Federal Register / Vol. 75, No. 16 / Tuesday, January 26, 2010 / Notices 4111 

NATIONAL CREDIT UNION 

ADMINISTRATION 

Sunshine Act; Notice of Agency 

Meeting 

TIME AND DATE: 10 a.m., Friday, January 

29, 2010. 

PLACE: Board Room, 7th Floor, Room 

7047, 1775 Duke Street, Alexandria, VA 

22314–3428. 

STATUS: Open. 

MATTERS TO BE CONSIDERED: 

1. Withdrawal of Final Rule—Part 706 

of NCUA’s Rules and Regulations, 

Unfair or Deceptive Acts or Practices. 

2. Insurance Fund Report. 

RECESS: 11 a.m. 

TIME AND DATE: 11:15 a.m., Friday, 

January 29, 2010. 

PLACE: Board Room, 7th Floor, Room 

7047, 1775 Duke Street, Alexandria, VA 

22314–3428. 

STATUS: Closed. 

MATTERS TO BE CONSIDERED: 

1. Consideration of Supervisory 

Activities. Closed pursuant to 

Exemptions (8), (9)(A)(ii) and 9(B). 

2. Personnel (3). Closed pursuant to 

some or all of the following: Exemptions 

(2), (6) and (9)(A)(ii). 

FOR FURTHER INFORMATION CONTACT: 

Mary Rupp, Secretary of the Board, 

Telephone: 703–518–6304. 

Mary Rupp, 

Board Secretary. 

[FR Doc. 2010–1652 Filed 1–22–10; 4:15 pm] 

BILLING CODE 7535–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0017] 

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) 

is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued from December 

31, 2009 to January 13, 2010. The last 

biweekly notice was published on 

January 12, 2010 (75 FR 1655). 

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 

Title 10 of the Code of Federal 

Regulations (10 CFR), Section 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 (RDB), 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. Written comments may also be 

faxed to the RDB at 301–492–3446. 

Documents may be examined, and/or 

copied for a fee, at the NRC’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 O1F21, 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 by the above 

date, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

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4112 Federal Register / Vol. 75, No. 16 / Tuesday, January 26, 2010 / Notices 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the requestor/ 

petitioner 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 requestor/petitioner 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 requestor/petitioner 

intends to rely in proving the contention 

at the hearing. The requestor/petitioner 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the requestor/petitioner 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 requestor/ 

petitioner to relief. A requestor/ 

petitioner 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, 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 

(72 FR 49139, August 28, 2007). The EFiling 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 an exemption in 

accordance with the procedures 

described below. 

To comply with the procedural 

requirements of E-Filing, at least ten 

(10) days prior to the filing deadline, the 

participant should contact the Office of 

the Secretary by e-mail at 

hearing.docket@nrc.gov, or by telephone 

at (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 (2) advise the 

Secretary that the participant will be 

submitting a request or petition for 

hearing (even in instances in which the 

participant, or its counsel or 

representative, already holds an NRCissued digital ID certificate). Based upon 

this information, the Secretary will 

establish an electronic docket for the 

hearing in this proceeding if the 

Secretary has not already established an 

electronic docket. 

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/ 

apply-certificates.html. System 

requirements for accessing the ESubmittal server are detailed in NRC’s 

‘‘Guidance for Electronic Submission,’’ 

which is available on the agency’s 

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

site-help/e-submittals.html. Participants 

may attempt to use other software not 

listed on the Web site, but should note 

that the NRC’s E-Filing system does not 

support unlisted software, and the NRC 

Meta System Help Desk will not be able 

to offer assistance in using unlisted 

software. 

If a participant is electronically 

submitting a document to the NRC in 

accordance with the E-Filing rule, the 

participant must file the document 

using the NRC’s online, Web-based 

submission form. In order to serve 

documents through EIE, users will be 

required to install a Web browser plugin from the NRC Web site. Further 

information on the Web-based 

submission form, including the 

installation of the Web browser plug-in, 

is available on the NRC’s public Web 

site at http://www.nrc.gov/site-help/esubmittals.html. 

Once a participant has obtained a 

digital ID certificate and a docket has 

been created, the participant 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 documents are 

submitted through the NRC’s E-Filing 

system. To be timely, an electronic 

filing must be submitted to the E-Filing 

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 

E-Filing system also distributes an email 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 using 

the agency’s adjudicatory E-Filing 

system may seek assistance by 

contacting the NRC Meta System Help 

Desk through the ‘‘Contact Us’’ link 

located on the NRC Web site at http:// 

www.nrc.gov/site-help/esubmittals.html, by e-mail at 

MSHD.Resource@nrc.gov, or by a tollfree call at (866) 672–7640. The NRC 

Meta System Help Desk is available 

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

Time, Monday through Friday, 

excluding government holidays. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file an 

exemption request, 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, 

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Federal Register / Vol. 75, No. 16 / Tuesday, January 26, 2010 / Notices 4113 

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. A presiding 

officer, having granted an exemption 

request from using E-Filing, may require 

a participant or party to use E-Filing if 

the presiding officer subsequently 

determines that the reason for granting 

the exemption from use of E-Filing no 

longer exists. 

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, or the 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. 

Petitions for leave to intervene must 

be filed no later than 60 days from 

January 26, 2010. Non-timely filings 

will not be entertained absent a 

determination by the presiding officer 

that the petition or request should be 

granted or the contentions should be 

admitted, based on a balancing of the 

factors specified in 10 CFR 

2.309(c)(1)(i)–(viii). 

For further details with respect to this 

license amendment application, 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 

O1F21, 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. Persons who do not have 

access to ADAMS or who encounter 

problems in accessing the documents 

located in ADAMS, should contact the 

NRC PDR Reference staff at 1–800–397– 

4209, 301–415–4737, or by e-mail to 

pdr.resource@nrc.gov. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units 1, 2, and 3, 

Maricopa County, Arizona 

Date of amendment request: October 

30, 2009. 

Description of amendment request: 

The amendments would revise License 

Condition C.(1) for Units 1 and 3, and 

the Technical Specifications (TS) for all 

three units, to remove requirements no 

longer applicable due to the completion 

of power uprate, replacement of steam 

generators, removal of part-length 

control element assemblies (CEAs), and 

completion of a core protection 

calculator (CPC) upgrade, and to make 

a minor administrative change to the 

nomenclature of the containment sump 

trash racks and screens. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment includes the 

following changes that are considered to be 

administrative and/or editorial changes: 

A. Remove superseded references to 3876 

megawatts thermal (MWt) and related 

information to this value from Unit 1 and 

Unit 3 Operating Licenses and Unit 1, 2, and 

3 Technical Specifications. 

This change is administrative. The change 

only removes the references to 3876 MWt 

and related information to this value and 

leaves the references to 3990 MWt. 

B. Remove references to Part Length 

Control Element Assemblies. 

This change is administrative because it 

only removes references to part length CEAs 

which have been replaced by part strength 

CEAs. 

C. Remove outdated pages and other 

references as a result of the CPC upgrade, and 

adjust the indentation of the logical 

connectors AND and OR in TS 3.2.4, between 

Required Actions B.1, B.2.1, and B.2.2. 

This change is administrative because it 

removes the redundant TS pages identified as 

‘‘(Before CPC Upgrade) or (Before CPCS 

Upgrade)’’ and removes the reference to 

‘‘(After CPC Upgrade) or (After CPCS 

Upgrade)’’ from various TS pages that will be 

renumbered and remain in place. The CPC 

upgrade has been completed. The adjustment 

of the indentation of the logical connectors 

AND and OR in TS 3.2.4 is consistent with 

the Action numbers and with TS 1.2. 

