[Federal Register Volume 76, Number 220 (Tuesday, November 15, 2011)]
[Notices]
[Pages 70768-70777]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-29435]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2011-0261]


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving No Significant Hazards Considerations

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)

[[Page 70769]]

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 October 20, 2011 to November 2, 2011. The 
last biweekly notice was published on November 1, 2011 (76 FR 67485).

ADDRESSES: Please include Docket ID NRC-2011-0261 in the subject line 
of your comments. For additional instructions on submitting comments 
and instructions on accessing documents related to this action, see 
``Submitting Comments and Accessing Information'' in the SUPPLEMENTARY 
INFORMATION section of this document. You may submit comments by any 
one of the following methods:
     Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for documents filed under Docket ID NRC-
2011-0261. Address questions about NRC dockets to Carol Gallagher, 
telephone: (301) 492-3668; email: [email protected].
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
     Fax comments to: RADB at (301) 492-3446.

SUPPLEMENTARY INFORMATION: 

Submitting Comments and Accessing Information

    Comments submitted in writing or in electronic form will be posted 
on the NRC Web site and on the Federal rulemaking Web site, http://www.regulations.gov. Because your comments will not be edited to remove 
any identifying or contact information, the NRC cautions you against 
including any information in your submission that you do not want to be 
publicly disclosed.
    The NRC requests that any party soliciting or aggregating comments 
received from other persons for submission to the NRC inform those 
persons that the NRC will not edit their comments to remove any 
identifying or contact information, and therefore, they should not 
include any information in their comments that they do not want 
publicly disclosed.
    You can access publicly available documents related to this 
document using the following methods:
     NRC's Public Document Room (PDR): The public may examine 
and have copied, for a fee, publicly available documents at the NRC's 
PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, 
Rockville, Maryland 20852.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): Publicly available documents created or received at the NRC 
are available online in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can gain entry into ADAMS, 
which provides text and image files of the NRC's public documents. If 
you do not have access to ADAMS or if there are problems in accessing 
the documents located in ADAMS, contact the NRC's PDR reference staff 
at 1 (800) 397-4209, (301) 415-4737, or by email to 
[email protected].
     Federal Rulemaking Web Site: Public comments and 
supporting materials related to this notice can be found at http://www.regulations.gov by searching on Docket ID NRC-2011-0261.

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) 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; (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.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ``Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 
Rockville Pike (first floor), Rockville, Maryland 20852. NRC 
regulations are accessible electronically from the NRC Library on the 
NRC Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If 
a request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the

[[Page 70770]]

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, then 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 E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at (301) 415-1677, to request (1) A digital identification 
(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 NRC-issued 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 the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the 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 the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in 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/e-submittals.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 the NRC guidance 
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.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 email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at http://www.nrc.gov/site-help/e-submittals.html, by email at 
[email protected], or by a toll-free 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

[[Page 70771]]

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 first-class 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 the 
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. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through ADAMS in the NRC Library at 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's PDR Reference staff at 1-(800) 397-
4209, (301) 415-4737, or by email to [email protected].

Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone 
Power Station, Unit 3, New London County, Connecticut

