and licensing bases. 

Plant safety margins are established 

through Limiting Conditions for 

Operation, Limiting Safety System 

Settings and Safety Limits specified in 

the Technical Specifications. Because 

there is no change to the physical design 

of the plant, there is no change to any 

of these margins. 

Therefore, the proposed amendment 

does not involve a significant reduction 

in the margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: 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 C. Wong. 

Duke Energy Carolinas, LLC, et al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina 

Date of amendment request: October 

14, 2008. 

Description of amendment request: 

The amendments would revise the 

Technical Specification [TS] 

Administrative Controls, ‘‘Inservice 

Testing Program,’’ for consistency with 

the requirements of Title 10 of the Code 

of Federal Regulations (10 CFR) 

50.55a(f)(4) for pumps and valves which 

are classified as American Society of 

Mechanical Engineers [ASME] Code 

Class 1, Class 2, and Class 3. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 

10 CFR 50.91(a), the licensee has 

provided its analysis of the issue of no 

significant hazards consideration, which 

is presented below: 

1. Does the proposed change involve 

a significant increase in the probability 

or consequences of an accident 

previously evaluated? 

Response: No. 

The proposed changes revise TS 5.5.8, 

‘‘Inservice Testing Program,’’ for 

consistency with the requirements of 10 

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

testing of pumps and valves which are 

classified as ASME Code Class 1, Class 

2, and Class 3. The proposed changes 

incorporate revisions to the ASME Code 

that result in a net improvement in the 

measures for testing pumps and valves. 

The proposed changes do not impact 

any accident initiators or analyzed 

events or assumed mitigation of 

accident or transient events. The 

proposed change does not involve the 

addition or removal of any equipment, 

or any design changes to the facility. 

Therefore, these proposed changes do 

not involve a significant increase in the 

probability or consequences of an 

accident previously evaluated. 

2. Does the proposed change create 

the possibility of a new or different kind 

of accident from any accident 

previously evaluated? 

Response: No. 

The proposed changes revise TS 5.5.8, 

‘‘Inservice Testing Program,’’ for 

consistency with the requirements of 10 

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

testing of pumps and valves which are 

classified as ASME Code Class 1, Class 

2, and Class 3. The proposed changes 

incorporate revisions to the ASME Code 

that result in a net improvement in the 

measures for testing pumps and valves. 

The proposed changes do not involve 

a modification to the physical 

configuration of the plant nor does it 

involve a change in the methods 

governing normal plant operation. The 

proposed changes will not impose any 

new or different requirements or 

introduce a new accident initiator, 

accident precursor, or malfunction 

mechanism. Additionally, there is no 

change in the types or increases in the 

amounts of any effluent that may be 

released offsite and there is no increase 

in individual or cumulative 

occupational exposure. Therefore, the 

15770 Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 

proposed changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve 

a significant reduction in a margin of 

safety? 

Response: No. 

The proposed changes revise TS 5.5.8, 

‘‘Inservice Testing Program,’’ for 

consistency with the requirements of 10 

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

testing of pumps and valves which are 

classified as ASME Code Class 1, Class 

2, and Class 3. The proposed changes do 

not involve a modification to the 

physical configuration of the plant nor 

does it change the methods governing 

normal plant operation. The proposed 

changes incorporate revisions to the 

ASME Code that result in a net 

improvement in the measures for testing 

pumps and valves. The safety function 

of the affected pumps and valves will be 

maintained. Therefore, the proposed 

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: 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 C. Wong. 

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: October 

2, 2008. 

Description of amendment request: 

The proposed amendments would 

revise technical specifications (TS) 

associated with the verification of ice 

condenser door operability. The 

proposed amendment affects the current 

TS surveillance requirements 3.6.13.5 

and 3.6.13.6. 

Basis for proposed no significant 

hazards consideration determination: 

As required by Title 10 of the Code of 

Federal Regulations (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 only analyzed accidents of 

possible consideration in regards to 

changes potentially affecting the ice 

condenser are a loss of coolant accident 

(LOCA) and a high energy line break 

(HELB) inside Containment. However, 

the ice condenser is not postulated as 

being the initiator of any LOCA or 

HELB. This is because it is designed to 

remain functional following a design 

basis earthquake, and the ice condenser 

does not interconnect or interact with 

any systems that interconnect or interact 

with the Reactor Coolant or Main Steam 

Systems. Since these proposed changes 

do not result in, or require, any physical 

change to the ice condenser that could 

introduce an interaction with the 

Reactor Coolant or Main Steam Systems, 

then there can be no change in the 

probability of an accident previously 

evaluated. Regarding consequences of 

analyzed accidents, the ice condenser is 

an engineered safety feature designed, 

in part, to limit the Containment subcompartment and Containment vessel 

pressure immediately following the 

initiation of a LOCA or HELB. 

Conservative sub-compartment and 

Containment pressure analysis shows 

these criteria will be met if the total ice 

mass within the ice bed is maintained 

in accordance with the DBA analysis; 

therefore, the proposed TS [technical 

specification] SR [surveillance 

requirement] changes of these 

requirements will not increase the 

consequences of any accident 

previously evaluated. 

Thus, based on the above, 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. 

As previously described, the ice 

condenser is not postulated as being the 

initiator of any design basis accident. 

The proposed changes do not impact 

any plant system, structure or 

component that is an accident initiator. 

The proposed TSs and TS Bases changes 

do not involve any hardware changes to 

the ice condenser or other change that 

could create any new accident 

mechanisms. Therefore, there can be no 

new or different accidents created from 

those already identified and evaluated. 

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 and the reactor coolant system 

will not be impacted by the proposed 

changes. The Application provides a 

description of additional subcompartment and Containment pressure 

response analysis that has been 

performed. This analysis demonstrates 

that Containment will remain fully 

capable of performing its design 

function with implementation of the 

proposed changes. Therefore, no safety 

margin will be significantly impacted. 