D. Change ‘‘trash racks and screens’’ to 

‘‘strainers.’’ 

This change is administrative. The change 

from ‘‘trash racks and screens’’ to ‘‘strainers’’ 

does not change the intent of the 

Surveillance Requirement 3.5.3.8 to verify, 

by visual inspection, that each [emergency 

core cooling system] ECCS train containment 

sump suction inlet is not restricted by debris 

and the suction inlet strainers show no 

evidence of structural distress or abnormal 

corrosion. 

E. Delete inspection requirements for 

Steam Generators (SG) with Alloy 600 MA 

tubes. 

This change is administrative because APS 

[Arizona Public Service Company] has 

completed the SG replacement project which 

removed all SGs containing Alloy 600 MA 

tubes. 

As discussed above, the proposed 

amendment involves administrative and/or 

editorial changes only. The proposed 

amendment does not impact any accident 

initiators, analyzed events, or assumed 

mitigation of accident or transient events. 

The proposed changes do not involve the 

addition or removal of any equipment or any 

design changes to the facility. The proposed 

changes do not affect any plant operations, 

design function, or analysis that verifies the 

capability of structures, systems, and 

components (SSCs) to perform a design 

function. The proposed changes do not 

change any of the accidents previously 

evaluated in the UFSAR [updated final safety 

analysis report]. The proposed changes do 

not affect SSCs, operating procedures, and 

administrative controls that have the 

function of preventing or mitigating any of 

these accidents. 

Therefore, the proposed changes do not 

represent a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

As stated in response to standard 1, the 

proposed amendment only involves 

administrative and/or editorial changes. No 

actual plant equipment or accident analyses 

will be affected by the proposed changes. The 

proposed changes will not change the design 

function or operation of any SSCs. The 

proposed changes will not result in any new 

failure mechanisms, malfunctions, or 

accident initiators not considered in the 

design and licensing bases. The proposed 

amendment does not impact any accident 

initiators, analyzed events, or assumed 

mitigation of accident or transient events. 

Therefore, this proposed change does not 

create the possibility of an accident of a new 

or different kind than previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

As stated in response to standard 1, the 

proposed amendment only involves 

administrative and/or editorial changes. The 

proposed change does not involve any 

physical changes to the plant or alter the 

manner in which plant systems are operated, 

maintained, modified, tested, or inspected. 

The proposed change does not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The safety analysis 

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4114 Federal Register / Vol. 75, No. 16 / Tuesday, January 26, 2010 / Notices 

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. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on that 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the request 

for amendments involves no significant 

hazards consideration. 

Attorney for licensee: Michael G. 

Green, Senior Regulatory Counsel, 

Pinnacle West Capital Corporation, P.O. 

Box 52034, Mail Station 8695, Phoenix, 

Arizona 85072–2034. 

NRC Branch Chief: Michael T. 

Markley. 

Carolina Power & Light Company, 

Docket Nos. 50–325 and 50–324, 

Brunswick Steam Electric Plant, Units 1 

and 2, Brunswick County, North 

Carolina 

Date of amendments request: October 

27, 2009. 

Description of amendments request: 

The proposed amendments would 

modify technical specifications (TSs) 

requirements related to primary 

containment isolation instrumentation 

in accordance with the Nuclear 

Regulatory Commission-approved 

Technical Specification Task Force 

(TSTF), Improved Standard Technical 

Specifications change traveler, TSTF– 

306, Revision 2, ‘‘Add action to LCO 

3.3.6.1 to give option to isolate the 

penetration.’’ The proposed amendment 

would revise TS Section 3.3.6.1, 

‘‘Primary Containment Isolation 

Instrumentation,’’ by adding an 

ACTIONS note allowing intermittent 

opening, under administrative control, 

of penetration flow paths that are 

isolated. Additionally, the traversing incore probe (TIP) system would be added 

as a separate isolation function with an 

associated Required Action to isolate 

the penetration within 24 hours rather 

than immediately initiating a unit 

shutdown. 

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 addition of the note that the 

penetration flow path may be unisolated 

under administrative control simply provides 

consistency with what is already allowed 

elsewhere in TSs. The isolation function of 

the TIP valves is mitigative, and does not 

create any increased possibility of an 

accident. Also, the operation of the manual 

shear valves is unaffected by this activity. 

The ability to manually isolate the TIP 

system by either the normal isolation ball 

valves or the shear valves would be 

unaffected by the inoperable 

instrumentation. The Required Actions and 

their associated Completion Times are not 

initiating conditions for any accident 

previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as result of the proposed changes. 

All systems, structures, and components 

previously required for the mitigation of a 

transient remain capable of fulfilling their 

intended design functions. The proposed 

changes have no adverse effects on any 

safety-related system or component and do 

not challenge the performance or integrity of 

any safety-related system. As a result no new 

failure modes are being introduced. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change will not affect the 

operation of plant equipment or the function 

of any equipment assumed in the accident 

analysis. The allowance to unisolate a 

penetration flow path will not have a 

significant effect on the margin of safety 

because the penetration flow path can be 

isolated manually, if needed. This change 

simply provides consistency with what is 

already allowed elsewhere in TSs. The 

option to isolate a TIP penetration will 

ensure the penetration will perform as 

designed in the accident analysis. The ability 

to manually isolate the TIP system is 

unaffected by the inoperable 

instrumentation. The proposed change does 

not impact any safety analysis assumptions 

or results. 

Therefore, the proposed change does not 

result in 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: David T. 

Conley, Associate General Counsel II— 

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, NC 27602. 

NRC Branch Chief: Thomas H. Boyce. 

Dominion Nuclear Connecticut Inc., et 

al., Docket No. 50–423, Millstone Power 

Station, Unit No. 3, New London 

County, Connecticut 

Date of amendment request: 

November 23, 2009. 

Description of amendment request: 

The proposed license amendment 

request would revise the Millstone 

Power Station, Unit 3 Technical 

Specification (TS) 6.8.4.g, ‘‘Steam 

Generator Program,’’ to exclude a 

portion of the tubes below the top of the 

steam generator tubesheet from periodic 

steam generator tube inspections. This 

request would also remove reference to 

the previous Cycle 13 interim alternate 

repair criteria. 

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 previously analyzed accidents are 

initiated by the failure of plant structures, 

systems, or components. The proposed 

change that alters the steam generator 

inspection criteria and the steam generator 

inspection reporting criteria does not have a 

detrimental impact on the integrity of any 

plant structure, system, or component that 

initiates an analyzed event. The proposed 

change will not alter the operation of, or 

otherwise increase the failure probability of 

any plant equipment that initiates an 

analyzed accident. 

Of the applicable accidents previously 

evaluated, the limiting transients with 

consideration to the proposed change to the 

steam generator tube inspection and repair 

criteria are the steam generator tube rupture 

(SGTR) event and the feedline break (FLB) 

postulated accidents. 

During the SGTR event, the required 

structural integrity margins of the steam 

generator tubes and the tube-to-tubesheet 

joint over the H* distance will be 

maintained. Tube rupture in tubes with 

cracks within the tubesheet is precluded by 

the constraint provided by the tube-totubesheet joint. This constraint results from 

the hydraulic expansion process, thermal 

expansion mismatch between the tube and 

tubesheet, and from the differential pressure 

between the primary and secondary side. 