    Date of amendment request: July 5, 2011, as supplemented by letter 
dated September 12, 2011.
    Description of amendment request: The proposed amendment would 
modify the Millstone Power Station, Unit 3 (MPS3), Technical 
Specifications (TSs) by relocating specific surveillance frequencies to 
a licensee-controlled program, the Surveillance Frequency Control 
Program (SFCP). The proposed changes are based on the Nuclear 
Regulatory Commission (NRC)-approved Technical Specification Task Force 
(TSTF)-425, Revision 3, ``Relocate Surveillance Frequencies to Licensee 
Control--RITSTF [Risk-Informed TSTF] Initiative 5b'' (Agencywide 
Documents Access and Management System (ADAMS) Package Accession No. 
ML090850642). Plant-specific deviations from TSTF-425 are proposed to 
accommodate differences between the MPS3 TSs and the model TSs 
originally used to develop TSTF-425. The proposed plant-specific 
deviations involve fixed periodic frequency surveillances, and are 
therefore consistent with TSTF-425, and editorial deviations.
    The NRC staff issued a Notice of Availability for TSTF-425 in the 
Federal Register on July 6, 2009 (74 FR 31996). The notice included a 
model safety evaluation and a model no significant hazards 
consideration (NSHC) determination. In its application dated July 5, 
2011, as supplemented by letter dated September 12, 2011, Dominion 
Nuclear Connecticut, Inc. (DNC or the licensee) provided its analysis 
of the issue of NSHC based on the model NSHC determination for TSTF-
425.
    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 any accident previously evaluated?
    Response: No.
    The proposed changes relocate the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
significantly increased. The systems and components required by the 
TSs for which the surveillance frequencies are relocated are still 
required to be operable, meet the acceptance criteria for the 
surveillance requirements, and be capable of performing any 
mitigation function assumed in the accident analysis. As a result, 
the consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
changes. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, Dominion 
will perform a probabilistic risk evaluation using the guidance 
contained in NRC approved NEI [Nuclear Energy Institute] 04-10, Rev. 
1, [``Risk-Informed Technical Specifications Initiative 5b Risk-
Informed Method for Control of Surveillance Frequencies,''] in 
accordance with the TS SFCP [Surveillance Frequency Control 
Program]. NEI 04-10, Rev. 1, methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase 
of proposed changes to surveillance frequencies consistent with 
Regulatory Guide 1.177 [``An

[[Page 70772]]

Approach for Plant-Specific, Risk-Informed Decision Making: 
Technical Specifications''].
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
    NRC Branch Chief: Harold K. Chernoff.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: July 22, 2011.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TS) by relocating specific 
Surveillance Frequencies to a licensee-controlled program with the 
adoption of Technical Specification Task Force (TSTF)-425, Revision 3, 
``Relocate Surveillance Frequencies to Licensee Control-Risk Informed 
Technical Specification Task Force (RITSTF) Initiative 5b.''
    The existing Bases information describing the basis for the 
Surveillance Frequency will be relocated to the licensee-controlled 
Surveillance Frequency Control Program. Additionally, the change would 
add a new program, TS 5.5.15, ``Surveillance Frequency Control 
Program,'' to TS Section 5.5, ``Programs and Manuals.''
    The changes are consistent with NRC approved TSTF-425, Revision 3, 
(Rev. 3) (ADAMS Package Accession No. ML090850642). The Federal 
Register notice published on July 6, 2009 (74 FR 31996), announced the 
availability of this TS improvement.
    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 any accident previously 
evaluated?
    Response: No.
    The proposed change relocates the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
significantly increased. The systems and components required by the 
technical specifications for which the surveillance frequencies are 
relocated are still required to be operable, meet the acceptance 
criteria for the surveillance requirements, and be capable of 
performing any mitigation function assumed in the accident analysis. 
As a result, the consequences of any accident previously evaluated 
are not significantly increased. Therefore, the proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, Entergy 
will perform a probabilistic risk evaluation using the guidance 
contained in NRC approved NEI 04-10, Rev. 1 in accordance with the 
TS SFCP [Surveillance Frequency Control Program]. NEI 04-10, Rev. 1, 
methodology provides reasonable acceptance guidelines and methods 
for evaluating the risk increase of proposed changes to surveillance 
frequencies consistent with Regulatory Guide 1.177.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Nancy L. Salgado.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

Renewed Facility Operating License No. DPR-059
    Date of amendment request: August 16, 2011.
    Description of amendment request: The proposed amendment to the 
Renewed Facility Operating License would revise the James A. 
FitzPatrick Nuclear Power Plant (JAF) current licensing basis (CLB) to 
allow the use of On Load Tap Changers (OLTCs) with new Reserve Station 
Service Transformers (RSST) that provide offsite power to JAF.
    The OLTCs are sub-components of two new RSSTs that will be 
installed at JAF in September 2012, during the scheduled refueling 
outage. The OLTCs are designed to compensate for offsite voltage 
variations and will provide added assurance that acceptable bus voltage 
is maintained for safety-related equipment.
    The proposed amendment requests NRC approval to operate the OLTCs 
in the automatic mode. Operation of the OLTCs in the automatic mode was 
evaluated under 10 CFR 50.59 and it was determined that it requires NRC 
approval because such operation creates the possibility for a 
malfunction of a structure, system, or component important to safety 
with a different result than any previously evaluated in the Updated 
Final Safety Analysis Report (UFSAR). The proposed amendment would 
change the UFSAR and the Technical Specification (TS) Bases. There 
would be no changes to the plant TS associated with this request.
    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 amendment will allow operation of the OLTCs in 
automatic mode. The only accident previously evaluated where the 
probability of an accident is potentially affected by the change is 
the loss