The changes proposed in this LAR 

[license amendment request] do not 

make any physical alteration to the ice 

condenser doors, nor does it affect the 

required functional capability of the 

doors in any way. The intent of the 

proposed changes to the ice condenser 

door surveillance requirements is to 

eliminate an unnecessary and overly 

restrictive Lower Inlet Door torque 

surveillance test. There will be no 

degradation in the operable status of the 

ice condenser doors and the ability to 

confirm operability for the ice 

condenser doors will be maintained, 

such that the doors will continue to 

fully perform their safety function as 

assumed in the plant’s safety analyses. 

Thus, it can be concluded that the 

proposed TS and TS Bases 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 Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of amendment request: February 

24, 2009. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specification (TS) 

Surveillance Requirement (SR) that 

governs operability testing of the 

pressure suppression chamber-drywell 

vacuum breakers to incorporate the SR 

contained within the Standard 

Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 15771 

Technical Specifications (STS), 

NUREG–1433 and delete the SR that 

requires inspection of the pressure 

suppression chamber-drywell vacuum 

breakers. Periodic inspections of the 

pressure suppression chamber-drywell 

vacuum breakers are not required by the 

STS. 

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. The operation of Vermont Yankee 

Nuclear Power Station (VY) in 

accordance with the proposed 

amendment will not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The proposed amendment does not 

impact the operability of any structure, 

system or component that affects the 

probability of an accident or that 

supports mitigation of an accident 

previously evaluated. The proposed 

amendment does not affect reactor 

operations or accident analysis and has 

no radiological consequences. The 

operability requirements for accident 

mitigation systems remain consistent 

with the licensing and design basis. 

Therefore, the proposed amendment 

does not involve a significant increase 

in the probability or consequences of an 

accident previously evaluated. 

2. The operation of VY in accordance 

with the proposed amendment will not 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated. 

The proposed amendment does not 

change the design or function of any 

component or system. No new modes of 

failure or initiating events are being 

introduced. Therefore, operation of VY 

in accordance with the proposed 

amendment will not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. The operation of VY in accordance 

with the proposed amendment will not 

involve a significant reduction in a 

margin of safety. 

The proposed amendment does not 

change the design or function of any 

component or system. The proposed 

amendment does not involve any safety 

limits, safety settings or safety margins. 

The ability of the pressure suppression 

chamber-drywell vacuum breakers to 

perform its intended function will 

continue to be required in accordance 

with the VY Technical Specifications. 

Since the proposed controls are 

adequate to ensure the operability of the 

pressure suppression chamber-drywell 

vacuum breakers, there will still be high 

assurance that the components are 

operable and capable of performing 

their respective functions. Therefore, 

operation of VY in accordance with the 

proposed amendment will not involve a 

significant reduction in the margin to 

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., 400 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Mark G. Kowal. 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, LaSalle 

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Date of amendment request: October 

23, 2008. 

Description of amendment request: 

The proposed amendments would 

revise the Technical Specifications (TS) 

to support the application of alternative 

source term (AST) methodology with 

respect to the loss-of-coolant accident 

and the fuel handling accident. The 

proposed request is to support a fullscope application of an AST 

methodology, with the exception that 

Technical Information Document (TID)– 

14844, ‘‘Calculation of Distance Factors 

for Power and Test Reactor Sites,’’ will 

continue to be used as the radiation 

dose basis for equipment qualification. 

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 implementation of AST 

assumptions has been evaluated in 

revisions to the analyses of the 

following limiting design basis 

accidents at LSCS [LaSalle County 

Station]: 

• Loss-of-Coolant Accident, and 

• Fuel Handling Accident 

Based upon the results of these 

analyses, it has been demonstrated that, 

with the requested changes, the dose 

consequences of these limiting events 

are within the regulatory requirements 

and guidance provided by the NRC for 

use with AST. The regulatory 

requirements and guidance is presented 

in 10 CFR 50.67, ‘‘Accident source 

term,’’ and associated NRC Regulatory 

Guide 1.183 and Standard Review Plan 

section 15.0.1. The AST is an input to 

calculations used to evaluate the 

consequences of an accident, and does 

not by itself affect the plant response, or 

the actual pathway of the radiation 

released from the fuel. It does, however, 

better represent the physical 

characteristics of the release, so that 

appropriate mitigation techniques may 

be applied. Therefore, the consequences 

of an accident previously evaluated are 

not significantly increased. 

The equipment affected by the 

proposed change is mitigative in nature, 

and relied upon after an accident has 

been initiated. Application of the AST 

does not involve any physical changes 

to the TS, while they revise certain 

performance requirements, do not 

involve any physical modifications to 

the plant. As a result, the proposed 

changes do not affect any of the 

parameters or conditions that could 

contribute to the initiation of any 

accidents. As such, removal of 

operability requirements during the 

specified conditions will not 

significantly increase the probability of 

occurrence for an accident previously 

analyzed. Since plant design basis 

accidents initiators are not being altered 

by adoption of the AST analyses, the 

probability of an accident previously 

evaluated is not affected. 

Therefore, the proposed change does 

not involve a significant increase in the 

probability or consequences of an 

accident previously evaluated. 

2. Does the proposed change create 

the possibility of a new or different kind 

of accident from any accident 

previously evaluated? 

Response: No. 

The proposed change does not 

involve a physical alteration of the plant 

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

equipment will be installed and there 

are no physical modifications to existing 

equipment associated with the proposed 

change). Similarly, it does not 

physically change any structures, 

systems, or components involved in the 

mitigation of any accidents. Thus, no 

new initiators or precursors of a new or 

different kind of accident 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? 

Response: No. 

15772 Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 

Safety margins and analytical 

conservatisms have been evaluated and 

have been found to be acceptable. The 

analyzed events have been carefully 

selected and margin has been retained 

to ensure that the analyses adequately 

bound postulated event scenarios. The 

dose consequences due to design basis 

accidents comply with the requirements 

of 10 CFR 50.67 and guidance of 

Regulatory Guide 1.183. 

The proposed change is associated 

with the implementation of a new 

licensing basis for LSCS design basis 

accidents. Approval of the change from 

the original source term to a new source 

term taken from Regulatory Guide 1.183 

is being requested. The results of the 

accident analyses, revised in support of 

the proposed license amendment, are 

subject to revised acceptance criteria. 

The analyses have been performed using 

conservative methodologies, as 

specified in Regulatory Guide 1.183. 