Based on this design, the structural margins 

against burst, as discussed in Regulatory 

Guide (RG) 1.121, ‘‘Bases for Plugging 

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Degraded [pressurized-water reactor] PWR 

Steam Generator Tubes,’’ are maintained for 

both normal and postulated accident 

conditions. 

The proposed change has no impact on the 

structural or leakage integrity of the portion 

of the tube outside of the tubesheet. The 

proposed change maintains structural 

integrity of the steam generator tubes and 

does not affect other systems, structures, 

components, or operational features. 

Therefore, the proposed change results in no 

significant increase in the probability of the 

occurrence of a SGTR accident. 

At normal operating pressures, leakage 

from primary water stress corrosion cracking 

below the proposed limited inspection depth 

is limited by both the tube-to-tubesheet 

crevice and the limited crack opening 

permitted by the tubesheet constraint. 

Consequently, negligible normal operating 

leakage is expected from cracks within the 

tubesheet region. The consequences of an 

SGTR event are affected by the primary-to 

secondary leakage flow during the event. 

However, primary-to-secondary leakage flow 

through a postulated broken tube is not 

affected by the proposed changes since the 

tubesheet enhances the tube integrity in the 

region of the hydraulic expansion by 

precluding tube deformation beyond its 

initial hydraulically expanded outside 

diameter. Therefore, the proposed changes do 

not result in a significant increase in the 

consequences of a SGTR. 

The consequences of a steam line break 

(SLB) are also not significantly affected by 

the proposed changes. During a SLB 

accident, the reduction in pressure above the 

tubesheet on the shell side of the steam 

generator creates an axially uniformly 

distributed load on the tubesheet due to the 

reactor coolant system pressure on the 

underside of the tubesheet. The resulting 

bending action constrains the tubes in the 

tubesheet thereby restricting primary-tosecondary leakage below the midplane. 

Primary-to-secondary leakage from tube 

degradation in the tubesheet area during the 

limiting accident (i.e., a SLB) is limited by 

flow restrictions. These restrictions result 

from the crack and tube-to-tubesheet contact 

pressures that provide a restricted leakage 

path above the indications and also limit the 

degree of potential crack face opening as 

compared to free span indications. 

The leakage factor of 2.49 for Millstone 

Power Station Unit 3 (MPS3), for a postulated 

SLB/FLB, has been calculated as shown in 

Table RA124–2 of Enclosure 5. The leakage 

factor of 2.49 is a bounding value for all 

steam generators, both hot and cold legs, in 

Table RA124–2. Specifically, for the 

condition monitoring (CM) assessment, the 

component of leakage from the prior cycle 

from below the H* distance will be 

multiplied by a factor of 2.49 and added to 

the total leakage from any other source and 

compared to the allowable accident induced 

leakage limit. For the operational assessment 

(OA), the difference in the leakage between 

the allowable accident induced leakage and 

the accident induced leakage from sources 

other than the tubesheet expansion region 

will be divided by 2.49 and compared to the 

observed operational leakage. 

The probability of a SLB is unaffected by 

the potential failure of a steam generator tube 

as the failure of the tube is not an initiator 

for a SLB event. SLB leakage is limited by 

leakage flow restrictions resulting from the 

leakage path above potential cracks through 

the tube-to-tubesheet crevice. The leak rate 

during postulated accident conditions 

(including locked rotor) has been shown to 

remain within the accident analysis 

assumptions for all axial and or 

circumferentially orientated cracks occurring 

13.1 inches below the top of the tubesheet. 

The accident induced leak rate limit is 1.0 

gpm. The technical specification (TS) 

operational leak rate is 150 gpd (0.1 gpm) 

through any one steam generator. 

Consequently, there is significant margin 

between accident leakage and allowable 

operational leakage. The SLB/FLB leak rate 

ratio is only 2.49 resulting in significant 

margin between the conservatively estimated 

accident leakage and the allowable accident 

leakage (1.0 gpm). 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

Response: No. 

The proposed change that alters the steam 

generator inspection criteria and the steam 

generator inspection reporting criteria does 

not introduce any new equipment, create 

new failure modes for existing equipment, or 

create any new limiting single failures. Plant 

operation will not be altered, and all safety 

functions will continue to perform as 

previously assumed in accident analyses. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the change involve a significant 

reduction in a margin of safety? 

Response: No. 

The proposed change that alters the steam 

generator inspection criteria and the steam 

generator inspection reporting criteria 

maintains the required structural margins of 

the steam generator tubes for both normal 

and accident conditions. Nuclear Energy 

Institute (NEI) 97–06, Revision 2, ‘‘Steam 

Generator Program Guidelines’’ and RG 1.121, 

are used as the bases in the development of 

the limited tubesheet inspection depth 

methodology for determining that steam 

generator tube integrity considerations are 

maintained within acceptable limits. RG 

1.121 describes a method acceptable to the 

Nuclear Regulatory Commission (NRC) for 

meeting General Design Criteria (GDC) 14, 

‘‘Reactor Coolant Pressure Boundary,’’ GDC 

15, ‘‘Reactor Coolant System Design,’’ GDC 

31, ‘‘Fracture Prevention of Reactor Coolant 

Pressure Boundary,’’ and GDC 32, ‘‘Inspection 

of Reactor Coolant Pressure Boundary,’’ by 

reducing the probability and consequences of 

a SGTR. RG 1.121 concludes that by 

determining the limiting safe conditions for 

tube wall degradation the probability and 

consequences of a SGTR are reduced. This 

RG uses safety factors on loads for tube burst 

that are consistent with the requirements of 

Section III of the American Society of 

Mechanical Engineers (ASME) Code. 

For axially oriented cracking located 

within the tubesheet, tube burst is precluded 

due to the presence of the tubesheet. For 

circumferentially oriented cracking, 

Westinghouse Electric Company, LLC 

(Westinghouse) report WCAP–1 7071 -P, ‘‘H*: 

Alternate Repair Criteria for the Tubesheet 

Expansion Region in Steam Generators with 

Hydraulically Expanded Tubes (Model F),’’ 

defines a length of degradation free expanded 

tubing that provides the necessary resistance 

to tube pullout due to the pressure induced 

forces, with applicable safety factors applied. 

Application of the limited hot and cold leg 

tubesheet inspection criteria will preclude 

unacceptable primary-to-secondary leakage 

during all plant conditions. The methodology 

for determining leakage provides for large 

margins between calculated and actual 

leakage values in the proposed limited 

tubesheet inspection depth criteria. 

Therefore, the proposed change does not 

involve a significant reduction in any margin 

of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Lillian M. 

Cuoco, Senior Counsel, Dominion 

Resource Services, Inc., 120 Tredegar 

Street, RS–2, Richmond, VA 23219. 

NRC Branch Chief: Harold K. 

Chernoff. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit No. 2, 

Westchester County, New York 

Date of amendment request: 

November 19, 2009. 