[[Page 70773]]

of offsite power (LOOP) Abnormal Operational Transient (AOT). 
Failure of an OLTCs while in the automatic mode of operation that 
results in decreased voltage to the engineered safety features (ESF) 
buses could cause a LOOP if voltage decreased below the degraded 
voltage relay (DVR) setpoint. The two postulated failure scenarios 
are: (1) Failure of an [a] primary microcontroller that results in 
rapidly decreasing voltage supplied to the ESF buses and; (2) 
failure of an [a] primary microcontroller to respond to decreasing 
grid voltage. For the first scenario, a backup microcontroller is 
provided for each OLTC, which makes this failure unlikely. For the 
second scenario, since grid voltage changes typically occur 
relatively slowly and the magnitude of the resulting change would be 
limited to the effect of the change in grid voltage, operators would 
have ample time to address the condition utilizing identified 
procedures. In addition, the frequency of occurrence of these 
failure modes is small, based on the operating history of similar 
equipment at other plants. Furthermore, in both of the above 
potential failure modes, operators can take manual control of the 
OLTC to mitigate the effects of the failure. Thus, the probability 
of a LOOP will not be significantly increased by operation of the 
OLTCs in the automatic mode.
    The proposed amendment has no effect on the consequences of a 
LOOP, since the emergency diesel generators (EDGs) provide power to 
safety-related equipment following a LOOP. The design and function 
of the EDGs are not affected by the proposed change. The probability 
of other previously evaluated accidents is not affected, since the 
proposed amendment does not affect the way plant equipment is 
operated and thus does not contribute to the initiation of any of 
the previously evaluated accidents. The OLTC is equipped with a 
backup microcontroller, which inhibits gross improper action of the 
OLTC in the event of primary microcontroller failure. Additionally, 
the operator has procedurally identified actions available to 
prevent a sustained high voltage condition from occurring. Damage 
due to overvoltage is time-dependent, requiring a sustained high 
voltage condition. Therefore, damage to safety-related equipment is 
unlikely, and the consequences of previously evaluated accidents are 
not significantly increased. Therefore, this proposed amendment does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The proposed amendment involves electrical transformers that 
provide offsite power to safety-related equipment for accident 
mitigation. The proposed change does not alter the design, physical 
configuration, or mode of operation of any other plant structure, 
system, or component. No physical changes are being made to any 
other portion of the plant, so no new accident causal mechanisms are 
being introduced. Although the proposed change potentially affects 
the consequences of previously evaluated accidents (as discussed in 
the response to Question 1), it does not result in any new 
mechanisms that could initiate damage to the reactor or its 
principal safety barriers (i.e., fuel cladding, reactor coolant 
system, or primary containment).
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
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 amendment does not affect the inputs or assumptions 
of any of the analyses that demonstrate the integrity of the fuel 
cladding, reactor coolant system, or containment during accident 
conditions. The allowable values for the degraded voltage protection 
function are unchanged and will continue to ensure that the degraded 
voltage protection function actuates when required, but does not 
actuate prematurely to unnecessarily transfer safety-related loads 
from offsite power to the emergency diesel generators. Automatic 
operation of the OLTCs increases the margin of safety by reducing 
the potential for transferring loads to the EDGs during an under 
voltage or over voltage event on the offsite power sources.
    Therefore, the proposed amendment to the JAF design basis does 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William C. Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White 
Plains, NY 10601.
    NRC Branch Chief: Nancy L. Salgado.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant (PNP), Van Buren County, Michigan