Safety margins have been evaluated and 

analytical conservatism has been 

utilized to ensure that the analyses 

adequately bound the postulated 

limiting event scenario. The dose 

consequences of these design basis 

accidents remain within the acceptance 

criteria presented in 10 CFR 50.67 and 

Regulatory Guide 1.183. 

The proposed change continues to 

ensure that the doses at the exclusion 

area boundary and low population zone 

boundary, as well as the control room, 

are within corresponding regulatory 

limits. 

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 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Nuclear, 4300 Winfield Road, 

Warrenville, IL 60555. 

NRC Branch Chief: Russell Gibbs. 

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: February 

11, 2009. 

Brief description of amendment: The 

proposed amendment consists of 

administrative revision to the operating 

licenses and Technical Specifications 

(TSs) to revise the station name from 

Comanche Peak Steam Electric Station 

(CPSES) to Comanche Peak Nuclear 

Power Plant (CPNPP); remove the Table 

of Contents from TSs and maintain and 

revise it in accordance with plant 

administrative procedures; delete TSs 

3.2.1.1, 3.2.3.1, 5.5.9.1, 5.6.10 and 

several footnotes from Tables 3.3.1–1, 

3.3.2–1, and TS 3.4.10 since these TSs 

and footnotes are no longer applicable 

to CPSES, Units 1 and 2 operation; 

delete several topical reports from the 

list of approved analytical methods used 

to determine core operating limits in TS 

5.6.5, no longer in use, since these 

topical reports have been replaced by 

standard Westinghouse methods and 

Westinghouse methods have been 

approved for use at CPSES, Units 1 and 

2, under a separate amendment request; 

make editorial corrections; and reprint 

and reissue the entire TS due to 

adoption of ‘FrameMaker’ software in 

place of ‘Microsoft Word’ software. 

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 revises the 

station name, removes the Table of 

Contents from the Technical 

Specifications, deletes several Technical 

Specifications and footnotes which are 

no longer applicable to [CPSES] Unit 1 

or Unit 2 operation, renumbers 

subsequent Technical Specifications, 

deletes several topical reports from the 

list of approved analytical methods used 

to determine core operating limits, and 

corrects various editorial and formatting 

errors. The Table of Contents does not 

include information required by 10 CFR 

50.36 [Title 10 of the Code of Federal 

Regulations, Section 50.36] to be 

reviewed by the NRC [U.S. Nuclear 

Regulatory Commission] staff and is not 

required by the regulation. The 

Technical Specifications and footnotes 

which are being deleted were only 

applicable during previous operational 

cycles and are now defunct 

requirements since both Units have 

completed the applicable operational 

cycles. The topical reports deleted from 

Technical Specification 5.6.5b are no 

longer used to determine the core 

operating limits for Comanche Peak 

Nuclear Power Plant. The remaining 

topical reports listed in Technical 

Specification 5.6.5b will be used to 

determine the core operating limits for 

both Comanche Peak Nuclear Power 

Plant units. All other changes proposed 

are corrections of previous inadvertent 

editorial errors or changes in format to 

increase conformity with the guidelines 

described in TSTF–RPT–01, ‘‘Writer’s 

Guide for Plant-Specific Improved 

Technical Specifications’’, published in 

June, 2005. All of the proposed changes 

are administrative changes which do not 

change the meaning, intent, 

interpretation, or application of the 

Technical Specifications. None of the 

proposed changes affect the operation, 

physical configuration, or function of 

plant equipment or systems. The 

changes do not affect the initiators or 

assumptions of analyzed events; nor do 

they impact the mitigation of accidents 

or transient events. 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 revises the 

station name, removes the Table of 

Contents from the Technical 

Specifications, deletes several Technical 

Specifications and footnotes which are 

no longer applicable to [CPSES,] Unit 1 

or Unit 2 operation, renumbers 

subsequent Technical Specifications, 

deletes several topical reports from the 

list of approved analytical methods used 

to determine core operating limits, and 

corrects various editorial and formatting 

errors. The Table of Contents does not 

include information required by 10 CFR 

50.36 to be reviewed by the Nuclear 

Regulatory Commission staff and is not 

required by the regulation. The 

Technical Specifications and footnotes 

which are being deleted were only 

applicable during previous operational 

cycles and are now defunct 

requirements since both Units have 

completed the applicable operational 

cycles. The topical reports deleted from 

Technical Specification 5.6.5b are no 

longer used to determine the core 

operating limits for Comanche Peak 

Nuclear Power Plant. The remaining 

topical reports listed in Technical 

Specification 5.6.5b will be used to 

determine the core operating limits for 

both Comanche Peak Nuclear Power 

Plant units. All other changes proposed 

are corrections of previous inadvertent 

editorial errors or changes in format to 

increase conformity with the guidelines 

described in TSTF–RPT–01, ‘‘Writer’s 

Guide for Plant-Specific Improved 

Technical Specifications’’, published in 

June, 2005. All of the proposed changes 

are administrative changes which do not 

change the meaning, intent, 

Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 15773 

interpretation, or application of the 

Technical Specifications. None of the 

changes alter the plant configuration, 

require installation of new equipment, 

alter assumptions about previously 

analyzed accidents, or impact the 

operation or function of any plant 

equipment or systems. Therefore, the 

proposed changes do not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of 

safety? 

Response: No. 

