Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 9821 

Dated: February 5, 2011. 

Richard W. Sherman, 

Deputy General Counsel. 

[FR Doc. 2011–3781 Filed 2–18–11; 8:45 am] 

BILLING CODE P 

NATIONAL CREDIT UNION 

ADMINISTRATION 

Agency Information Collection 

Activities: Submission to OMB for 

Review; Comment Request 

AGENCY: National Credit Union 

Administration (NCUA). 

ACTION: Request for comment. 

SUMMARY: The NCUA intends to submit 

the following information collection to 

the Office of Management and Budget 

(OMB) for review and clearance under 

the Paperwork Reduction Act of 1995 

(Pub. L. 104–13, 44 U.S.C. Chapter 35). 

This information collection is published 

to obtain comments from the public. 

DATES: Comments will be accepted until 

April 25, 2011. 

ADDRESSES: Interested parties are 

invited to submit written comments to 

NCUA Clearance Officer listed below: 

Clearance Officer: Tracy Sumpter, 

National Credit Union Administration, 

1775 Duke Street, Alexandria, Virginia 

22314–3428. Fax No. 703–837–2861, Email: OCIOmail@ncua.gov. 

FOR FURTHER INFORMATION CONTACT: 

Requests for additional information or a 

copy of the information collection 

request should be directed to Tracy 

Sumpter at the National Credit Union 

Administration, 1775 Duke Street, 

Alexandria, VA 22314–3428, or at (703) 

518–6440. 

SUPPLEMENTARY INFORMATION: Proposal 

for the following collection of 

information: 

OMB Number: 3133–0121. 

Form Number: 4063 and 4063a. 

Type of Review: Reinstatement, 

without change, of a previously 

approved collection. 

Title: Notice of Change of Official or 

Senior Executive Officer and Individual 

Application for Approval of Official or 

Senior Executive Officer. 

Description: In order to comply with 

statutory requirements, the agency must 

obtain sufficient information from new 

officials or senior executives officers of 

troubled or newly chartered credit 

unions to determine their fitness for the 

position. These forms standardize the 

information gathered to evaluate the 

individual’s fitness for the position. The 

format is similar to the one used by the 

FFIEC agencies and the FRB. 12 CFR 

701.14 and 741.205. 

Estimated No. of Respondents/ 

Recordkeepers: 650. 

Estimated Burden Hours per 

Response: 2.0 hours. 

Frequency of Response: Reporting and 

on occasion. 

Estimated Total Annual Burden 

Hours: 1300. 

Estimated Total Annual Cost: $ 0. 

By the National Credit Union 

Administration Board on February 15, 2011. 

Mary Rupp, 

Secretary of the Board. 

[FR Doc. 2011–3792 Filed 2–18–11; 8:45 am] 

BILLING CODE 7535–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2011–0040] 

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 January 27, 

2011, to February 10, 2011. The last 

biweekly notice was published on 

February 8, 2011 (76 FR 6830). 

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, Rules, 

Announcements and Directives Branch 

(RADB), 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 RADB 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, Room O1– 

F21, 11555 Rockville Pike (first floor), 

Rockville, Maryland 20852. 

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 

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9822 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 

‘‘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, Room 

O1–F21, 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/doccollections/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 

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 

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Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 9823 

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 1–866–672–7640. 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, 

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:// 

ehd1.nrc.gov/EHD/, 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 the 

date of publication of this notice. Nontimely 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, Room O1–F21, 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. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–333, James A. FitzPatrick 

Nuclear Power Plant, Oswego County, 

New York 

Date of amendment request: January 

13, 2011. 

Description of amendment request: 

The proposed amendment would 

modify the Facility Operating License 

(FOL) by deleting references to specific 

Safety Evaluation Reports (SER), 

Technical Specification (TS) 

Amendments, and Exemptions from 

License Condition 2.C(3), Fire 

Protection, and replacing them with the 

words ‘‘as supplemented.’’ This is an 

administrative amendment to the FOL. 

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 FOL change is 

administrative and does not involve a plant 

or design function change. It has no effect on 

reactor operation or accident analyses, and 

thus, the proposed FOL change does not 

increase the probability or consequence 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 FOL change is 

administrative and does not involve a plant 

or design function change. Because the 

proposed amendment would not change the 

design, configuration, or method of operation 

of the plant, it would 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 FOL change is 

administrative and does not involve a plant 

or design function change. No design or 

safety margin is involved. Therefore, the 

proposed change does not involve a 

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. 

Entergy Nuclear Vermont Yankee (VY), 

LLC and Entergy Nuclear Operations, 

Inc., Docket No. 50–271, Vermont 

Yankee Nuclear Power Station, Vernon, 

Vermont 

Date of amendment request: 

December 21, 2010. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specifications (TS) Section 

3.6.A ‘‘Pressure and Temperature 

Limitation’’ to reflect the pressure and 

temperature (P–T) limits for the reactor 

coolant system through, approximately 

the end of the prospective 20-year 

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9824 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 

renewed license period, depending on 

the plant capacity factor. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration which is presented below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the period of 

applicability of the P–T limits. The technical 

bases for the new period of applicability have 

been previously reviewed and approved by 

the NRC as discussed in the submittal. 

Because the applicable regulatory 

requirements continue to be met, the change 

does not significantly increase the probability 

of any accident previously evaluated. The 

proposed change provides the same 

assurance of RPV integrity as previously 

provided. 

The change will require that the reactor 

pressure vessel and interfacing coolant 

system continue to be operated within their 

design, operational or testing limits. Also, the 

change will not alter any assumptions 

previously made in evaluating the 

radiological consequences of accidents. 

Therefore, the proposed amendment does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

modification of the design of plant structures, 

systems, or components. The change will not 

impact the manner in which the plant is 

operated and will not degrade the reliability 

of structures, systems, or components 

important to safety as equipment protection 

features will not be deleted or modified, 

equipment redundancy or independence will 

not be reduced, supporting system 

performance will not be affected and no 

severe testing of equipment will be imposed. 

No new failure modes or mechanisms will be 

introduced as a result of this proposed 

change. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Appendix G to 10 CFR 50 describes the 

conditions that require pressure-temperature 

(P–T) limits and provides the general bases 

for these limits. Operating limits based on the 

criteria of Appendix G, as defined by 

applicable regulations, codes and standards, 

provide reasonable assurance that nonductile or rapidly propagating failure will not 

occur. The P–T limits are prescribed for all 

plant modes to avoid encountering pressure, 

temperature and temperature rate of change 

conditions that might cause undetected flaws 

to propagate and cause non-ductile failure of 

the reactor coolant pressure boundary. 