Description of amendment request: 

The proposed change will correct 

identified non-conservatisms in 

Technical Specification 5.5.9 

‘‘Ventilation Filter Testing Program’’ by 

modifying the charcoal testing criteria to 

account for the 95% charcoal efficiency 

assumed for elemental iodine in the 

accident analyses for alternate source 

term. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The proposed change revises testing 

acceptance criteria for the existing Indian 

Point 2 Control Room filtration system in 

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Technical Specification (TS) 5.5.9 

‘‘Ventilation Filter Testing Program’’ to reflect 

current assumptions of iodine removal in 

accident dose calculations. The revised 

testing criteria does not add equipment or 

change the process for taking the test sample 

and only changes the test in the laboratory 

to be more restrictive. Therefore it cannot 

increase the probability of an accident 

occurring. The revised testing criteria is more 

stringent and therefore does not increase the 

consequences of an accident since it is more 

capable of mitigating control room doses and 

is consistent with existing analyses. 

Therefore the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. The proposed change revises the 

testing acceptance criteria for the existing 

Control Room filtration system. The 

proposed change does not involve 

installation of new equipment, modification 

of existing equipment, or result in a change 

to the way that the equipment or facility is 

operated so that no new equipment failure 

modes are introduced. Therefore the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

No. The proposed change revises the 

testing acceptance criteria for the existing 

Control Room filtration system. There is no 

change to the design requirements or the 

surveillance interval. The proposed change 

reflects the accident analysis dose calculation 

assumptions that assumed increased iodine 

removal. The factor of safety applied to the 

testing acceptance criteria remains the same. 

The new acceptance criterion is well within 

the system design capabilities. 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: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Nancy L. Salgado. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–286, Indian Point 

Nuclear Generating Unit No. 3 (IP3), 

Westchester County, New York 

Date of amendment request: 

December 15, 2009, as supplemented on 

December 22, 2009, January 4, 2010, and 

January 11, 2010. 

Description of amendment request: 

The proposed amendment would allow 

a one-time extension of the 72-hour 

completion time of Technical 

Specification (TS) 3.7.5, Condition B, 

Action B.1 ‘‘Restore AFW [auxiliary 

feedwater] train to OPERABLE status’’ 

by 34 hours. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The proposed change revises the 

allowed outage time (AOT) for the steam 

driven Auxiliary Boiler Feedwater Pump 

(ABFP) on a one time basis. Revising the 

AOT is not an accident initiator since an 

ABFP is a mitigating system. Therefore the 

proposed changes do not increase the 

probability of an accident occurring. The 

proposed AOT change is a one time increase 

that will allow repairs without the transient 

of shutdown. The plant is designed for single 

failure and recognizes that inoperability for 

short periods does not cause a significant 

increase in the consequences of an accident. 

The one time increase in this outage time is 

compensated with measures to reduce the 

potential need for the ABFP and the effects 

of events that could require the pump. 

Therefore the increase does not significantly 

increase the consequences of an accident. 

Therefore the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

No. The proposed change revises the 

allowed outage time for the ABFP on a one 

time basis. The proposed change does not 

involve installation of new equipment or 

modification of existing equipment, so no 

new equipment failure modes are introduced. 

The proposed revision is not a change to the 

way that the equipment or facility is operated 

or analyzed and no new accident initiators 

are created. Therefore the proposed change 

does not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

No. The reduction in the margin of safety 

associated with continued IP3 operation with 

Auxiliary Boiler Feedwater (ABF) pump 32 

out of service during a 34 hour period 

beyond current allowed outage time is 

represented by an increase of approximately 

50 percent in the allowed outage time. This 

change in the margin of safety has been 

compensated for by specific compensatory 

measures to reduce the potential need for the 

pump and to address postulated events that 

could require the pump. The increase in core 

damage frequency (CDF) associated with 

continued IP3 operation with ABFP 32 out of 

service for a duration of 106 hours which 

represents a 34 hour period beyond the 

current allowed outage time is 3.9E–5 per 

reactor year (ry). This results in an 

incremental conditional core damage 

probability (ICCDP) of 4.8E–07, which is 

below the ICCDP guidance threshold of 5E– 

07 identified in NRC Inspection Manual Part 

9900. The ICCDP includes risk due to 

external events due to seismic, fire, and 

flood. The increase in large early release 

frequency (LERF) was estimated as 4.2E–7/ry 

(including external events), which results in 

an incremental conditional large early release 

probability (ICLERP) of 5.1E–9. 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: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Nancy L. Salgado. 

Entergy Nuclear Operations, Inc., 

Docket Nos. 50–247 and 50–286, Indian 

Point Nuclear Generating Unit Nos. 2 

and 3, Westchester County, New York 

Date of amendment request: 

November 17, 2009. 

Description of amendment request: 

The proposed change will correct 

identified non-conservatisms in the 

calculation of Emergency Diesel 

Generator (EDG) air receiver pressure 

requirements for Technical 

Specification (TS) 3.8.3. In addition, the 

proposed change will modify the 

number of normal EDG starts the air 

receiver is capable of providing as listed 

in the Final Safety Analysis Report. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The proposed change revises the 

pressure at which the Emergency Diesel 

[G]enerator (EDG) air receiver is required to 

be kept to meet surveillance requirements, 

revises the minimum EDG air receiver 

pressure required for one start of the EDG, 

and changes the number of normal starts in 

the air receiver. Revising the air receiver 

upper and lower pressure limits and 

reducing the number of starts in the air 

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receiver are not accident initiators since an 

EDG is a mitigating system. Therefore the 

proposed changes do not increase the 

probability of an accident occurring. The 

proposed changes will assure that each EDG 

is capable of starting consistent with 

assumed accident analyses. These analyses 

assume that an EDG starts the first time and 

accident analyses do not credit subsequent 

starts. The proposed new TS limits on the 

EDG air receiver will assure that air pressure 

is adequate to assure one attempt to start the 

EDG is available at the lower limit and will 

provide additional normal starts at the upper 

pressure established in the surveillance. 

Establishing acceptance criteria that replace 

non conservative criteria and assure the 

design bases is met assures the capability of 

equipment to mitigate accident conditions. 

Therefore the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

No. The proposed change revises the 

pressure limit for the air receiver to initiate 

an alarm for low pressure, revises the lower 

pressure limit that must be maintained to 

assure that air is sufficient for at least one 

EDG start and revises the number of normal 

starts in the air receiver based on the revised 

calculations. The proposed change does not 

involve installation of new equipment or 

modification of existing equipment, so no 

new equipment failure modes are introduced. 

The proposed revision to the air receiver 

pressure limits and minimum air receiver 

EDG starts is also is [sic] not a change to the 

way that the equipment or facility is operated 

or analyzed and no new accident initiators 

are created. 

Therefore the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

No. The conduct of surveillance tests, the 

conditions for failure of those tests and the 

number of EDG starts in the air receiver are 

means of assuring that the equipment is 

capable of maintaining the margin of safety 

established in the safety analyses for the 

facility. The proposed change in the EDG 

surveillance test acceptance criteria is 

consistent with values assumed in existing 

safety analyses which assume one start 

attempt for each EDG. The requirement for a 

minimum air pressure in the EDG air start 

receiver assures that there will be adequate 

air to allow at least one EDG start attempt 

which meets the intent of the existing TS. 

The reduction in the number of starts 

maintained in the air receiver does not affect 

the margins in accident analyses for this 

reason and because an EDG failure to start 

would reduce the air pressure below that 

required for one start before the overcrank 

timer would lock out a further start attempt. 

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: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Nancy L. Salgado. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–333, James A. FitzPatrick 

Nuclear Power Plant, Oswego County, 

New York 

Date of amendment request: 

November 23, 2009. 