The proposed change revises the 

station name, removes the Table of 

Contents from the Technical 

Specifications, deletes several Technical 

Specifications and footnotes which are 

no longer applicable to [CPSES,] Unit 1 

or Unit 2 operation, renumbers 

subsequent Technical Specifications, 

deletes several topical reports from the 

list of approved analytical methods used 

to determine core operating limits, and 

corrects various editorial and formatting 

errors. The Table of Contents does not 

include information required by 10 CFR 

50.36 to be reviewed by the Nuclear 

Regulatory Commission staff and is not 

required by the regulation. The 

Technical Specifications and footnotes 

which are being deleted were only 

applicable during previous operational 

cycles and are now defunct 

requirements since both Units have 

completed the applicable operational 

cycles. The topical reports deleted from 

Technical Specification 5.6.5b are no 

longer used to determine the core 

operating limits for Comanche Peak 

Nuclear Power Plant. The remaining 

topical reports listed in Technical 

Specification 5.6.5b will be used to 

determine the core operating limits for 

both Comanche Peak Nuclear Power 

Plant units. All other changes proposed 

are corrections of previous inadvertent 

editorial errors or changes in format to 

increase conformity with the guidelines 

described in TSTF–RPT–01, ‘‘Writer’s 

Guide for Plant-Specific Improved 

Technical Specifications’’, published in 

June, 2005. All of the proposed changes 

are administrative changes which do not 

change the meaning, intent, 

interpretation, or application of the 

Technical Specifications. None of the 

proposed changes alter the effective 

technical content of the Technical 

Specifications. Therefore the proposed 

changes do 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. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

No. 1, Washington County, Nebraska 

Date of amendment request: October 

31, 2008. 

Description of amendment request: 

The proposed amendment modifies the 

surveillance requirements in Technical 

Specification (TS) 3.6(3), ‘‘Containment 

Recirculating Air Cooling and Filtering 

System,’’ and removes the license 

conditions related to the replacement 

and testing of containment air cooling 

and filtering (CACF) unit high-efficiency 

particulate air (HEPA) filters and 

surveillance testing of the CACF unit 

relief ports. These license conditions 

were committed to by the licensee in its 

letter dated April 10, 2008 (Agencywide 

Documents Access and Management 

System (ADAMS) Accession No. 

ML081010122), and implemented via 

TS Amendment No. 255 (ADAMS 

Accession No. ML081140390), dated 

May 2, 2008. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated? 

Response: No. 

The containment air cooling and 

filtering system (CACFS) is not an 

initiator of any accident previously 

evaluated at the Fort Calhoun Station 

(FCS). The CACFS is an accident 

mitigation system. The design basis 

function of the CACFS is to limit the 

containment pressure rise by providing 

a means for cooling the containment 

following a loss-of-coolant accident 

(LOCA) or main steam line break 

(MSLB). In accordance with TS 

Amendment No. 255, the CACFS high 

efficiency particulate air (HEPA) filters 

are also credited to reduce post-LOCA 

radioactive leakage from containment. 

The proposed changes provide 

additional assurance that the CACFS is 

capable of performing its design and 

licensing basis functions to mitigate 

these design basis accidents (DBAs). 

The CACFS face and bypass dampers 

are aligned to their accident positions 

permanently causing the CACFS to 

operate in filtered air mode. 

Surveillance testing has shown that 

operating the system in this alignment 

over long periods does not jeopardize 

filter performance. Over the lifetime of 

the plant, the differential pressures 

measured across the combined HEPA 

and charcoal filter banks have met test 

acceptance criteria. 

Increasing the number of surveillance 

requirements will not adversely affect 

the function of the CACFS but rather 

provides additional assurance that the 

CACFS is capable of responding to a 

DBA. 

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 CACFS was designed to remove 

heat released to the containment 

atmosphere during a DBA to the extent 

necessary to maintain the containment 

structure below its design pressure. The 

containment airflow continually passes 

through the cooling coils. The proposed 

changes to the surveillance 

requirements do not affect the active 

function of the CACFS. 

The CACFS will continue to operate 

in normal and accident conditions to 

remove heat and radioactive particulates 

and aerosols. The proposed changes 

enhance surveillance testing of the 

CACFS by requiring more frequent 

exercising of the fans, imposing a more 

stringent pressure drop limit, specifying 

a HEPA filter replacement interval, and 

instituting a requirement to exercise the 

relief ports. These changes ensure that 

the CACFS is capable of long-term 

operation in filtered air mode while 

remaining capable of providing cooling 

and filtering sufficient to mitigate 

design basis accidents. 

No credible new failure mechanisms, 

malfunctions, or accident initiators not 

previously considered in the design and 

licensing basis are created and none of 

the initial condition assumptions of any 

accident evaluated in the safety analysis 

are impacted. 

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. 

The containment building and 

associated penetrations are designed to 

15774 Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 

withstand an internal pressure of 60 

psig [pounds per square inch gauge] at 

305°F, including all thermal loads 

resulting from the temperature 

associated with this pressure, with a 

leakage rate of 0.1 percent by weight or 

less of the contained volume per 24 

hours. [Omaha Public Power District] 

credits the CACFS in the containment 

pressure analysis for a LOCA, and for 

the containment pressure response to a 

main steam line break (MSLB). 

The proposed changes impose more 

stringent surveillance test requirements. 

This provides additional assurance that 

the CACFS will perform its design basis 

and licensing basis functions to be 

capable of long-term post-DBA 

operation in filtered air mode to limit 

the containment temperature and 

pressure increase to within design limits 

and to reduce post-LOCA radioactive 

leakage from containment. 

Neither the design basis nor the 

licensing basis for post-DBA 

containment heat removal is adversely 

affected by the proposed changes. The 

ability to maintain design limits for 

containment peak pressure and 

temperature, as well as long-term 

containment pressure and temperature, 

are preserved. 

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: David A. Repka, 

Esq., Winston & Strawn, 1700 K Street, 

NW., Washington, DC 20006–3817. 

NRC Branch Chief: Michael T. 

Markley. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

No. 1, Washington County, Nebraska 

Date of amendment request: January 

30, 2009. 

Description of amendment request: 

The proposed amendment would 

modify the Fort Calhoun Station (FCS), 

Unit No. 1, Renewed Operating License 

No. DPR–40, by adding operability and 

surveillance testing requirements to the 

FCS Technical Specifications (TS) for 

the steam generator (SG) blowdown 

isolation on a reactor trip. Specifically, 

the proposed changes will revise TS 

Limiting Conditions for Operation 

(LCO) 2.15, Instrumentation and Control 

Systems, Table 2–4, Instrument 

Operating Conditions for Isolation 

Functions, to include operability 

requirements for SG blowdown isolation 

on a reactor trip and to add applicable 

footnotes. In addition, TS 3.1, 

Instrumentation and Control, Table 3–2, 

Minimum Frequencies for Checks, 

Calibrations and Testing of Engineered 

Safety Features, Instrumentation and 

Controls, is being revised to include the 

surveillance test requirements for SG 

blowdown isolation on a reactor trip. 