Calculation of P–T limits in accordance with 

the criteria of Appendix G to 10 CFR 50 and 

applicable regulatory requirements ensures 

that adequate margins of safety are 

maintained and there is no significant 

reduction in a margin of safety. 

The proposed change does not alter the 

manner in which safety limits, limiting safety 

system settings, or limiting conditions for 

operation are determined. There is no change 

or impact on any safety analysis assumption 

or in any other parameter affecting the course 

of an accident analysis supporting the basis 

of any Technical Specification. The proposed 

change does not involve any increase in 

calculated off-site dose consequences. 

Therefore, operation of VY in accordance 

with the proposed amendment will not 

involve a significant reduction in a margin to 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 400 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Nancy Salgado. 

FirstEnergy Nuclear Operating 

Company (FENOC), et al., Docket No. 

50–440, Perry Nuclear Power Plant, Unit 

1 (PNPP), Lake County, Ohio 

Date of amendment request: 

December 15, 2010. 

Description of amendment request: 

The proposed amendment would 

modify the requirements for testing 

control rod scram times following fuel 

movement within the reactor pressure 

vessel by incorporating Nuclear 

Regulatory Commission (NRC) approved 

Technical Specification Task Force 

(TSTF) change traveler TSTF–222–A, 

Revision 1. 

Basis for proposed no significant 

hazards consideration determination: 

As required by Title 10 of the Code of 

Federal Regulations (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 control rod drive system is not an 

initiator to any accident sequence analyzed 

in the PNPP Updated Final Safety Analysis 

Report (USAR), including Appendix 15C, 

‘‘Anticipated Transients Without Scram 

(ATWS).’’ The proposed TS changes improve 

existing surveillance requirements by 

eliminating unnecessary control rob scram 

time testing, while continuing to provide 

adequate assurance of control rod 

performance for those control rods in core 

cells in which fuel is moved or replaced, or 

control rod maintenance was performed. 

Historically, testing results indicate the 

control rod drive system is highly reliable. 

Since the fall 1996 implementation of 

Improved Technical Specifications, during 

6036 control rod tests covering all 177 

control rods, only 7 control rod tests (0.12 

percent) yielded results slower than the 

required insertion time limit, and no control 

rods were inoperable as a result of scream 

time testing. All seven slow insertion time 

test results have been attributed to control 

rod scream solenoid pilot valves (SSPVs). 

These seven slow tests occurred prior to May 

1999, and prior to a control rod SSPV 

upgrade program during which all 177 

SSPV’s were replaced. 

As such, this type of change does not affect 

initiators of analyzed events and does not 

affect the mitigation of any accidents or 

transients. 

Therefore, the proposed TS changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed TS changes do not involve 

a physical alteration of the plant. No new 

equipment is being introduced, and installed 

equipment is not being operated in a new or 

different manner. There are no setpoints 

affected by the changes at which protective 

or mitigative actions are initiated. The 

changes will not alter the manner in which 

equipment operation is initiated, nor will the 

functional demands on credited equipment 

be changed. No alterations in the procedures 

that ensure the plant remains within 

analyzed limits are being proposed, and no 

changes are being made to the procedures 

relied upon to respond to an off-normal event 

as described in the USAR. This change does 

not alter assumptions made in the safety 

analysis and licensing basis. As such, no new 

failures modes are being introduced. 

Accordingly, the proposed changes do not 

create any new credible failure mechanisms, 

malfunction, or accident initiators not 

previously considered in PNPP design and 

licensing basis. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Margin of safety is related to the ability of 

the fission product barriers to perform their 

design functions during and following 

accident conditions. These barriers include 

the fuel cladding, the reactor coolant system, 

and the containment. This request does not 

involve a change to the fuel cladding, the 

reactor coolant system, or the containment. 

The proposed TS changes associated with 

TSTF–222–1 modify current frequency 

requirements for scram time testing control 

rods following refueling outages and for 

control rod requiring testing due to work 

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Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 9825 

activities. Scram times for control rods not 

affected by fuel movement or control rod 

maintenance remain unaffected. 

The proposed TS changes have no affect on 

any safety analysis assumptions or methods 

of performing safety analyses. The changes 

do not adversely affect system design or 

operational requirements, and the equipment 

continues to be tested in a manner and at a 

frequency necessary to provide confidence 

that the equipment can perform its intended 

safety functions. 

Therefore, the proposed TS changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David W. 

Jenkins, Attorney, FirstEnergy 

Corporation, Mail Stop. A–GO–15, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: Robert D. Carlson. 

FirstEnergy Nuclear Operating 

Company (FENOC), et al., Docket No. 

50–440, Perry Nuclear Power Plant, Unit 

1 (PNPP), Lake County, Ohio 

Date of amendment request: 

December 15, 2010 

Description of amendment request: 

The proposed amendment would revise 

the required testing frequency of 

Surveillance Requirement (SR) 3.1.4.2 

from ‘‘120 days cumulative operation in 

MODE 1’’ to ‘‘200 days cumulative 

operation in MODE 1’’ by incorporating 

Nuclear Regulatory Commission (NRC) 

approved Technical Specification Task 

Force (TSTF) change traveler TSTF–460, 

Revision 0. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration which is presented below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change extends the 

frequency for testing control rod scram time 

testing from every 120 days of cumulative 

Mode 1 operation to 200 days of cumulative 

Mode 1 operation. The frequency of 

surveillance testing is not an initiator of any 

accident previously evaluated. The frequency 

of surveillance testing does not affect the 

ability to mitigate any accident previously 

evaluated, as the tested component is still 

required to be operable. 

Therefore, the proposed TS changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change extends the 

frequency for testing control rod scram time 

testing from every 120 days of cumulative 

Mode 1 operation to 200 days of cumulative 

Mode 1 operation. The proposed change does 

not result in any new or different modes of 

plant operation. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change extends the 

frequency for testing control rod scram time 

testing from every 120 days of cumulative 

Mode 1 operation to 200 days of cumulative 

Mode 1 operation. The proposed change 

continues to test the control rod scram time 

to ensure the assumptions in the safety 

analysis are protected. 

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: David W. 

Jenkins, Attorney, FirstEnergy 

Corporation, Mail Stop A–GO–15, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: Robert D. Carlson. 

FPL Energy Duane Arnold, LLC, Docket 

No. 50–331, Duane Arnold Energy 

Center, Linn County, Iowa 

Date of amendment request: October 

15, 2010. 

Description of amendment request: 

The proposed amendment would revise 

Operating License No. DPR–49 by 

modifying the License to delete the 

parent guarantee License Condition. 