Description of amendment request: 

The proposed amendment would 

modify the Technical Specification (TS) 

5.5.7, Inservice Testing Program, by 

replacing the references from the 

American Society of Mechanical 

Engineers (ASME) Boiler and Pressure 

Vessel Code to the current code of 

record, the ASME Operation and 

Maintenance Nuclear Power Plants 

Code (ASME OM Code), the code of 

record for the James A. FitzPatrick 

Nuclear Power Plant (JAF) Inservice 

Testing Program for Inservice Testing 

Program. This is an administrative 

amendment to maintain the TS current 

with the NRC accepted code of record 

for JAF. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Will operation of the facility in 

accordance with this proposed change 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated? 

Response: No. 

The proposed TS changes are nontechnical, and are provided for consistency. 

There is no plant change involved, and thus, 

proposed TS changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Will operation of the facility in 

accordance with this proposed change create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed TS changes are nontechnical, i.e., there is no plant change 

involved, and thus, do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Will operation of the facility in 

accordance with this proposed change 

involve a significant reduction in a margin of 

safety? 

Response: No. 

The proposed TS changes are nontechnical, i.e., there is no plant change 

involved. The changes are consistent with 

the regulations, and only update the TS to 

refer to the current code of reference. No 

design or safety margin is involved. 

Therefore, the proposed changes do not 

involve a significant reduction in any margin 

of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Nancy L. Salgado. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Steam Electric Station 

(CPSES), Units 1 and 2, Somervell 

County, Texas 

Date of amendment request: October 

26, 2009. 

Description of amendment request: 

The proposed change will revise 

Technical Specification (TS) 3.8.1 

entitled ‘‘AC Sources—Operating’’ to 

extend, on a one-time basis, the 

allowable Completion Time (CT) of 

Required Action A.3 for one offsite 

circuit inoperable, from 72 hours to 14 

days. This change is only applicable to 

startup transformer (ST) XST2 and will 

expire on March 1, 2011. This change is 

needed to allow sufficient time to make 

final terminations as part of a plant 

modification to facilitate connection of 

either ST XST2 or the spare ST to the 

Class 1E buses. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change will revise the CT for 

the loss of one offsite source from 72 hours 

to 14 days. The proposed one-time extension 

of the CT for the loss of one offsite power 

circuit does not significantly increase the 

probability of an accident previously 

evaluated. The startup transformers are not 

the initiator of any previously evaluated 

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accidents involving a loss of offsite power 

(LOOP). 

The TS will continue to require equipment 

that will power safety related equipment 

necessary to perform any required safety 

function. The one-time extension of the CT 

to 14 days does not affect the design of the 

STs, the interface of the STs with other plant 

systems, the operating characteristic of the 

STs, or the reliability of the STs. 

Per Regulatory Guide (RG) 1.177, the risk 

acceptance guideline presented in RG 1.174 

shows that Unit 1 met all the risk acceptance 

guidelines for delta core damage frequency 

(CDF), delta large early release frequency 

(LERF), incremental conditional core damage 

probability (ICCDP), and incremental 

conditional large early release probability 

(ICLERP). [CPSES,] Unit 2 met the same risk 

acceptance guidelines of delta LERF and 

ICLERP; however, the delta CDF and ICCDP 

were above the acceptance value. Since the 

increase above the regulatory guidance is 

small, and the risk reduction measures 

quantitatively addressed, the values for Unit 

2 delta CDF and ICCDP would fall below the 

regulatory guidance as well as decrease the 

other risk metrics for both Units. 

The consequence of a LOOP event has been 

evaluated in the CPNPP [Comanche Peak 

Steam Electric Station] Final Safety Analysis 

Report [ ] and the Station Blackout 

evaluation. Increasing the CT for one offsite 

power source on a one-time basis from 72 

hours to 14 days does not increase the 

consequences of a LOOP event nor change 

the evaluation of LOOP events. 

Therefore, the proposed change does 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 does not result in a 

change in the manner in which the electrical 

distribution subsystems provide plant 

protection. The proposed change will only 

affect the time allowed to restore the 

operability of the offsite power source 

through a startup transformer. The proposed 

change does not affect the configuration or 

operation of the plant. The proposed change 

to the CT will facilitate installation of a plant 

modification which will improve plant 

design and will eliminate the necessity to 

shut down both Units if [ST] XST2 fails or 

requires maintenance that goes beyond the 

current TS CT of 72 hours. This change will 

improve the long-term reliability of the 

345kV [kiloVolt] offsite circuit STs which are 

common to both CPNPP Units. 

There are no changes to the STs or the 

supporting systems operating characteristics 

or conditions. The change to the CT does not 

change any existing accident scenarios, nor 

create any new or different accident 

scenarios. In addition, the change does not 

impose any new or different requirements or 

eliminate any existing requirements. The 

change does not alter any of the assumptions 

made in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change does not affect the 

acceptance criteria for any analyzed event 

nor is there a change to any safety limit. The 

proposed change does not alter the manner 

in which safety limits, limiting safety system 

settings, or limiting conditions for operation 

are determined. Neither the safety analyses 

nor the safety analysis acceptance criteria are 

affected by this change. The proposed change 

will not result in plant operation in a 

configuration outside the current design 

basis. The proposed activity only increases, 

for a one-time pre-planned occurrence, the 

period when the plant may operate with one 

offsite power source. The margin of safety is 

maintained by maintaining the ability to 

safely shut down the plant and remove 

residual heat. 

Therefore, the proposed change does not 

involve a reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Timothy P. 

Matthews, Esq., Morgan, Lewis and 

Bockius, 1800 M Street, NW., 

Washington, DC 20036. 

NRC Branch Chief: Michael T. 

Markley. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of amendment request: 

November 4, 2009. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specifications (TSs) to: (1) 

Delete TS 4.0.5, which pertains to 

surveillance requirements (SRs) for 

inservice inspection (ISI) and inservice 

testing (IST) of American Society of 

Mechanical Engineers (ASME) Boiler 

and Pressure Vessel Code (Code) Class 

1, 2 and 3 components; (2) add a new 

TS for the IST Program to Section 6.0, 

‘‘Administrative Controls,’’ of the TSs; 

(3) change TSs that currently reference 

TS 4.0.5 to reference the IST Program or 

ISI Program, as applicable; and (4) 

revise TS 6.10.3.h to reflect the deletion 

of the ISI Program from the TSs. The 

new TS for the IST Program, TS 6.8.4.i, 

will indicate that the program will 

include testing frequencies applicable to 

the ASME Code for Operation and 

Maintenance of Nuclear Power Plants 

(OM Code), replacing the current 

reference to Section XI of the ASME 

Code specified in TS 4.0.5. In addition, 

TS 6.8.4.i would revise the 

requirements, currently contained in TS 

4.0.5, regarding the applicability of the 

surveillance interval extension 

provisions of SR 4.0.2. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes revise TS 4.0.5, 

Surveillance Requirements for Inservice 

Inspections and Testing of ASME Code 

Components, for consistency with 10 CFR 

50.55a(f)(4) requirements regarding inservice 

testing of pumps and valves. The proposed 

change incorporates revisions to the ASME 

OM Code and clarifies testing frequency 

requirements for testing pumps and valves. 

The proposed change also relocates the ISI 

and IST Programs consistent with NUREG– 

1433. A commitment is made to maintain 

[Generic Letter (GL)] 88–01 inspection 

requirements in the ISI Program. 