An administrative change is also being 

made to TS LCO 2.15(1), to delete the 

words ‘‘key operated’’ as the ‘‘key’’ 

associated with the bypass switches is 

not a critical element in controlling the 

use of bypass switches. This 

amendment will allow FCS to credit an 

automatic SG blowdown isolation 

interlock being installed during the 

2009 Refueling Outage (RFO). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated? 

Response: No. 

The proposed change provides 

Technical Specification (TS) operability 

and surveillance testing requirements 

for automatic steam generator (SG) 

blowdown isolation on a reactor trip in 

the event of a loss of main feedwater 

(LMFW). Automatic isolation will 

ensure that the existing 15-minute 

requirement in the Updated Safety 

Analysis Report (USAR) Chapter 14.10 

safety analysis is met without the risk 

that an unanticipated distraction could 

prevent manual action from occurring at 

the proper time. The installation of this 

feature will eliminate the need for 

manual isolation of blowdown and thus 

will eliminate the associated operator 

challenge. 

Automatic isolation of blowdown will 

reduce the consequences of the LMFW 

event by providing automatic isolation 

prior to manual isolation being initiated 

by the operators. Automatic isolation at 

the time of reactor trip will reduce the 

severity of the LMFW event by isolating 

the SGs earlier in the event, thereby 

conserving SG inventory. 

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. 

No new malfunctions are being 

introduced by this activity, and based 

on the current redundancy in the 

design, there are no malfunctions of the 

SG blowdown isolation valves that 

challenge nuclear safety. 

The SG blowdown isolation valves 

will continue to function as currently 

credited for the LMFW event; thus, this 

proposed change does not alter their 

ability to function as containment 

isolation valves to maintain 

containment integrity. The manual 

isolation capability remains unchanged. 

A failure analysis has been prepared 

which shows that the addition of the 

automatic isolation feature does not 

introduce a new failure mode or 

malfunction to the valve circuits. An 

isolation of SG blowdown, either 

through the designed circuit following a 

reactor trip, or during normal 

operations, does not present a nuclear 

safety challenge. The capability exists 

for operators to bypass the isolation 

signal and restore blowdown as plant 

conditions warrant. 

Therefore, the proposed change does 

not create the possibility of a new or 

different kind of accident from any 

accident previously evaluated. 

3. Does the proposed amendment 

involve a significant reduction in a 

margin of safety? 

Response: No. 

The addition of an automatic isolation 

interlock to the SG blowdown isolation 

valve circuits that close the valves on a 

reactor trip actually increases the 

margin of safety by isolating the SG 

early in the event to maintain SG 

inventories. 

A reactor trip signal is generated in 

the first seconds of an LMFW due to 

reduced SG inventories. Because it is 

desirable to isolate blowdown as soon as 

possible following the LMFW event, for 

maximum margin, a reactor trip signal 

will be used for the SG blowdown 

isolation interlock. Isolating blowdown 

earlier in an event provides greater 

operating margin in terms of 

maximizing SG inventories. More 

margin allows operators more time to 

address operator demands that occur 

during transient events. 

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. 

Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 15775 

Attorney for licensee: David A. Repka, 

Esq., Winston & Strawn, 1700 K Street, 

NW., Washington, DC 20006–3817. 

NRC Branch Chief: Michael T. 

Markley. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

No. 1, Washington County, Nebraska 

Date of amendment request: January 

30, 2009. 

Description of amendment request: 

The proposed amendment would delete 

those portions of the Technical 

Specifications (TS) superseded by Title 

10 of the Code of Federal Regulations 

(10 CFR) Part 26, Subpart I. The licensee 

is proposing to adopt the approved 

Technical Specification Task Force 

(TSTF) change traveler TSTF–511, 

Revision 0, ‘‘Eliminate Working Hour 

Restrictions from TS 5.2.2 to Support 

Compliance with 10 CFR Part 26.’’ 

The NRC staff issued a ‘‘Notice of 

Availability of Model Safety Evaluation, 

Model No Significant Hazards 

Determination, and Model Application 

for Licensees That Wish To Adopt 

TSTF–511, Revision 0, ‘‘Eliminate 

Working Hour Restrictions From TS 

5.2.2 To Support Compliance With 10 

CFR Part 26,’’ in the Federal Register on 

December 30, 2008 (73 FR 79923). The 

notice included a model safety 

evaluation, a model no significant 

hazards consideration (NSHC) 

determination, and a model license 

amendment request, using the 

consolidated line item improvement 

process. In its application dated January 

30, 2009, the licensee affirmed the 

applicability of the model NSHC 

determination, which is presented 

below. 

Basis for proposed (NSHC) 

determination: As required by 10 CFR 

50.91(a), an analysis of the issue of 

NSHC determination is presented 

below: 

Criterion 1—The Proposed Change Does 

Not Involve a Significant Increase in the 

Probability or Consequences of an 

Accident Previously Evaluated 

The proposed change removes 

Technical Specification restrictions on 

working hours for personnel who 

perform safety related functions. The 

Technical Specification restrictions are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. 

Removal of the Technical Specification 

requirements will be performed 

concurrently with the implementation 

of the 10 CFR Part 26, Subpart I, 

requirements. The proposed change 

does not impact the physical 

configuration or function of plant 

structures, systems, or components 

(SSCs) or the manner in which SSCs are 

operated, maintained, modified, tested, 

or inspected. Worker fatigue is not an 

initiator of any accident previously 

evaluated. Worker fatigue is not an 

assumption in the consequence 

mitigation of any accident previously 

evaluated. Therefore, it is concluded 

that this change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Criterion 2—The Proposed Change Does 

Not Create the Possibility of a New or 

Different Kind of Accident From Any 

Accident Previously Evaluated 

The proposed change removes 

Technical Specification restrictions on 

working hours for personnel who 

perform safety related functions. The 

Technical Specification restrictions are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. 