    Date of amendment request: August 16, 2011, as supplemented by 
letter dated October 6, 2011.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 5.5.14, ``Containment Leak 
Rate Testing Program'' to increase the value of the calculated peak 
containment internal pressure from 53 pounds per square inch gauge 
(psig) to 54.2 psig. This increase is due to an increase in the 
calculated mass and energy release during the blowdown phase of the 
design basis loss-of-coolant accident (LOCA). The increase in the 
predicted mass and energy release is due to the correction of an error 
in the calculation of the current value of Pa. The 
regulations at 10 CFR part 50 Appendix J Option B define Pa 
as the calculated peak containment internal pressure related to the 
design basis LOCA as specified in the TS and specifies the requirements 
for containment leakage rate testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to Pa does not alter the assumed 
initiators to any analyzed event. The probability of an accident 
previously evaluated will not be increased by this proposed change.
    The change in Pa will not affect radiological dose 
consequence analyses. PNP radiological dose consequence analyses 
assume a certain containment atmosphere leak rate based on the 
maximum allowable containment leakage rate, which is not affected by 
the change in calculated peak containment internal pressure. The 
Appendix J containment leak rate testing program will continue to 
ensure that containment leakage remains within the leakage assumed 
in the offsite dose consequence analyses. The consequences of an 
accident previously evaluated will not be increased by this proposed 
change.
    Therefore, operation of the facility in accordance with the 
proposed change to Pa will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change provides a higher Pa than 
currently described in the TS. This change is a result of an 
increase in the mass and energy release input for the loss of 
coolant accident containment response analysis. The calculated peak 
containment pressure remains below the containment design pressure 
of 55 psig. This change does not involve any alteration in the plant 
configuration (no new or different type of equipment will be 
installed) or make changes in the methods governing normal plant 
operation. The change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, operation of the facility in accordance with the 
proposed change to TS Section 5.5.14 would not create the 
possibility of a new or different kind of accident from any 
previously evaluated.

[[Page 70774]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The calculated peak containment pressure remains below the 
containment design pressure of 55 psig. Since PNP radiological 
consequence analyses are based on the maximum allowable containment 
leakage rate, which is not being revised, the change in the 
calculated peak containment pressure does not represent a 
significant change in the margin of safety.
    Therefore, operation of the facility in accordance with the 
proposed change to TS Section 5.5.14 does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Branch Chief: Robert J. Pascarelli.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center (DAEC), Linn County, Iowa

    Date of amendment request: May 31, 2011.
    Description of amendment request: The proposed amendment would 
upgrade selected DAEC Emergency Action Levels (EALs) based on NEI 99-
01, Revision 5, ``Methodology for Development of Emergency Action 
Levels,'' using the guidance of NRC Regulatory Issue Summary 2003-18, 
Supplement 2, ``Use of Nuclear Energy Institute (NEI) 99-01, 
Methodology for Development of Emergency Action Levels.'' NextEra 
Energy Duane Arnold currently uses an emergency classification scheme 
based on NEI 99-01, Revision 4.
    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 any accident previously 
evaluated?
    Response: No.
    These changes affect the DAEC Emergency Plan and do not alter 
any of the requirements of the Operating License or the Technical 
Specifications. The proposed changes do not modify any plant 
equipment and do not impact any failure modes that could lead to an 
accident. Additionally, the proposed changes do not impact the 
consequence of any analyzed accident since the changes do not affect 
any equipment related to accident mitigation.
    Based on this discussion, the proposed amendment does not 
increase the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    These changes affect the DAEC Emergency Plan and do not alter 
any of the requirements of the Operating License or the Technical 
Specifications. They do not modify any plant equipment and there is 
no impact on the capability of the existing equipment to perform 
their intended functions. No system setpoints are being modified and 
no changes are being made to the method in which plant operations 
are conducted. No new failure modes are introduced by the proposed 
changes. The proposed amendment does not introduce accident 
initiator or malfunctions that would cause a new or different kind 
of accident.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    These changes affect the DAEC Emergency Plan and do not alter 
any of the requirements of the Operating License or the Technical 
Specifications. The proposed changes do not affect any of the 
assumptions used in the accident analysis, nor do they affect any 
operability requirements for equipment important to plant safety.
    Therefore, the proposed changes will not result in a significant 
reduction in the margin of safety as defined in the bases for 
technical specifications covered in this license amendment request.