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 is an 

administrative change deleting the parent 

guarantee License Condition, as well as other 

minor editorial changes in format. Deletion of 

this License Condition does not involve any 

modifications to the safety-related structures, 

systems or components (SSCs). Deletion of 

this License Condition will not alter 

previously evaluated Final Safety Analysis 

Report (FSAR) design basis accident analysis 

assumptions, add any accident initiators, or 

affect the function of the plant safety-related 

SSCs as to how they are operated, 

maintained, modified, tested, or inspected. 

Therefore, the proposed amendment does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment only deletes the 

parent guarantee License Condition and 

makes other minor editorial changes. 

Deletion of this License Condition does not 

result in the need for any new or different 

FSAR design basis accident analysis. It does 

not introduce new equipment that could 

create a new or different kind of accident, 

and no new equipment failure modes are 

created. As a result, no new accident 

scenarios, failure mechanisms, or limiting 

single failures are introduced as a result of 

this proposed amendment. Therefore, the 

proposed amendment does not create a 

possibility for an accident of a new or 

different type than those previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The margin of safety is associated with the 

confidence in the ability of the fission 

product barriers (i.e., fuel cladding, reactor 

coolant pressure boundary, and containment 

structure) to limit the level of radiation to the 

public. The proposed amendment would not 

alter the way any safety-related SSC 

functions and would not alter the way the 

plant is operated. The amendment only 

involves deletion of the parent guarantee 

License Condition and minor editorial 

changes. The proposed amendment would 

not introduce any new uncertainties or 

change any existing uncertainties associated 

with any safety limit. The proposed 

amendment would have no impact on the 

structural integrity of the fuel cladding, 

reactor coolant pressure boundary, or 

containment structure. Based on the above 

considerations, the proposed amendment 

would not degrade the confidence in the 

ability of the fission product barriers to limit 

the level of radiation to the public. Therefore, 

the proposed change does not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ms. Marjan 

Mashhadi, Florida Power & Light 

Company, 801 Pennsylvania Avenue, 

NW., Suite 220, Washington, DC 20004. 

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9826 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 

NRC Branch Chief: Robert J. 

Pascarelli. 

Indiana Michigan Power Company (the 

licensee), Docket No. 50–315, Donald C. 

Cook Nuclear Plant, Unit 1 (DCCNP–1), 

Berrien County, Michigan 

Date of amendment request: 

December 16, 2010. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 4.2.1, 

adding Optimized ZIRLO TM fuel rods to 

the fuel matrix in addition to Zircaloy 

or ZIRLO fuel rods that are currently in 

use. The proposed amendment would 

also add a Westinghouse topical report 

regarding Optimized ZIRLO TM as 

reference 8 in TS 5.6.5.b, which lists the 

analytical methods used to determine 

the core operating limits. 

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 would allow the use 

of Optimized ZIRLO TM clad nuclear fuel in 

the reactors. The NRC approved topical 

report WCAP–12610–P–A and CENPD–404– 

P–A, Addendum 1–A ‘‘Optimized ZIRLO TM,’’ 

prepared by Westinghouse Electric Company 

LLC (Westinghouse), addresses Optimized 

ZIRLO TM and demonstrates that Optimized 

ZIRLO TM has essentially the same properties 

as currently licensed ZIRLO TM. The fuel 

cladding itself is not an accident initiator and 

does not affect accident probability. Use of 

Optimized ZIRLO TM fuel cladding has been 

shown to meet all 10 CFR 50.46 acceptance 

criteria and, therefore, will not 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 proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Use of Optimized ZIRLO TM clad fuel will 

not result in changes in the operation or 

configuration of the facility. Topical Report 

WCAP–12610–P–A and CENPD–404–P–A 

demonstrated that the material properties of 

Optimized ZIRLO TM are similar to those of 

standard ZIRLO TM. Therefore, Optimized 

ZIRLO TM fuel rod cladding will perform 

similarly to those fabricated from standard 

ZIRLO TM, thus precluding the possibility of 

the fuel becoming an accident initiator and 

causing a new or different type of accident. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change will not involve a 

significant reduction in the margin of safety 

because it has been demonstrated that the 

material properties of the Optimized 

ZIRLO TM are not significantly different from 

those of standard ZIRLO TM. Optimized 

ZIRLO TM is expected to perform similarly to 

standard ZIRLO TM for all normal operating 

and accident scenarios, including both loss of 

coolant accident (LOCA) and non-LOCA 

scenarios. For LOCA scenarios, where the 

slight difference in Optimized ZIRLO TM

material properties relative to standard 

ZIRLO TM could have some impact on the 

overall accident scenario, plant-specific 

LOCA analyses using Optimized ZIRLO TM

properties will be performed prior to the use 

of fuel assemblies with fuel rods containing 

Optimized ZIRLO TM. These LOCA analyses 

will demonstrate that the acceptance criteria 

of 10 CFR 50.46 will be satisfied when 

Optimized ZIRLO TM fuel rod cladding is 

implemented. 

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: James M. Petro, 

Jr., Senior Nuclear Counsel, Indiana 

Michigan Power Company, One Cook 

Place, Bridgman, MI 49106. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Nine Mile Point Nuclear Station, LLC, 

(NMPNS) Docket No. 50–220, Nine Mile 

Point Nuclear Station Unit 1 (NMP1), 

Oswego County, New York 

Date of amendment request: 

September 29, 2010. 

Description of amendment request: 

The proposed amendment would revise 

the NMP1 Technical Specifications 

(TSs) Section 3/4.1.5, ‘‘SolenoidActuated Pressure Relief Valves 

(Automatic Depressurization System),’’ 

and 3/4.2.9, ‘‘Pressure Relief Systems— 

Solenoid-Actuated Pressure Relief 

Valves (Overpressurization),’’ to provide 

for an alternative means of testing the 

main steam electromatic relief valves 

(ERVs). Specifically, the proposed 

amendment would revise TS 

Surveillance Requirements (SRs) 4.1.5.a 

and 4.2.9.b to verify each ERV actuator 

strokes when manually actuated at least 

once each operating cycle. The 

functional testing requirements for the 

ERVs would be described in the 

Inservice Testing (IST) Program and 

controlled pursuant to TS 

Administrative Controls Section 6.5.4, 

‘‘Inservice Testing Program.’’ The 

proposed change would allow 

demonstration of the capability of the 

valves to perform their safety function 

without requiring the ERVs to be cycled 

with reactor steam pressure while 

installed in the plant. 