Working hours will continue to be 

controlled in accordance with NRC 

requirements. The new rule allows for 

deviations from controls to mitigate or 

prevent a condition adverse to safety or 

as necessary to maintain the security of 

the facility. This ensures that the new 

rule will not unnecessarily restrict 

working hours and thereby create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. The proposed change does 

not alter the plant configuration, require 

new plant equipment to be installed, 

alter accident analysis assumptions, add 

any initiators, or effect the function of 

plant systems or the manner in which 

systems are operated, maintained, 

modified, tested, or inspected. 

Therefore, the proposed change does not 

create the possibility of a new or 

different kind of accident from any 

previously evaluated. 

Criterion 3—The Proposed Change Does 

Not Involve a Significant Reduction in 

a Margin of Safety 

The proposed change removes 

Technical Specification restrictions on 

working hours for personnel who 

perform safety related functions. The 

Technical Specification restrictions are 

superseded by the worker fatigue 

requirements in 10 CFR Part 26. The 

proposed change does not involve any 

physical changes to plant or alter the 

manner in which plant systems are 

operated, maintained, modified, tested, 

or inspected. The proposed change does 

not alter the manner in which safety 

limits, limiting safety system settings or 

limiting conditions for operation are 

determined. The safety analysis 

acceptance criteria are not affected by 

this change. The proposed change will 

not result in plant operation in a 

configuration outside the design basis. 

The proposed change does not adversely 

affect systems that respond to safely 

shutdown the plant and to maintain the 

plant in a safe shutdown condition. 

Removal of plant-specific Technical 

Specification administrative 

requirements will not reduce a margin 

of safety because the requirements in 10 

CFR Part 26 are adequate to ensure that 

worker fatigue is managed. Therefore, 

the proposed change does not involve a 

significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

analysis adopted by the licensee and, 

based on this review, it appears that the 

three standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves NSHC. 

Attorney for licensee: David A. Repka, 

Esq., Winston & Strawn, 1700 K Street, 

NW., Washington, DC 20006–3817. 

NRC Branch Chief: Michael T. 

Markley. 

R.E. Ginna Nuclear Power Plant, LLC, 

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

Power Plant, Wayne County, New York 

Date of amendment request: 

December 19, 2008. 

Description of amendment request: 

The proposed amendment would 

modify the Technical Specifications 

(TSs) to (1) correct an error in TS Table 

3.3.2–1, ‘‘Engineered Safety Feature 

Actuation System Instrumentation,’’ 

Function 1.a, to reflect the correct 

CONDITIONS for applicable Modes 1, 2, 

3, and 4, (2) revise TS Limiting 

Condition for Operation (LCO) 3.3.4 

degraded voltage relay and loss of 

voltage relay Limiting Safety System 

Settings values to reflect the revised 

analysis, and (3) revise the load 

requirement of Surveillance 

Requirement 3.8.1.3 to reflect values 

supported by the diesel generator 

accident loading analyses. 

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 to LCO 3.3.2 

correct an administrative error which 

directed inadequate action in the event 

that a channel of instrumentation is lost 

for manual safety injection initiation. 

The amendment places the plant in a 

15776 Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 

more conservative condition, Mode 5, if 

the other Required Actions cannot be 

executed within their periodicity. 

The proposed changes to LCO 3.3.4 

provide setpoint changes based on a 

revised calculation, which generated 

new setpoints for the loss of voltage 

relays and degraded voltage relays. The 

new setpoints ensure the protective 

relays will function when required, will 

ensure protection from thermal damage 

to loads on the 480V busses, and will 

not cause unintended diesel generator 

starts even in worst case scenarios, with 

power provided from offsite. 

The proposed changes to LCO 3.8.1 

involve an increase in the minimum 

load band value for diesel generator 

surveillance SR 3.8.1.3. This change 

ensures that the diesel generators are 

capable of synchronizing with the 

offsite electrical system and accepting 

loads greater than or equal [to] the 

equivalent of the maximum expected 

accident loads. The new load band 

value is more conservative than the 

existing value and provides a more 

thorough test to ensure equipment 

emergency response capability. 

Therefore, the probability or 

consequences of an accident previously 

evaluated will not be significantly 

increased. 

2. Do the proposed amendments 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated? 

Response: No. 

The proposed changes involve 

correcting an administrative error and 

revising previously established values 

associated with the diesel generators to 

increase conservatism. None of these 

proposed changes involve a physical 

alteration of the plant (i.e., no new or 

different types of equipment will be 

installed) or a change in methods 

governing normal plant operation. The 

proposed changes preserve the safety 

analysis assumptions related to accident 

mitigation. No initiators or accident 

precursors are created by this change. 

Therefore, the possibility of a new or 

different kind of accident not previously 

evaluated is not created. 

3. Do the proposed amendments 

involve a significant reduction in a 

margin of safety? 

Response: No. 

The level of safety of facility 

operation is unaffected by any of the 

proposed changes. The requested 

administrative change is conservative 

compared to the existing requirement. 

The response of the diesel generators to 

accident transients reported in the 

Updated Final Safety Analysis Report 

(UFSAR) is unaffected by these changes. 

The proposed changes preserve the 

safety analysis assumptions related to 

accident mitigation. Therefore, these 

changes do not involve a significant 

reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Carey Fleming, 

Sr. Counsel—Nuclear Generation, 

Constellation Group, LLC, 750 East Pratt 

Street, 17 Floor, Baltimore, MD 21202. 

NRC Branch Chief: Mark G. Kowal. 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

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. 

Carolina Power & Light Company, 

Docket Nos. 50–325 and 50–324, 

Brunswick Steam Electric Plant, Units 1 

and 2, Brunswick County, North 

Carolina 

Date of application for amendments: 

July 7, 2008, as supplemented by letters 

dated December 17, 2008, and March 9, 

2009. 

Brief description of amendments: The 

amendments revise Surveillance 

Requirement (SR) 3.6.1.6.1 to add a new 

requirement to verify that each vacuum 

breaker is closed within 6 hours 

following an operation that causes any 

of the vacuum breakers to open and, 

also, revise SR 3.6.1.6.2 by removing the 

requirement to perform functional 

testing of each vacuum breaker within 

12 hours following an operation that 

causes any of the vacuum breakers to 

open. 