    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, 801 Pennsylvania Avenue 
NW., Suite 220, Washington, DC 20004.
    NRC Branch Chief: Robert J. Pascarelli.

South Carolina Electric and Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station (VCSNS), 
Unit 1, Fairfield County, South Carolina

    Date of amendment request: August 23, 2011.
    Description of amendment request: The proposed amendment would 
delete the license condition, 2.G.1 of the Facility Operating License, 
that requires reporting of violations of Section 2.C of the Facility 
Operating License consistent with the Federal Register notice dated 
November 4, 2005 (70 FR 67202) as part of the consolidated line item 
improvement process (CLIIP). The proposed amendment would also delete a 
reporting requirement in the VCSNS Technical Specifications (TS), 
Section 6.6, which is duplicative of NRC regulations, and make 
appropriate adjustments to the TS index to reflect that deletion.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
referenced the NRC staffs model no significant hazards consideration, 
presented in a Federal Register notice (70 FR 51098; August 29, 2005), 
and made available for use by Federal Register notice (70 FR 67202; 
November 4, 2005), and 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 involves the deletion of a reporting 
requirement. The change does not affect plant equipment or operating 
practices and therefore does not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change is administrative in that it deletes a 
reporting requirement. The change does not add new plant equipment, 
change existing plant equipment, or affect the operating practices 
of the facility. Therefore, the change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change deletes a reporting requirement. The change 
does not affect plant equipment or operating practices and therefore 
does not involve a significant reduction in a margin of safety.

    Based on the above, the NRC staff proposes that the change presents 
no significant hazards consideration under the standards set forth in 
10 CFR 50.92(c).
    Attorney for licensee: J. Hagood Hamilton, Jr., South Carolina 
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 
29218.
    NRC Branch Chief: Gloria Kulesa.

[[Page 70775]]

Tennessee Valley Authority (TVA), Docket No. 50-328, Sequoyah Nuclear 
Plant, Unit 2, Hamilton County, Tennessee