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 revises the TS 

Surveillance Requirements (SRs) to provide 

for an alternative means of testing the main 

steam ERVs. The ERVs perform automatic 

depressurization system (ADS) and 

overpressurization relief mode safety 

functions to mitigate the consequences of a 

small break loss of coolant accident 

(SBLOCA) and other accidents and 

transients. The ERVs are not considered an 

initiator for any accident previously 

evaluated except for the stuck-open ERV 

event, which is evaluated in Section XV– 

B.3.11 of the NMP1 Updated Final Safety 

Analysis Report (UFSAR). The proposed 

amendment would allow demonstration of 

the capability of the valves to perform their 

safety function through a series of tests, 

inspections, and maintenance activities 

without requiring the ERVs to be cycled with 

reactor steam pressure while installed in the 

plant, thereby eliminating the possibility of 

a stuck-open ERV event due to testing. Thus, 

the proposed amendment does not increase 

the probability of a stuck-open ERV event. 

The testing methodology, comprehensive 

inspections and preventive maintenance, and 

associated programmatic controls will 

provide an equivalent level of assurance that 

the ERVs are capable of performing their 

intended accident mitigation safety functions 

and, as such, will have no effect on the types 

or amounts of radiation released or the 

predicted offsite doses in the event of an 

accident. Accordingly, the proposed 

amendment does not alter the initial 

conditions, assumptions, or conclusions of 

any accident 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 amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment does not affect 

the assumed accident performance of the 

ERVs, or of any plant structure, system, or 

component previously evaluated. The 

proposed amendment does not involve the 

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Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 9827 

installation of new equipment, and installed 

equipment is not being operated in a new or 

different manner. The proposed amendment 

provides for an alternative means of testing 

the ERVs that does not involve opening the 

valves with reactor steam while installed in 

the plant. The alternative testing and 

associated programmatic controls will 

provide an equivalent level of assurance that 

the ERVs are capable of performing their 

accident mitigation safety functions. No 

setpoints are being changed that would alter 

the dynamic response of plant equipment. As 

such, the proposed amendment will not 

introduce any new failure modes. 

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 amendment provides for an 

alternative means of testing the ERVs, in that 

the testing requirements will be satisfied by 

a combination of required testing in 

accordance with the Inservice Testing 

Program (controlled in accordance with TS 

administrative controls) and the revised TS 

SRs. The proposed changes will provide a 

complete verification of the functional 

capability of the ERVs by performing a series 

of tests, inspections, and maintenance 

activities without opening the valves with 

reactor steam while installed in the plant. 

The alternative testing and associated 

programmatic controls will provide an 

equivalent level of assurance that the ERVs 

are capable of performing their intended 

accident mitigation safety functions. The 

proposed amendment does not affect the 

valve setpoints or adversely affect any other 

operational criteria assumed for accident 

mitigation. No changes are proposed that 

alter the setpoints at which protective actions 

are initiated, and there is no change to the 

operability requirements for equipment 

assumed to operate for accident mitigation. 

Moreover, it is expected that the alternative 

testing methodology will increase the margin 

of safety by reducing the potential for ERV 

leakage resulting from testing the ERVs with 

reactor steam pressure while installed in the 

plant. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Carey W. 

Fleming, Senior Counsel, Constellation 

Energy Nuclear Group, LLC, 100 

Constellation Way, Suite 200C, 

Baltimore, MD 21202. 

NRC Branch Chief: Nancy L. Salgado. 

Northern States Power Company— 

Minnesota, Docket Nos. 50–282 and 50– 

306, Prairie Island Nuclear Generating 

Plant, Units 1 and 2, Goodhue County, 

Minnesota 

Date of amendment request: February 

3, 2011. 

Description of amendment request: 

The proposed amendments would 

revise the Technical Specification (TS) 

3.8.1, ‘‘AC Sources—Operating’’, 

Surveillance Requirement 3.8.1.10 

footnote, which concerns battery 

charger modifications to be installed 

during or prior to the Unit 1 2011 

refueling outage. The proposed change 

will allow use of different battery 

charger modifications to those 

considered when the footnote was 

added to the TS. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This license amendment request proposes 

to revise the footnote to the emergency diesel 

generator Technical Specification 

surveillance requirement for loss of offsite 

power with safety injection actuation. The 

proposed footnote revision removes some 

specific requirements for battery charger 

modifications but will continue to assure that 

the applicable emergency diesel generator 

and its associated battery charger perform 

their required safety functions. 

The emergency diesel generators and their 

associated battery chargers are not accident 

initiators and therefore, these changes do not 

involve a significant increase [in] the 

probability of an accident. 

The proposed changes to the Technical 

Specification footnote will assure that the 

emergency diesel generator and the 

associated battery charger continue to 

perform their required safety function. Since 

the emergency diesel generator and the 

associated battery charger will provide 

required electrical power as assumed in the 

accident analyses, the results of the previous 

accident analyses are not changed and the 

changes proposed in this license amendment 

request do not involve a significant increase 

in the consequences of an accident. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

This license amendment request proposes 

to revise the footnote to the emergency diesel 

generator Technical Specification 

surveillance requirement for loss of offsite 

power with safety injection actuation. The 

proposed footnote revision removes some 

specific requirements for battery charger 

modifications but will continue to assure that 

the applicable emergency diesel generator 

and its associated battery charger perform 

their required safety functions. 

No new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of the proposed 

change. The proposed change does not 

challenge the performance or integrity of any 

safety-related system. The proposed change 

does involve modification of plant battery 

chargers, however, failures of battery 

chargers has been previously considered and 

bounded by assuming one safety related train 

of equipment fails. The modified battery 

chargers do not create new failure modes or 

mechanisms and no new accident precursors 

are generated. Surveillance testing 

requirements for the emergency diesel 

generator and battery charger will continue to 

demonstrate that the Limiting Conditions for 

Operation are met and the system 

components are functional. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

This license amendment request proposes 

to revise the footnote to the emergency diesel 

generator Technical Specification 

surveillance requirement for loss of offsite 

power with safety injection actuation. The 

proposed footnote revision removes some 

specific requirements for battery charger 

modifications but will continue to assure that 

the applicable emergency diesel generator 

and its associated battery charger perform 

their required safety functions. 

The proposed Technical Specification 

footnote change does not affect the 

availability, operability, or performance of 

safety-related systems and components: The 

affected emergency diesel generator and its 

associated battery will continue to perform 

their safety functions. The ability of operable 

structures, systems, and components to 

perform their designated safety function is 

unaffected by this proposed change. The 

proposed change does not involve a 

significant reduction in a margin of safety 

because the proposed footnote changes do 

not reduce the margin of safety that exists in 

the present Technical Specifications or 

Updated Safety Analysis Report. The 

operability requirements of the Technical 

Specifications are consistent with the initial 

condition assumptions of the safety analyses 

and the surveillance testing requirements 

will continue to demonstrate the operability 

of the emergency diesel generator. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

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9828 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment requests involve no 

significant hazards consideration. 