Date of issuance: March 11, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment Nos.: 251 and 279. 

Facility Operating License Nos. DPR– 

71 and DPR–62: Amendments change 

the Technical Specifications. 

Date of initial notice in Federal 

Register: September 23, 2008 (73 FR 

54864). The supplemental letter 

provided clarifying information that was 

within the scope of the initial notice 

and did not change the initial proposed 

no significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 11, 2009. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and STN 50– 

457, Braidwood Station, Units 1 and 2, 

Will County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–454 and STN 50– 

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

Ogle County, Illinois 

Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 15777 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station, Unit No. 1, DeWitt County, 

Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, LaSalle 

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Exelon Generation Company, LLC, 

Docket No. 50–352 and No. 50–353, 

Limerick Generating Station, Unit 1 and 

2, Montgomery County, Pennsylvania 

Exelon Generation Company, LLC, 

Docket No. 50–219, Oyster Creek 

Nuclear Generating Station, Ocean 

County, New Jersey 

Exelon Generation Company, LLC, and 

PSEG Nuclear LLC, Docket Nos. 50–277 

and 50–278, Peach Bottom Atomic 

Power Station, Units 2 and 3, York and 

Lancaster Counties, Pennsylvania 

Exelon Generation Company, LLC, 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Units 1 

and 2, Rock Island County, Illinois 

Exelon Generation Company, LLC, 

Docket No. 50–289, Three Mile Island 

Nuclear Station, Unit 1 (TMI–1), 

Dauphin County, Pennsylvania 

Date of application for amendments: 

April 21, 2008, as supplemented on 

March 11, 2009. 

Brief description of amendments: The 

proposed amendment removes 

references to and limits provided by 

Nuclear Regulatory Commission Generic 

Letter (GL) 82–12, ‘‘Nuclear Power Plant 

Staff Working Hours,’’ from the subject 

plants’ technical specifications (TS). 

The references and limitations have 

been superseded by the requirements of 

Title 10 of the Code of Federal 

Regulations, Part 26 (10 CFR 26), 

Subpart I, ‘‘Managing Fatigue.’’ 

Date of issuance: March 23, 2009. 

Effective date: As of the date of 

issuance and shall be implemented by 

October 1, 2009. 

Amendment Nos.: 157, 157, 162, 162, 

185, 231, 224, 192, 179, 198, 159, 274, 

271, 275, 243, 238, 270. 

Facility Operating License Nos. NPF– 

72, NPF–77, NPF–37, NPF–66, NPF–62, 

DPR–19, DPR–25, NPF–11, NPF–18, 

NPF–39, NPF–85, DPR–16, DPR–44, 

DPR–56, DPR–29, DPR–30, DPR–50: The 

amendments revised the Technical 

Specifications/Licenses. 

Date of initial notice in Federal 

Register: June 3, 2008 (73 FR 31721). 

The March 11, 2009, supplement 

contained clarifying information and 

did not change the NRC staff’s initial 

proposed finding of no significant 

hazards consideration. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 23, 2009. 

No significant hazards consideration 

comments received: No. 

Nebraska Public Power District, Docket 

No. 50–298, Cooper Nuclear Station, 

Nemaha County, Nebraska 

Date of amendment request: March 

24, 2008, as supplemented by letters 

dated September 11 and 19, 2008, 

November 6, 2008, and February 26, 

2009. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) Section 3.7.3, 

‘‘Reactor Equipment Cooling (REC) 

System,’’ to allow credit for the ability 

to align the service water system to the 

REC system. 

Date of issuance: March 20, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No.: 232. 

Facility Operating License No. DPR– 

46: Amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: April 22, 2008 (73 FR 21660). 

The supplemental letters dated 

September 11 and 19, 2008, November 

6, 2008, and February 26, 2009, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 20, 2009. 

No significant hazards consideration 

comments received: No. 

Nine Mile Point Nuclear Station, LLC, 

Docket No. 50–220, Nine Mile Point 

Nuclear Station, Unit No. 1 (NMP1), 

Oswego County, New York 

Date of application for amendment: 

August 15, 2008, as supplemented on 

December 4, 2008. 

Brief description of amendments: The 

amendment revises NMP1 Technical 

Specification (TS) 6.5.7, ‘‘10 CFR 50 

[Part 50 of Title 10 of the Code of 

Federal Regulations Appendix J Testing 

Program Plan,’’ to allow a one-time 

extension of the Integrated Leak Rate 

Test (ILRT) interval for no more than 5 

years. The amendment allows the next 

ILRT for NMP1 to be performed within 

15 years from the last ILRT as opposed 

to the current 10-year interval. 

Date of issuance: March 11, 2009. 

Effective date: As of the date of 

issuance to be implemented within 30 

days. 

Amendment No.: 202. 

Renewed Facility Operating License 

No. DPR–063: The amendment revises 

the License and TSs. 

Date of initial notice in Federal 

Register: October 21, 2008 (73 FR 

62566). The supplement dated 

December 4, 2008, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the Nuclear 

Regulatory Commission staff’s initial 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 11, 2009. 

No significant hazards consideration 

comments received: No. 

Northern States Power Company— 

Minnesota, LLC, Docket No. 50–263, 

Monticello Nuclear Generating Plant, 

Wright County, Minnesota 

Date of application for amendment: 

April 3, 2008, as supplemented on 

February 23, 2009. 

Brief description of amendment: The 

amendment adopted the proposed 

requirements regarding control room 

envelope habitability set forth in 

Technical Specifications Task Force 

(TSTF) change traveler TSTF–448, 

Revision 3. Specifically, the amendment 

revised the requirements in TS Section 

3.7.4, ‘‘Control Room Emergency 

Filtration (CREF) System,’’ adds a new 

TS Section 5.5.13, ‘‘Control Room 

Envelope Habitability Program,’’ and 

added a license condition to the 

operating license to implement the TS 

changes. 

Date of issuance: March 17, 2009. 

Effective date: As of the date of 

issuance and shall be implemented by 

November 1, 2009. 

Amendment No.: 160. 

Facility Operating License No. DPR– 

22. Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: May 6, 2008 (73 FR 25043). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated March 17, 2009. 