The proposed changes do not impact any 

accident initiators or analyzed events or 

assumed mitigation of accident or transient 

events. They do not involve the addition or 

removal of any equipment, or any design 

changes to the facility. 

Therefore, the proposed changes do not 

represent a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not involve a 

modification to the physical configuration of 

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

installed) or change in the methods 

governing normal plant operation. The 

proposed change will not impose any new or 

different requirements or introduce a new 

accident initiator, accident precursor, or 

malfunction mechanism. Therefore, this 

proposed change does not create the 

possibility of an accident of a different kind 

than previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes revise and relocate 

TS 4.0.5, Surveillance Requirements for 

Inservice Inspections and Testing of ASME 

Code Components, for consistency with (1) 

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

regarding the inservice testing of pumps and 

valves and (2) NUREG–1433. The proposed 

change updates references to the ASME OM 

Code, clarifies testing frequency 

requirements for testing pumps and valves, 

and relocates the IST Program to Section 6.0 

of TS, and the ISI Program to a licensee 

controlled document. The safety function of 

the affected pumps and valves will be 

maintained; the programs will continue to be 

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Federal Register / Vol. 75, No. 16 / Tuesday, January 26, 2010 / Notices 4119 

implemented with the required regulations 

and codes. A commitment is made to 

maintain GL 88–01 inspection requirements 

in the ISI Program; there will be no change 

to these requirements. 

Therefore, this 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: Vincent 

Zabielski, PSEG Nuclear LLC–N21, P.O. 

Box 236, Hancocks Bridge, NJ 08038. 

NRC Branch Chief: Harold K. 

Chernoff. 

PSEG Nuclear LLC, Docket No. 50–354, 

Hope Creek Generating Station, Salem 

County, New Jersey 

Date of amendment request: 

December 1, 2009. 

Description of amendment request: 

The proposed amendment would revise 

the Technical Specifications (TSs) to 

change the required frequency of testing 

control rod scram times from ‘‘at least 

once per 120 days of POWER 

OPERATION’’ to ‘‘at least once per 200 

days of POWER OPERATION.’’ This 

change is based on TS Task Force 

(TSTF) change traveler TSTF–460, 

Revision 0, ‘‘Control Rod Scram Time 

Testing Frequency.’’ TSTF–460 has been 

approved generically by the Nuclear 

Regulatory Commission (NRC) for 

incorporation into the boiling water 

reactor (BWR) Standard TS (STS); 

NUREG–1433 (BWR/4) and NUREG– 

1434 (BWR/6). The NRC staff published 

a notice announcing the availability of 

this proposed TS change using the 

consolidated line item improvement 

process (CLIIP) in the Federal Register 

on August 23, 2004 (69 FR 51864). Since 

Hope Creek Generating Station has not 

adopted the STS, the licensee has 

proposed variations from the CLIIP to 

ensure consistency with NUREG–1433, 

Revision 3, ‘‘Standard Technical 

Specifications, General Electric Plants, 

BWR/4.’’ The changes to align with 

NUREG–1433 involve the adoption of a 

revised control rod scram time test 

methodology and an establishment of a 

category of operable but ‘‘slow’’ control 

rods. 

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 extend the 

frequency and revise the evaluation 

methodology for control rod scram times, and 

identify a new category of ‘‘slow’’ control rods 

for assessing control rod operability. The 

frequency of control rod scram testing is not 

an initiator of any accident previously 

evaluated. The frequency of surveillance 

testing does not affect the ability to mitigate 

any accident previously evaluated, because 

the tested component is still required to be 

operable. The proposed evaluation 

methodology is consistent with industry 

approved methods and ensures control rod 

operability requirements for the number and 

distribution of operable, slow, and stuck 

control rods [and] continue[s] to satisfy 

scram reactivity rate assumptions used in 

plant safety analysis. 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 changes do not involve any 

physical alteration of the plant (no new or 

different type of equipment is being 

installed) and do not involve a change in the 

design, normal configuration, or basic 

operation of the plant. The proposed changes 

do not introduce any new accident initiators. 

The proposed changes do not involve 

significant changes in the fundamental 

methods governing normal plant operation 

and do not require unusual or uncommon 

operator actions. The proposed changes 

provide assurance that the plant will not be 

operated in a mode or condition that violates 

the assumptions or initial conditions in the 

plant safety analyses and that [structures, 

systems and components] remain capable of 

performing their intended safety functions as 

assumed in the same analyses. Consequently, 

the response of the plant and the plant 

operator to postulated events will not be 

significantly different. Therefore, the 

proposed TS change does not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

Margin of safety is related to confidence in 

the ability of the fission product barriers to 

perform their design functions during and 

following an accident situation. The 

proposed changes address control rod scram 

test performance and acceptance criteria as 

well as control rod operability requirements. 

The scram test acceptance criteria and 

control rod operability restrictions are based 

on industry approved methodology and will 

continue to ensure control rod scram design 

functions and reactivity insertion 

assumptions used in plant safety analyses 

continue to be protected. The proposed 

changes also extend the frequency of testing 

control rod scram times while at-power from 

120 days to 200 days. The proposed change 

continues to test the control rod scram time 

to ensure the assumptions in the plant safety 

analysis are protected. The demonstrated 

reliability of the control rod scram function 

justifies the extension of the surveillance 

frequency. Therefore, the proposed changes 

do not involve a significant reduction in the 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, 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: Vincent 

Zabielski, PSEG Nuclear LLC–N21, P.O. 

Box 236, Hancocks Bridge, NJ 08038. 

NRC Branch Chief: Harold K. 

Chernoff. 

Virginia Electric and Power Company, 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Unit Nos. 1 and 2, 

Louisa County, Virginia 

Date of amendment request: 

December 16, 2009. 

Description of amendment request: 

The amendment would revise the 

Technical Specifications (TS) to adopt 

Nuclear Regulatory Commission (NRC)- 

approved Revision 2 to Technical 

Specification Task Force (TSTF) 

Standard Technical Specification 

Change Traveler, TSTF–427, 

‘‘Allowance for Non Technical 

Specification Barrier Degradation on 

Support System Operability.’’ The 

proposed amendment would modify the 

requirements for unavailable barriers by 

adding Limiting Condition for 

Operation 3.0.9. 

The NRC staff published a notice of 

opportunity for comment in the Federal 

Register on June 2, 2006 (71 FR 32145), 

on possible amendments adopting 

TSTF–427, including a model safety 

evaluation and model no significant 

hazards consideration (NSHC) 

Determination, using the consolidated 

line-item improvement process. The 

NRC staff subsequently issued a notice 

of availability of the models for 

referencing in license amendment 

applications in the Federal Register 

October 3, 2006 (71 FR 58444). The 

licensee affirmed the applicability of the 

following NSHC determination in its 

application dated December 16, 2009. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration is presented 

below: 

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4120 Federal Register / Vol. 75, No. 16 / Tuesday, January 26, 2010 / Notices 

Criterion 1—The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability of Consequences of an Accident 

Previously Evaluated 

The proposed change allows a delay time 

for entering a supported system technical 

specification (TS) when the inoperability is 

due solely to an unavailable hazard barrier if 

risk is assessed and managed. The postulated 

initiating events which may require a 

functional barrier are limited to those with 

low frequencies of occurrence, and the 

overall TS system safety function would still 

be available for the majority of anticipated 

challenges. Therefore, the probability of an 

accident previously evaluated is not 

significantly increased, if at all. The 

consequences of an accident while relying on 

the allowance provided by proposed LCO 

3.0.9 are no different than the consequences 

of an accident while relying on the TS 

required actions in effect without the 

allowance provided by proposed LCO 3.0.9. 