Attorney for licensee: Peter M. Glass, 

Assistant General Counsel, Xcel Energy 

Services, Inc., 414 Nicollet Mall, 

Minneapolis, MN 55401 

NRC Branch Chief: Robert J. Pascarell. 

PPL Susquehanna, LLC, Docket Nos. 50– 

387 and 50–388, Susquehanna Steam 

Electric Station, Units 1 and 2, Luzerne 

County, Pennsylvania 

Date of amendment request: 

November 10, 2010. 

Description of amendment request: 

The change to the PPL Susquehanna, 

LLC (PPL) Unit 1 and Unit 2 Technical 

Specification (TS) Surveillance 

Requirement (SR) 3.4.3.1 ‘‘Safety/Relief 

Valves (S/RVs)’’ proposes a new safety 

function lift setpoint lower tolerance for 

the S/RVs. The proposed change will 

revise the lower tolerances from ¥3% 

to ¥5%. This change would be limited 

to the lower tolerances and does not 

affect the upper tolerances. This change 

only applies to the lower as-found 

tolerance and not to the as-left 

tolerance, which will remain unchanged 

at ±1% of the safety lift setpoint. The asfound tolerances are used for 

determining past operability and to 

increase sample sizes for S/RV testing 

should the upper tolerance be exceeded. 

There will be no revision to the actual 

setpoints of the valves installed in the 

plant due to this change. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This change has no influence on the 

probability or consequences of any accident 

previously evaluated. The lower setpoint 

tolerance change does not affect the 

operation of the valves and it does not 

change the as-left setpoint tolerance. The 

change only affects the lower tolerance for 

opening the valve and does not change the 

upper tolerance, which is the limit that 

protects from overpressurization. 

The proposed action does not involve 

physical changes to the valves, nor does it 

change the safety function of the valves. The 

proposed TS revision involves no significant 

changes to the operation of any systems or 

components in normal or accident operating 

conditions and no changes to existing 

structures, systems, or components. 

The proposed action does not change any 

other behavior or operation of any S/RVs, 

and, therefore, has no significant impact on 

reactor operation. It also has no significant 

impact on response to any perturbation of 

reactor operation including transients and 

accidents previously analyzed in the Final 

Safety Analysis Report (FSAR). 

Therefore, the proposed amendment does 

not result in a significant increase in the 

probability or consequences of any 

previously evaluated accident. 

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 lower setpoint tolerance 

change only affects the criteria to determine 

when an as-found S/RV test is considered to 

be acceptable. This change does not affect the 

criteria for the upper setpoint tolerance. 

The proposed lower setpoint tolerance 

change does not adversely affect the 

operation of any safety-related components 

or equipment. Since the proposed action 

does not involve hardware changes, 

significant changes to the operation of any 

systems or components, nor change to 

existing structures, systems, or components, 

there is no possibility that a new or different 

kind of accident is created. 

The proposed change does not involve 

physical changes to the S/RVs, nor does it 

change the safety function of the S/RVs. The 

proposed change does not require any 

physical change or alteration of any existing 

plant equipment. No new or different 

equipment is being installed, and installed 

equipment is not being operated in a new or 

different manner. There is no alteration to the 

parameters within which the plant is 

normally operated. This change does not 

alter the manner in which equipment 

operation is initiated, nor will the functional 

demands on credited equipment be changed. 

No alterations in the procedures that ensure 

the plant remains within analyzed limits are 

being proposed, and no changes are being 

made to the procedures relied upon to 

respond to an off-normal event as described 

in the FSAR. As such, no new failure modes 

are being introduced. The change does not 

alter assumptions made in the safety analysis 

and licensing basis. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed lower setpoint tolerance 

change only affects the criteria to determine 

when an as-found S/RV test is considered to 

be acceptable. This change does not affect the 

criteria for the upper setpoint tolerance. The 

TS setpoints for the S/RVs are not changed. 

The as-left setpoint tolerances are not 

changed by this proposed change. 

The margin of safety is established through 

the design of the plant structures, systems, 

and components, the parameters within 

which the plant is operated, and the 

establishment of the setpoints for the 

actuation of equipment relied upon to 

respond to an event. The proposed change 

does not significantly impact the condition or 

performance of structures, systems, and 

components relied upon for accident 

mitigation. 

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: Bryan A. Snapp, 

Esquire, Assoc. General Counsel, PPL 

Services Corporation, 2 North Ninth St., 

GENTW3, Allentown, PA 18101–1179. 

NRC Branch Chief : Nancy L. Salgado. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–424 and 50–425, 

Vogtle Electric Generating Plant, Units 1 

and 2, Burke County, Georgia, and 

Southern Nuclear Operating Company, 

Inc., Georgia Power Company, 

Oglethorpe Power Corporation, 

Municipal Electric Authority of Georgia, 

City of Dalton, Georgia, Docket Nos. 50– 

321 and 50–366, Edwin I. Hatch Nuclear 

Plant, Unit 1 and 2, Appling County, 

Georgia 

Date of amendment request: 

December 16, 2010. 

Description of amendment request: 

The proposed amendments would 

revise Technical Specification (TS) 

Section 2.0 ‘‘Safety Limits.’’ Specifically, 

the removal of the requirement to report 

a Safety Limit Violation, that is 

redundant to existing regulations, Title 

10 of the Code of Federal Regulations 

(10 CFR), Part 50.36(c)(8) ‘‘Written 

Reports.’’ The proposed change is 

described in Technical Specification 

Task Force Traveler TSTF–5–A, 

Revision 1, ‘‘Delete Safety Limit 

Violation Notification Requirements,’’ 

(Agencywide Documents Access and 

Management System (ADAMS) 

Accession No. ML052010227), and was 

described in the Notice of Availability 

published in the Federal Register (FR) 

on November 4, 2005 (70 FR 67202). 

The proposed changes are consistent 

with the NRC-approved TSTF–5–A, 

Revision 1. 

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. 

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Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 9829 

The proposed change to remove the 

duplicative safety limit reporting, 

notification, and restart constraint 

requirements from the TS does not affect the 

plant or operation of the plant. The change 

simply removes duplicative information from 

the TS that is covered in the NRC regulations. 

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 previously evaluated? 

Response: No. 