    Date of amendment request: August 31, 2011 (TS-SQN-2011-03).
    Description of amendment request: During Sequoyah Nuclear Plant 
(SQN), Unit 2, spring 2011 refueling outage (RFO), two penetrations 
through the shield building (SB) dome were created. To maintain SB 
integrity, these penetrations were closed with a steel hatch assembly 
prior to entering Mode 4 at the end of the RFO. The proposed amendment 
would temporarily revise the technical specifications to allow opening 
of one of the penetration hatches in the SB dome for up to 5 hours per 
day, 6 days per calendar week while in Modes 1 through 4 during SQN, 
Unit 2 Cycle 18, and until entering Mode 5 at the start of the SQN, 
Unit 2 fall 2012 RFO. The two approximately 18-inch diameter 
penetrations on the SB dome will provide steam generator replacement 
project workers an alternate path of moving materials inside the 
annulus for online work. Without use of the SB dome penetration 
hatches, materials would travel through the auxiliary building (AB), to 
the annulus access door, and be hoisted up the annual access ladders. 
Bypassing the AB and the annulus access ladders reduces the risk of 
potential adverse effects to sensitive equipment along the path. The 
alternate path is estimated to save approximately 2.8 roentgen 
equivalent man by allowing materials to be passed through the open SB 
dome penetration hatch in lieu of carrying the material past higher 
dose areas. In addition, passing material through the open SB dome 
hatch will significantly improve the industrial safety aspect of the 
work and will provide work efficiency gains since material will be 
provided closer to the point of use.
    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 bounding transients and accidents (i.e., loss-of-coolant-
accident (LOCA), tornado, and earthquake) that are potentially 
affected by the assumptions associated with the use of one of the 
Shield Building dome penetration hatches (2-EQH-410-0010 or 2-EQH-
410-0011) have been evaluated/analyzed. Weather and seismic related 
events are determined by regional conditions. Therefore, the 
probability of a tornado or earthquake is not affected by the use of 
one of the Shield Building dome penetration hatches. Failure of the 
Shield Building or Emergency Gas Treatment System (EGTS) is not an 
initiator of any of the accidents and transients described in the 
Updated Final Safety Analysis Report (UFSAR). Therefore, since no 
initiating event mechanisms are being changed, the use of one of the 
Shield Building dome penetration hatches will not result in an 
increase in probability of any previously evaluated accident.
    The use of one of the Shield Building dome penetration hatches 
affects the integrity of the Shield Building and the ability of the 
EGTS to maintain the annulus at a negative pressure relative to the 
outside atmosphere such that the function in mitigating the 
radiological consequences of an accident is affected. TVA's 
evaluation documents the radiological consequences of a LOCA 
assuming the open Shield Building dome penetration hatch is closed 
within 22.1 minutes and the operating EGTS trains draw down the 
annulus to -0.25 inches wg [water gauge] to effectively end the 
direct release of radionuclides to the environment 23.1 minutes 
after accident initiation. TVA's evaluation also documents the 
mission dose an individual may receive during ingress from the 
Control Building Habitability area to the Shield Building dome, 
closure of the steel hatch assembly, and egress from the Shield 
Building dome. Although the LOCA radiological consequences with the 
Shield Building dome penetration hatch open for 22.1 minutes (and 
assumed to be a direct release path for 23.1 minutes) are higher 
than those described in the UFSAR, the offsite and Control Room 
doses remain within the limits of 10 CFR 50.67, ``Accident source 
term,'' when applying the Alternate Source Term (AST) methodology in 
accordance with Regulatory Guide 1.183, ``Alternative Radiological 
Source Terms for Evaluating Design Basis Accidents at Nuclear Power 
Reactors,'' dated July 2000. The calculated mission doses are also 
less than the limits of 10 CFR 50.67, ``Accident source term,'' 
paragraph (b)(2)(iii) when applying the AST methodology in 
accordance with Regulatory Guide 1.183.
    Therefore, since the increase in radiological consequences of 
the previously evaluated LOCA remains bounded by the applicable 
regulatory limits, the increased consequences are not considered 
significant.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Loss of Shield Building integrity or EGTS failure is not an 
initiator of any of the accidents and transients described in the 
UFSAR. Shield Building integrity as the pressure boundary for the 
EGTS, and loss of Shield Building integrity due to an open 
penetration hatch in the Shield Building dome (Hatch 2-EQH-410-0010 
or 2-EQH-410-0011) during Modes 1 through 4 potentially renders both 
trains of EGTS incapable of establishing a post-accident annulus 
pressure. This condition would require SQN, Unit 2, to enter the 
Action of TS [Technical Specification] Limiting Condition for 
Operation (LCO) 3.6.1.8 (for the condition of one train of EGTS 
being inoperable) and enter TS LCO 3.0.3 (due to both trains of EGTS 
being inoperable). TS LCO 3.0.3 requires that the unit be shutdown 
within specified time periods. Closure of the open Shield Building 
dome penetration steel hatch assembly restores the integrity of the 
Shield Building such that both trains of EGTS would be operable as 
required by TS LCO 3.6.1.8. Failure of the Shield Building dome 
penetration steel hatch assemblies will not initiate any of the 
accidents and transients described in the UFSAR. Postulated failures 
of the Shield Building dome penetration steel hatch assemblies are 
degradation/damage to the seals or damage to the hatch hinges. Like 
any other Shield Building failure during Modes 1 through 4 that 
potentially renders both trains of EGTS inoperable, these postulated 
Shield Building dome penetration steel hatch assembly failures 
result in a loss of Shield Building integrity and require that the 
failed component be repaired or replaced within a specified time 
period or that plant shutdown be initiated.
    Therefore, a failure of a steel hatch assembly during use of the 
Shield Building dome penetration will not initiate an accident nor 
create any new failure mechanisms. The changes do not result in any 
event previously deemed incredible being made credible. The use of 
Shield Building dome Penetration Hatch 2-EQH-410-0010 or 2-EQH-410-
0011 is not expected to result in more adverse conditions in the 
annulus and is not expected to result in any increase in the 
challenges to safety systems.
    Manual action is required to close an open Shield Building dome 
penetration hatch and to configure the EGTS control loops following 
the opening and closing of a Shield Building dome penetration hatch 
such that the EGTS will respond as designed. NRC Information Notice 
(IN) 97-78, ``Crediting of Operator Actions in Place of Automatic 
Actions and Modifications of Operator Actions, Including Response 
Times,'' and American National Standards Institute/American Nuclear 
Society (ANSI/ANS)-58.8, ``Time Response Design Criteria for Safety-
Related Operator Actions,'' provide guidance for consideration of 
safety-related operator actions.
    The manual actions implemented as a result of this change can be 
completed within the guidance and criteria provided in Information 
Notice (IN) 97-78 and ANSI/ANS-58.8. Consequently, the manual 
actions can be credited in the mitigation of events that require 
Shield Building integrity. With credit for the manual actions to 
close an open Shield Building dome penetration hatch (2-EQH-410-0010 
or 2-EQH-410-0011) and reconfigure the EGTS control loops subsequent 
to an event, the types of accidents currently evaluated in the UFSAR 
remain the same.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