No significant hazards consideration 

comments received: No. 

15778 Federal Register / Vol. 74, No. 65 / Tuesday, April 7, 2009 / Notices 

PPL Susquehanna, LLC, Docket Nos. 50– 

387 and 50–388, Susquehanna Steam 

Electric Station, Units 1 and 2, Luzerne 

County, Pennsylvania 

Date of application for amendments: 

July 31, 2008. 

Brief description of amendments: The 

amendments changed the PPL 

Susquehanna, LLC (PPL) Units 1 and 2 

Technical Specification 3.6.1.3 

‘‘Primary Containment Isolation Valves 

(PCIVs).’’ It revised the Secondary 

Containment Bypass Leakage limit in 

Surveillance Requirement 3.6.1.3.11 

from ‘‘less than or equal to 9 standard 

cubic foot/feet per hour (scfh)’’ to ‘‘less 

than or equal to 15 scfh when 

pressurized to greater than or equal to 

Pa.’’ 

Date of issuance: March 18, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment Nos.: 251 for Unit 1 and 

231 for Unit 2. 

Facility Operating License Nos. NPF– 

14 and NPF–22: The amendments 

revised the License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: November 18, 2008 (73 FR 

68455). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation (SE) 

dated March 18, 2009. 

No significant hazards consideration 

comments received: No. However, 

comments have been received from the 

Commonwealth of Pennsylvania and 

have been addressed in the SE. 

Virginia Electric and Power Company, 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Units 1 and 2, 

Louisa County, Virginia 

Date of application for amendment: 

March 19, 2008, as supplemented 

October 7, 2008, November 17, 2008, 

and December 10, 2008. 

Brief description of amendment: The 

amendments revise the technical 

specifications (TSs) to (1) delete TS 

3.7.13, ‘‘MCR/ESGR Bottled Air 

System,’’ (2) create TS 3.3.6, ‘‘Main 

Control Room/Emergency Switchgear 

Room (MCR/ESGR) Envelope Isolation 

Actuation Instrumentation,’’ to establish 

the operability requirements for the 

MCR/ESGR envelope isolation function, 

and (3) incorporate TS 3.7.14, ‘‘MCR/ 

ESGR Emergency Ventilation During 

Movement of Recently Irradiated Fuel 

Assemblies,’’ into TS 3.7.10, ‘‘MCR/ 

ESGR Emergency Ventilation System.’’ 

The changes revise the TSs to be 

consistent with the assumptions of the 

current dose analysis of record, 

performed in accordance with Title 10 

of the Code of Federal Regulations, 

Section 50.67, ‘‘Accident Source Term,’’ 

and the results of the nonpressurized 

MCR/ESGR envelope tracer gas testing. 

Date of issuance: March 25, 2009. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment Nos.: 255/236. 

Renewed Facility Operating License 

Nos. NPF–4 and NPF–7: Amendments 

change the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: April 22, 2008 (73 FR 21661). 

The supplements dated October 7, 2008, 

November 17, 2008, and December 10, 

2008, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination. The Commission’s 

related evaluation of the amendments is 

contained in a Safety Evaluation dated 

March 25, 2009. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 30th of 

March, 2009. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E9–7494 Filed 4–6–09; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket Nos. 50–315 and 50–316; NRC– 

2009–0153] 

Indiana Michigan Power Company; 

Donald C. Cook Nuclear Plant, Units 1 

and 2; Exemption 

1.0 Background 

The Indiana Michigan Power 

Company (the licensee) is the holder of 

Facility Operating License Nos. DPR–58 

and DPR–74, which authorizes 

operation of the Donald C. Cook Nuclear 

Plant, Units 1 and 2. The licenses 

provide, among other things, that the 

facility is subject to all rules, 

regulations, and orders of the U.S. 

Nuclear Regulatory Commission (NRC, 

the Commission) now or hereafter in 

effect. 

The facility consists of two 

pressurized-water reactors located in 

Berrien County in Michigan. 

2.0 Request/Action 

Title 10 of the Code of Federal 

Regulations, Part 50, Section 36a(a)(2) 

(10 CFR 50.36a(a)(2)) requires each 

licensee to submit a report to the 

Commission annually that specifies the 

quantity of each of the principal 

radionuclides released to unrestricted 

areas in liquid and in gaseous effluents 

during the previous 12 months, 

including any other information as may 

be required by the Commission to 

estimate maximum potential annual 

radiation doses to the public resulting 

from effluent releases. The report must 

be submitted as specified in Section 

50.4, and the time between report 

submittals must be no longer than 12 

months. 

The licensee has proposed an 

amendment to Technical Specification 

5.6.3 to change the submittal date for 

the report from ‘‘within 90 days of 

January 1 of each year’’ to ‘‘prior to May 

1 of each year.’’ Therefore, the licensee 

has requested a one-time exemption 

from the 12-month reporting criteria 

specified in 10 CFR 50.36a(a)(2) for its 

submittal of the 2008 Radioactive 

Effluent Release Report. 

In summary, the exemption does not 

affect the information required to be 

submitted or the time period the report 

covers, only the date the report is 

submitted. 

3.0 Discussion 

Pursuant to 10 CFR 50.12, the 

Commission may, upon application by 

any interested person or upon its own 

initiative, grant exemptions from the 

requirements of 10 CFR Part 50, when 

(1) the exemptions are authorized by 

law, will not present an undue risk to 

public health or safety, and are 

consistent with the common defense 

and security; and (2) when special 

circumstances are present. These 

circumstances include the special 

circumstances that would provide only 

temporary relief from the applicable 

regulation and the licensee or applicant 

has made good faith efforts to comply 

with the regulation. 

Authorized by Law 

This exemption would allow the 

licensee to submit the 2008 Radioactive 

Effluent Release Report prior to May 1, 

2009, which would exceed the report 

submittal requirement of no longer than 

12 months specified in 10 CFR 

50.36a(a)(2). As stated above, 10 CFR 

50.12 allows the NRC to grant 

exemptions from the requirements of 10 

CFR Part 50. The NRC staff has 

determined that granting of the 

licensee’s proposed exemption will not