The proposed change to remove the 

duplicative safety limit reporting, 

notification, and restart constraint 

requirements from the TS does not introduce 

any new accident scenarios, failure 

mechanisms, or limiting single failures. All 

systems, structures, and components 

previously required for the mitigation of a 

transient remain capable of fulfilling their 

intended design functions. The proposed 

change has no adverse effect on any safetyrelated system or component and does not 

challenge the performance or integrity of any 

safety related system. This change is 

considered an administrative action to 

remove duplicative reporting, notification, 

and restart constraint requirements. 

Therefore, this proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes are administrative 

and do not involve any reduction in a margin 

of safety. All systems, structures, and 

components previously required for the 

mitigation of a transient remain capable of 

fulfilling their intended design functions. 

The proposed change has no adverse effect 

on any safety-related system or component 

and does not [involve a significant reduction 

in a margin of safety.] 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Arthur H. 

Domby, Troutman Sanders, 

NationsBank Plaza, Suite 5200, 600 

Peachtree Street, NE., Atlanta, Georgia 

30308–2216. 

NRC Branch Chief: Gloria Kulesa. 

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, Room O1–F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland 

20852. 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. 

Dairyland Power Cooperative, Docket 

No. 50–409, La Crosse Boiling Water 

Reactor, Vernon County, Wisconsin 

Date of application for amendment: 

July 28, 2009, and supplemented August 

7, 2009, May 19, 2010, and August 12, 

2010. 

Brief description of amendment: The 

amendment revises the La Crosse 

Boiling Water Rector (LACBWR) 

Technical Specifications, in support of 

the dry cask storage project at LACBWR. 

Date of issuance: January 25, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 71. 

Facility Operating License No. DPR–7: 

This amendment revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: October 6, 2009 (74 FR 51326). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 25, 

2011. 

No significant hazards consideration 

comments received: No. 

Dominion Electric Kewaunee, Inc. 

Docket No. 50–305, Kewaunee Power 

Station (KPS), Kewaunee County, 

Wisconsin 

Date of application for amendment: 

August 24, 2009 (Agencywide 

Documents and Management System 

(ADAMS) Accession No. 

ML092440398), as supplemented by 

letters dated October 22, 2009 (ADAMS 

Accession No. ML093070092), April 13, 

2010 (ADAMS Accession Nos. 

ML101060517 and ML101040090), May 

12, 2010 (ADAMS Accession No. 

ML101380399), July 1, 2010 (ADAMS 

Accession No. ML101890404), July 16, 

2010 (ADAMS Accession No. 

ML102370370), August 18, 2010 

(ADAMS Accession No. ML102371064), 

September 7, 2010 (ADAMS Accession 

No. ML102730383), September 8, 2010 

(ADAMS Accession No. ML102580700), 

October 15, 2010 (ADAMS Accession 

No. ML102920037), and December 2, 

2010 (ADAMS Accession No. 

ML103400328). 

Brief description of amendment: This 

amendment converts the current 

technical specifications (CTSs) to the 

improved TSs (ITSs) and relocates 

certain requirements to other licenseecontrolled documents. The ITSs are 

based on NUREG–1431, Rev. 3.0, 

‘‘Standard Technical Specifications, 

Westinghouse Plants,’’ Revision 3.0; 

‘‘NRC Final Policy Statement on 

Technical Specification Improvements 

for Nuclear Power Reactors,’’ dated July 

22, 1993 (58 FR 39132); and 10 CFR 

50.36, ‘‘Technical Specifications.’’ 

Technical Specification Task Force 

changes were also incorporated. The 

purpose of the conversion is to provide 

clearer and more readily understandable 

requirements in the TSs for KPS to 

ensure safe operation. In addition, the 

amendment includes a number of issues 

that were considered beyond the scope 

of NUREG–1431. 

Date of issuance: February 2, 2011. 

Effective date: As of the date of 

issuance and shall be implemented on 

or before February 23, 2011. 

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9830 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 

Amendment No.: 207. 

Facility Operating License No. DPR– 

43: Amendment revised the Technical 

Specifications and License. 

Date of initial notice in Federal 

Register: December 15, 2009 (74 FR 

66384). The supplements provided, 

contained clarifying information and 

did not expand the scope of the 

application as originally noticed. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 2, 

2011. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Nuclear 

Plant, Van Buren County, Michigan 

Date of application for amendment: 

January 27, 2010. 

Brief description of amendment: 

The amendment revises Section 2.E. 

of the Palisades Nuclear Plant (PNP) 

Renewed Facility Operating License to 

remove the name of the former operator 

of the plant in the title of the PNP 

physical security plan and replace it 

with Entergy Nuclear. The change also 

removes the security plan revision 

number and the date the plan was 

submitted to the Nuclear Regulatory 

Commission. 

Date of issuance: January 25, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 241. 

Facility Operating License No. DPR– 

20: Amendment revised the Technical 

Specifications. 

Public comments requested as to 

proposed no significant hazards 

considerations (NSHC): The notice 

provided an opportunity to submit 

comments on the Commission’s 

proposed NSHC determination. No 

comments have been received. 

Date of initial notice in Federal 

Register: November 18, 2010 (75 FR 

70708), followed by the repeat biweekly 

notice in the Federal Register on 

January 25, 2011 (76 FR 4389). 

The Commission’s related evaluation 

of the amendment, state consultation, 

and final NSHC determination are 

contained in a Safety Evaluation dated 

January 25, 2011. 

Attorney for licensee: Mr. William 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Ave., White Plains, NY 10601. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of application for amendment: 

January 24, 2010, as supplemented by 

letters dated September 7 and November 

4, 2010. 

Brief description of amendment: This 

amendment request would revise the 

Technical Specifications (TSs) Section 

1.0, Definitions, TS Section 3.6, Primary 

System Boundary Specifications 3.6.A, 

and TS Programs and Manuals Section 

5.5, to include reference to the Pressure 

and Temperature Limits Report (PTLR). 

The proposed PTLR would include 

revised 43 effective full-power years 

pressure-temperature curves, neutron 

fluence, and adjusted reference 

temperature values. 

Date of issuance: January 26, 2011. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 234. 

Facility Operating License No. DPR– 

35: The amendment revised the License 

and Technical Specifications. 

Date of initial notice in Federal 

Register: April 6, 2010 (75 FR 17443). 

The supplemental letters dated 

September 7 and November 4, 2010, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of this amendment is contained in a 

Safety Evaluation dated January 26, 

2011. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of amendment request: April 13, 

2010 as supplemented by letter dated. 

February 2, 2011. 