Therefore, the consequences of an accident 

previously evaluated are not significantly 

affected by this change. The addition of a 

requirement to assess and manage the risk 

introduced by this change will further 

minimize possible concerns. Therefore, this 

change does not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Previously 

Evaluated 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed). 

Allowing delay times for entering supported 

system TS when inoperability is due solely 

to an unavailable hazard barrier, if risk is 

assessed and managed, will not introduce 

new failure modes or effects and will not, in 

the absence of other unrelated failures, lead 

to an accident whose consequences exceed 

the consequences of accidents previously 

evaluated. The addition of a requirement to 

assess and manage the risk introduced by this 

change will further minimize possible 

concerns. Thus, this change does not create 

the possibility of a new or different kind of 

accident from an accident previously 

evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety 

The proposed change allows a delay time 

for entering a supported system TS when the 

inoperability is due solely to an unavailable 

barrier, if risk is assessed and managed. The 

postulated initiating events which may 

require a functional barrier are limited to 

those with low frequencies of occurrence, 

and the overall TS system safety function 

would still be available for the majority of 

anticipated challenges. The risk impact of the 

proposed TS changes was assessed following 

the three-tiered approach recommended in 

RG 1.177. A bounding risk assessment was 

performed to justify the proposed TS 

changes. This application of LCO 3.0.9 is 

predicated upon the licensee’s performance 

of a risk assessment and the management of 

plant risk. The net change to the margin of 

safety is insignificant as indicated by the 

anticipated low levels of associated risk 

(ICCDP [incremental conditional core damage 

probability] and ICLERP [incremental 

conditional large early release probability]) as 

shown in Table 1 of Section 3.1.1 in the 

Safety Evaluation [published in the Federal 

Register on October 3, 2006 (71 FR 58444)]. 

Therefore, this change does not involve a 

significant reduction in a margin of safety. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: Lillian M. 

Cuoco, Senior Counsel, Dominion 

Resources Services, Inc., 120 Tredegar 

Street, RS–2, Richmond, VA 23219. 

NRC Branch Chief: Gloria Kulesa. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: October 

10, 2009. 

Description of amendment request: 

The proposed changes will revise 

Technical Specification (TS) 3.1.7, ‘‘Rod 

Position Indication,’’ TS 3.2.1, ‘‘Heat 

Flux Hot Channel Factor (FQ(Z)) (FQ

Methodology),’’ TS 3.2.2, ‘‘Nuclear 

Enthalpy Rise Hot Channel Factor 

(FNDH), TS 3.2.4, ‘‘Quadrant Power Tilt 

Ratio (QPTR),’’ and TS 3.3.1, ‘‘Reactor 

Trip System (RTS) Instrumentation,’’ for 

use of the Best Estimate Analyzer for 

Core Operations—Nuclear (BEACON) 

Power Distribution Monitoring System 

(PDMS) described in WCAP–12472–P– 

A, ‘‘BEACON Core Monitoring and 

Operations Support System,’’ to perform 

power distribution surveillances. 

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 PDMS performs continuous core 

power distribution monitoring with data 

input from existing plant instrumentation. 

This system utilizes an NRC [U.S. Nuclear 

Regulatory Commission] approved 

Westinghouse proprietary computer code, 

i.e., Best Estimate Analyzer for Core 

Operations—Nuclear (BEACON), to provide 

data reduction for incore flux maps, core 

parameter analysis, load follow operation 

simulation, and core prediction. The PDMS 

does not provide any protection or control 

system function. Fission product barriers are 

not impacted by these proposed changes. The 

proposed changes occurring with PDMS will 

not result in any additional challenges to 

plant equipment that could increase the 

probability of any previously evaluated 

accident. The changes associated with the 

PDMS do not affect plant systems such that 

their function in the control of radiological 

consequences is adversely affected. These 

proposed changes will therefore not affect the 

mitigation of the radiological consequences 

of any accident described in the Updated 

Safety Analysis Report (USAR). 

Use of the PDMS supports maintaining the 

core power distribution within required 

limits. Further continuous on-line 

monitoring through the use of PDMS 

provides significantly more information 

about the power distributions present in the 

core than is currently available. This results 

in more time (i.e., earlier determination of an 

adverse condition developing) for operator 

action prior to having an adverse condition 

develop that could lead to an accident 

condition or to unfavorable initial conditions 

for an accident. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequence of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Other than use of the PDMS to monitor 

core power distribution, implementation of 

the PDMS and associated Technical 

Specification changes has no impact on plant 

operations or safety, nor does it contribute in 

any way to the probability or consequences 

of an accident. No safety-related equipment, 

safety function, or plant operation will be 

altered as a result of this proposed change. 

The possibility for a new or different type of 

accident from any accident previously 

evaluated is not created since the changes 

associated with implementation of the PDMS 

do not result in a change to the design basis 

of any plant component or system. The 

evaluation of the effects of using the PDMS 

to monitor core power distribution 

parameters shows that all design standards 

and applicable safety criteria limits are met. 

The proposed changes do not result in any 

event previously deemed incredible being 

made credible. Implementation of the PDMS 

will not result in any additional adverse 

condition and will not result in any increase 

in the challenges to safety systems. The 

cycle-specific variables required by the 

PDMS are calculated using NRC-approved 

methods. The Technical Specifications will 

continue to require operation within the 

required core operating limits, and 

appropriate actions will continue to be taken 

when or if limits are exceeded. 

The proposed change, therefore, does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

No margin of safety is adversely affected by 

the implementation of the PDMS. The 

margins of safety provided by current 

Technical Specification requirements and 

limits remain unchanged, as the Technical 

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Federal Register / Vol. 75, No. 16 / Tuesday, January 26, 2010 / Notices 4121 

Specifications will continue to require 

operation within the core limits that are 

based on NRC-approved reload design 

methodologies. Appropriate measures exist 

to control the values of these cycle-specific 

limits, and appropriate actions will continue 

to be specified and taken for when limits are 

violated. Such actions remain unchanged. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jay Silberg, Esq., 

Pillsbury Winthrop Shaw Pittman LLP, 

2300 N Street, NW., Washington, DC 

20037. 

NRC Branch Chief: Michael T. 

Markley. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: 

November 20, 2009. 

Description of amendment request: 

The proposed changes will revise 

Technical Specification (TS) 3.8.1, ‘‘AC 

[Alternating Current] Sources— 

Operating,’’ by adding a Note to the 

Required Actions B.3.1 and B.3.2 to 

indicate that the TS 3.8.1 Required 

Actions B.3.1 and B.3.2 are satisfied if 

the diesel generator (DG) became 

inoperable due to an inoperable support 

system, an independently testable 

component, or preplanned preventive 

maintenance or testing. The amendment 

also proposes to revise the Completion 

Times for Required Actions B.3.1 and 

B.3.2 to specify a Completion Time 

based on the discovery of an issue or 

failure of the DG. 

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. 