Description of amendment request: 

The amendment would revise Technical 

Specification (TS) to update the Table of 

Contents and the Applicability and 

Objective portions of TS 4.12 as a result 

of changes made by License 

Amendment Nos. 230 and 239 and to 

revise wording in TS 3.7.A.8. The 

changes are considered administrative 

in nature and do not materially change 

any technical requirement. 

Date of Issuance: February 9, 2011. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 245. 

Facility Operating License No. DPR– 

28: Amendment revised the License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: June 29, 2010 (75 FR 37474). 

The supplement letter dated February 2, 

2011, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of this amendment is contained in a 

Safety Evaluation dated February 9, 

2011. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50– 

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: July 20, 

2010. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) 3.7.1.2, ‘‘Emergency 

Feedwater System,’’ Limiting Condition 

for Operation (LCO) 3/4.7.1.2, 

‘‘Emergency Feedwater,’’ to clarify the 

acceptability of transitioning from Mode 

4, Hot Shutdown, to Mode 3, Hot 

Standby, with the turbine-driven 

emergency feedwater (EFW) pump 

inoperable but available. The 

amendment granted an exception to TS 

LCO 3.0.4 and Surveillance 

Requirement 4.0.4 allowing entry into 

operational Mode 3 with TS LCO 

equipment, the turbine-driven EFW 

pump, associated with a shutdown 

action inoperable. 

Date of issuance: January 31, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 60 

days from the date of issuance. 

Amendment No.: 230. 

Facility Operating License No. NPF– 

38: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: September 21, 2010 (75 FR 

57523). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 31, 

2011. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois 

Date of amendment request: February 

4, 2010 as supplemented by letters 

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Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 9831 

dated September 15, 2010, October 6, 

2010, and December 13, 2010. 

Description of amendment request: 

The proposed amendments would 

revise Technical Specification (TS) 

3.3.6.1, ‘‘Primary Containment Isolation 

Instrumentation,’’ ‘‘Table 3.3.6.1–1, 

‘‘Primary Containment Isolation 

Instrumentation,’’ Function 6.a 

‘‘Shutdown Cooling System Isolation, 

Recirculation Line Water Temperature— 

High,’’ to enable implementation with 

reactor pressure-based isolation 

instrumentation, for the Dresden 

Nuclear Power Station, Units 2 and 3. 

Date of issuance: February 7, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 236/229. 

Facility Operating License Nos. DPR– 

19 and DPR–25: The amendment 

revised the Technical Specifications and 

License. 

Date of initial notice in Federal 

Register: April 20, 2010 (75 FR 20635). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated February 7, 

2011. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

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

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Date of application for amendments: 

Dated October 5, 2009 as supplemented 

by letters dated June 10, November 23, 

December 14, and December 22, 2010, 

and January 11, 24, and 28, 2011. 

Brief description of amendments: The 

proposed amendment would revise 

Technical Specification (TS) 4.3.1, 

‘‘Criticality,’’ to address a nonconservative TS. The proposed change 

addresses the Boraflex degradation issue 

in the LSCS Unit 2 spent fuel storage 

racks by revising TS Section 4.3.1 to 

allow the use of NETCO–SNAP–IN®

rack inserts in LSCS Unit 2 spent fuel 

storage rack cells as a replacement for 

the neutron absorbing properties of the 

existing Boraflex panels. 

Date of issuance: January 28, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days after the end of Unit 2 

refueling outage 13. 

Amendment Nos.: 199 and 186. 

Facility Operating License Nos. NPF– 

11 and NPF–18: The amendments 

revised the Technical Specifications and 

License. 

Date of initial notice in Federal 

Register: January 5, 2010 (75 FR 463). 

The June 10, November 23, December 

14, and December 22, 2010, and January 

11, 24, and 28, 2011, submittals 

contained clarifying information and 

did not change the NRC staff’s initial 

proposed finding of no significant 

hazards consideration. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated January 28, 

2011. 

No significant hazards consideration 

comments received: No. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–346, 

Davis-Besse Nuclear Power Station, Unit 

No. 1, Ottawa County, Ohio 

Date of amendment request: April 15, 

2009, as supplemented by letters dated 

December 18, 2009, October 8, 2010 and 

January 10, 2011. 

Brief description of amendment 

request: The amendment request and 

proposed exemption request were to 

incorporate a new methodology for the 

development of Reactor Coolant System 

(RCS) pressure-temperature limits into 

Technical Specification (TS) 5.6.4, 

‘‘Reactor Coolant System (RCS) Pressure 

and Temperature Limits Report (PTLR).’’ 

The amendment also requested a 

revision to the period of validity of the 

analysis for the low temperature 

overpressure protection (LTOP) system 

contained in Operating License 

Condition 2.C(3)(d). An associated 

revision to the Technical Specification 

Basis 3.4.12 ‘‘Low Temperature 

Overpressure Protection (LTOP)’’ 

supports the change to the operating 

license condition. 

Date of issuance: January 28, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment No.: 282. 

Facility Operating License No. NPF–3: 

The amendment revised the TS and 

license. 

Date of initial notice in Federal 

Register: June 16, 2009 (72 FR 28577). 

The supplemental letters contained 

clarifying information, did not change 

the initial no significant hazards 

consideration determination, and did 

not expand the scope of the original 

Federal Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 28, 

2011. 

No significant hazards consideration 

comments received: No. 

Nine Mile Point Nuclear Station, LLC, 

Docket No. 50–410, Nine Mile Point 

Nuclear Station, Unit 2 (NMP2), Oswego 

County, New York 

Date of application for amendment: 

December 9, 2009. 

Brief description of amendment: The 

amendment changes the NMP2 

Technical Specification (TS) 3.8.4, ‘‘DC 

Sources—Operating,’’ to remove the 

Mode restrictions for performance of TS 

Surveillance Requirements (SRs) 3.8.4.7 

and 3.8.4.8 for the Division 3 direct 

current (DC) electrical power subsystem 

battery. The Division 3 DC electrical 

power subsystem feeds emergency DC 

loads associated with the high-pressure 

core spray (HPCS) system. These 

surveillances verify that the battery 

capacity is adequate for the battery to 

perform its required functions. The 

amendment removes these Mode 

restrictions for the Division 3 battery, 

thereby allowing performance of the SRs 

during Mode 1, 2, or 3 in conjunction 

with scheduled HPCS system outages. 

Date of issuance: January 31, 2011. 

Effective date: As of the date of 

issuance to be implemented within 90 

days. 

Amendment No.: 136. 

Renewed Facility Operating License 

No. NPF–069: The amendment revises 

the License and TSs. 

Date of initial notice in Federal 

Register: April 6, 2010 (75 FR 17444). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 31, 

2011. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–424 and 50–425, 

Vogtle Electric Generating Plant, Units 1 

and 2, Burke County, Georgia 

Date of application for amendments: 

February 2, 2010. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications (TSs) Table 3.3.1–1 

‘‘Reactor Trip System Instrumentation 

[RTS],’’ Function 3, ‘‘Power Range 

Neutron Flux High Positive Rate.’’ 

Specifically, the revision added 

surveillance requirement 3.3.1.15 to 

verify the RTS response time is within 

limits. 

Date of issuance: February 7, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment Nos.: 159 and 141. 

Facility Operating License Nos. NPF– 

68 and NPF–81: Amendments revised 

the licenses and the TSs. 

Date of initial notice in Federal 

Register: May 4, 2010 (75 FR 23817). 

The supplement dated October 29, 2010, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the NRC 

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9832 Federal Register / Vol. 76, No. 35 / Tuesday, February 22, 2011 / Notices 

staff’s original proposed no significant 

hazards consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated February 7, 

2011. 

No significant hazards consideration 

comments received: No. 

Virginia Electric and Power Company, 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Units 1 and 2, 

Louisa County, Virginia 

Date of application for amendment: 

March 30, 2010. 

Brief description of amendment: The 

amendments revised the North Anna 

Technical Specifications (TSs) by 

relocating specific surveillance 

frequencies to a licensee-controlled 

program with the implementation of 

Nuclear Energy Institute (NEI) 04–10, 

‘‘Risk-Informed Technical Specifications 

Initiative 5b, Risk-Informed Method for 

Control of Surveillance Frequencies.’’ 

Date of issuance: January 31, 2011. 

Effective date: As of the date of 

issuance and shall be implemented 

within 180 days from the date of 

issuance. 

Amendment Nos.: 262 and 243. 

Renewed Facility Operating License 

Nos. NPF–4 and NPF–7: Amendments 

changed the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: May 18, 2010 (75 FR 27833). 

The supplements dated August 30, 

2010, and January 18, 2011, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination. The 

Commission’s related evaluation of the 

amendments is contained in a Safety 

Evaluation dated January 31, 2011. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 10th day 

of February 2011. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2011–3721 Filed 2–18–11; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket No. 03034542; License No: 37– 

30412–01; EA–10–077; NRC–2011–0041] 

In the Matter of Superior Well Services, 

Ltd. Indiana, PA; Confirmatory Order 

Modifying License (Effective 

Immediately) 

I 

Superior Well Services, Ltd. (SWS or 

Licensee) is the holder of radioactive 

material License No. 37–30412–01 

issued by the U.S. Nuclear Regulatory 

Commission (NRC or agency) pursuant 

to 10 CFR Part 30. The license 

authorizes the possession, storage, and 

use of radioactive sources for oil and gas 

well logging at the Licensee’s facilities 

in Buckhannon, West Virginia, Sophia, 

West Virginia, and Gaylord, Michigan, 

and at temporary jobsites within the 

NRC’s jurisdiction, in accordance with 

conditions specified therein. 

II 

On October 21, 2010, the NRC issued 

a Notice of Violation (Notice) and 

Proposed Imposition of Civil Penalty 

(CP) in the amount of $34,000 for five 

violations that were categorized into 

two severity level (SL) III problems. The 

violations were identified during an 

NRC inspection as well as an 

investigation conducted by the NRC 

Office of Investigations (OI). (Reference: 

NRC Inspection Report No. 03034542/ 

2009001 and OI Investigation Report 

No. 1–2009–035). The violations were 

also discussed at a predecisional 

enforcement conference (PEC) on 

September 2, 2010. 

The first SL III problem described in 

the Notice related to an event that 

occurred on September 20, 2008, when 

two well-logging sealed sources fell off 

of a company truck during transport. 

One violation involved the failure to 

secure the packages containing the 

licensed material from shifting during 

transport. On September 20, 2008, when 

the truck transporting these sources 

reportedly hit a large pothole, the weld 

securing the source plate to the truck 

broke, and the sources fell off of the 

truck and remained unattended by the 

side of a public highway. The second 

violation involved the failure to control 

and maintain constant surveillance of 

the sources while they were on the 

highway (an unrestricted area). Since 

SWS did not recognize that the sources 

had fallen out of the truck until the 

truck reached its destination at the SWS 

facility in Buckhannon, WV, the sources 

remained unattended for approximately 

ninety minutes until SWS personnel 

located and retrieved the sources. The 

third violation involved the failure to 

immediately report this occurrence by 

telephone to the NRC Operations 

Center. The involved SWS employees, 

including the site Radiation Safety 

Officer (RSO) for the associated SWS 

facility, did not recognize the need to 

report this event to the NRC. As a result, 

SWS did not provide the required 

immediate telephone notification of this 

event to the NRC Operations Center 

until July 23, 2009, after an NRC 

inspector informed SWS of the 

reportability requirement while 

conducting a routine inspection. 

The second SL III problem described 

in the Notice involved SWS’s failure to 

conduct required radiological surveys of 

vehicles before transporting licensed 

material and the deliberate falsification 

of survey records for these vehicles. 

Specifically, former SWS employees 

informed the NRC inspector and 

investigator that on numerous 

occasions, they did not perform the 

surveys and they instead completed the 

survey forms by copying data from 

previously completed forms. The 

employees’ failure to perform the 

required radiological surveys of vehicles 

prevented SWS from assuring that the 

dose rates inside and outside the trucks 

did not exceed limits set by the NRC 

and the U.S. Department of 

Transportation. The employees who 

admitted to the NRC that they had 

falsified survey records indicated that 

they did so because they did not know 

how to use the survey instruments. 

III 

In response to the October 21, 2010, 

NRC letter, SWS requested the use of 

the NRC’s Alternative Dispute 

Resolution (ADR) process to resolve 

differences it had with the NRC 

regarding the Notice. ADR is a process 

in which a neutral professional 

mediator with no decision-making 

authority assists the parties in reaching 

an agreement to resolve any differences 

regarding the enforcement action. On 

January 4, 2011, the NRC and SWS met 

in an ADR mediation session, arranged 

through Cornell University’s Scheinman 

Institute on Conflict Resolution. During 

that ADR mediation session, an 

agreement in principle was reached. 

This Confirmatory Order is the result of 

that agreement, the elements of which 

consisted of the following: 

1. SWS did not take issue with the 

NRC conclusion set forth in the October 

21, 2010, letter and enclosed Notice that 

the subject violations regarding the 

temporary loss of two well-logging 

sources occurred as identified. Further, 

SWS did not take issue with the NRC 

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