WCNOC [Wolf Creek Nuclear Operating 

Corporation] is proposing to add a Note to 

Required Actions B.3.1 and B.3.2 to indicate 

that the TS 3.8.1 Required Actions of B.3 are 

satisfied if the DG became inoperable due to 

an inoperable support system, an 

independently testable component or 

preplanned preventative maintenance or 

testing. The proposed change to the TS does 

not involve a change in the operational limits 

or physical design of the emergency power 

system. Diesel generator (DG) OPERABILITY 

and reliability will continue to be assured 

while minimizing the potential number of 

required DG starts. The DGs are not an 

initiator of any accident previously 

evaluated. As a result, the probability of any 

accident previously evaluated is not 

significantly increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

No new or different accidents result for 

implementing the proposed change. 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 

operations. The change does not alter 

assumptions made in the safety analysis for 

DG performance. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change does not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. The safety analysis 

acceptance criteria are not impacted by this 

change. The proposed change will not result 

in operation in a configuration outside the 

design basis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jay Silberg, Esq., 

Pillsbury Winthrop Shaw Pittman LLP, 

2300 N Street, NW., Washington, DC 

20037. 

NRC Branch Chief: Michael T. 

Markley. 

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 

System (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. 

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of application for amendment: 

August 26, 2009. 

Brief description of amendment: The 

proposed amendment would revise the 

Technical Specification (TS) Section 6.5 

that governs administrative controls of 

High Radiation Areas (HRA) to 

incorporate the HRA administrative 

controls contained within the Standard 

Technical Specifications, NUREG–1433, 

Revision 3. 

Date of issuance: January 4, 2010. 

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Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 241. 

Facility Operating License No. DPR– 

28: Amendment revised the License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: October 20, 2009 (74 FR 

53778). 

The Commission’s related evaluation 

of this amendment is contained in a 

Safety Evaluation dated January 4, 2010. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of application for amendments: 

March 22, 2009. 

Brief description of amendments: The 

amendments revise the Technical 

Specification (TS) definition of the fully 

withdrawn position of the Rod Cluster 

Control Assemblies (RCCAs) to 

minimize localized RCCA wear. 

Previously, the fully withdrawn 

position for the RCCAs was defined in 

the TSs as being within the interval of 

222 to 228 steps withdrawn (i.e., steps 

above rod bottom). The approved 

change allows the fully withdrawn 

position to be defined as being within 

the interval of 222 to 230 steps 

withdrawn. 

Date of issuance: January 12, 2010. 

Effective date: As of the date of 

issuance. The Salem Unit No. 1 

amendment shall be implemented prior 

to entering Mode 2 following refueling 

outage 1R20 (currently scheduled for 

spring 2010). The Salem Unit No. 2 

amendment shall be implemented prior 

to entering Mode 2 following refueling 

outage 2R18 (currently scheduled for 

spring 2011). 

Amendment Nos.: 292 and 276. 

Facility Operating License Nos. DPR– 

70 and DPR–75: The amendments 

revised the TSs and the License. 

Date of initial notice in Federal 

Register: June 2, 2009 (74 FR 26435). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated January 12, 

2010. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 13th day 

of January 2010. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2010–1315 Filed 1–25–10; 8:45 am] 

BILLING CODE 7590–01–P 

SMALL BUSINESS ADMINISTRATION 

Data Collection Available for Public 

Comments and Recommendations 

ACTION: Notice and request for 

comments. 

SUMMARY: In accordance with the 

Paperwork Reduction Act of 1995, this 

notice announces the Small Business 

Administration’s intentions to request 

approval on a new and/or currently 

approved information collection. 

DATES: Submit comments on or before 

March 29, 2010. 

ADDRESSES: Send all comments 

regarding whether this information 

collection is necessary for the proper 

performance of the function of the 

agency, whether the burden estimates 

are accurate, and if there are ways to 

minimize the estimated burden and 

enhance the quality of the collection, to 

Sheila Thomas, Office of Business 

Development, Small Business 

Administration, 409 3rd Street, 8th 

Floor, Washington, DC 20416. 

FOR FURTHER INFORMATION CONTACT: 

Sheila Thomas, mail to: Office of 

Business Development, 202–205–5852 

sheila.thomas@sba.gov Curtis B. Rich, 

Management Analyst, 202–205–7030 

curtis.rich@sba.gov. 

SUPPLEMENTARY INFORMATION: This Form 

will be an Addendum to the 8(a) Annual 

Update Form (SBA Form 1450). The 

Section 8(a) Business Development (BD) 

Program was designed by Congress to 

provide socially and economically 

disadvantaged businesses with 

management and technical assistance to 

enhance their ability to compete in the 

American marketplace. The 8(a) 

Program utilizes various forms of 

assistance (e.g. procurement, financial, 

and management and technical 

assistance through 7(j) designated 

funds) to foster the business growth and 

development of 8(a) Program 

participants. 

In an effort to refocus the 8(a) 

Business Development Program to 

emphasize ‘‘business development’’ the 

SBA developed the 8(a) Business 

Development Assessment Tool (BDAT) 

that will be completed by the 8(a) 

Participant as part of the Annual Review 

Update process. The BDAT is an 

electronic questionnaire (which consists 

of topics ranging from general business 

questions to legal land insurance, 

business planning, financing, marketing 

and business operations) that allows the 

8(a) firm to answer a series of questions 

on a number of management and 

business skills. The 8(a) firm is then 

asked to rate their need for management 

and technical assistance in the specific 

skill area and a customized plan that 

addresses the firms’ stated needs is 

created. 

Title: ‘‘8(a) Annual Update 

Addendum.’’ 

Description of Respondents: 

Annually. 

Form Numbers: N/A. 

Annual Responses: 7,644. 

Annual Burden: 15,288. 

Jacqueline White, 

Chief, Administrative Information Branch. 

[FR Doc. 2010–1445 Filed 1–25–10; 8:45 am] 

BILLING CODE 8025–01–P 

SMALL BUSINESS ADMINISTRATION 

National Small Business Development 

Center Advisory Board 

AGENCY: U.S. Small Business 

Administration (SBA). 

ACTION: Notice of open Federal Advisory 

Committee meetings. 

SUMMARY: The SBA is issuing this notice 

to announce the location, date, time and 

agenda for the first quarter meetings of 

the National Small Business 

Development Center (SBDC) Advisory 

Board. 

DATES: The meetings for the fourth 

quarter will be held on the following 

dates: Tuesday, January 19, 2010 at 1 

p.m. EST. Tuesday, February 16, 2010 at 

1 p.m. EST. Tuesday, March 16, 2010 at 

1 p.m. EST. 

ADDRESSES: These meetings will be held 

via conference call. 

SUPPLEMENTARY INFORMATION: Pursuant 

to section 10(a) of the Federal Advisory 

Committee Act (5 U.S.C. Appendix 2), 

SBA announces the meetings of the 

National SBDC Advisory Board. This 

Board provides advice and counsel to 

the SBA Administrator and Associate 

Administrator for Small Business 

Development Centers. 

The purpose of these meetings is to 

discuss following issues pertaining to 

the SBDC Advisory Board: 

—ASBDC Spring Meeting 

—White Paper Issues 

—SBA Update 

—Member Roundtable 

FOR FURTHER INFORMATION CONTACT: The 

meeting is open to the public however 

advance notice of attendance is 

requested. Anyone wishing to be a 

listening participant must contact 

Alanna Falcone by fax or e-mail. Her 

contact information is Alanna Falcone, 

Program Analyst, 409 Third Street, SW., 

Washington, DC 20416, Phone, 202– 

619–1612, Fax 202–481–0134, e-mail, 

alanna.falcone@sba.gov. 

WReier-Aviles on DSKGBLS3C1PROD with NOTICES