[[Page 70776]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The manual actions to close an open Shield Building dome 
penetration hatch (2-EQH-410-0010 or 2-EQH-410-0011) and to 
configure the EGTS control loops following the opening and closing 
of a Shield Building dome penetration hatch ensure that the EGTS 
will respond as designed. Safety-related instrumentation is 
available to inform operators that a reactor trip has occurred, and 
dedicated trained individuals will be positioned to close an open 
Shield Building dome penetration hatch should an accident occur. The 
manual actions meet the criteria for safety-related operator actions 
contained in NRC IN 97-78 and ANSI/ANS-58.8. The use of manual 
actions maintains the margin of safety by assuring compliance with 
acceptance limits reviewed and approved by the NRC. The appropriate 
acceptance criteria for the various analyses and evaluations have 
been met; therefore, there has not been a reduction in any margin of 
safety.
    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, Tennessee 37902.
    NRC Branch Chief: Douglas A. Broaddus.

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 NRC's PDR, located at One White Flint North, Room O1-F21, 11555 
Rockville Pike (first floor), Rockville, Maryland 20852. Publicly 
available documents created or received at the NRC are accessible 
electronically through the Agencywide Documents Access and Management 
System (ADAMS) in the NRC Library at 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 email to 
[email protected].

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412 Beaver Valley Power Station, Unit 1 and 2, Beaver County, 
Pennsylvania

    Date of application for amendment: April 29, 2011.
    Brief description of amendment: The amendments will modify 
Technical Specification (TS) to define a new time limit for restoring 
inoperable reactor coolant system (RCS) leakage detection 
instrumentation to operable status and establish alternative methods of 
monitoring RCS leakage when one or more require monitors are 
inoperable. The changes are consistent with Nuclear Regulatory 
Commission-approved Technical Specification Task Force Traveler-513, 
Revision 3. The availability of this TS improvement was published in 
the Federal Register on January 3, 2011 (76 FR 189), as part of the 
consolidated line item improvement process.
    Date of issuance: October 25, 2011.
    Effective date: As of the date of issuance, and shall be 
implemented within 90 days from the. date of issuance.
    Amendment Nos: 288 and 175.
    Renewed Facility Operating License Nos. DPR-66 and NPF-73: The 
amendments revised the License and TS.
    Date of initial notice in Federal Register: July 12, 2011 (76 FR 
40940).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 25, 2011.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Nuclear Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of application for amendments: August 5, 2010, supplemented by 
letters dated February 22, May 20, September 14, and September 22, 
2011.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 5.5.1 Fuel Storage--Criticality, to include new 
spent fuel storage patterns that account for both the increase in fuel 
maximum enrichment from 4.5 weight (wt) percent (%) U-235 to 5.0 wt% U-
235 and the impact on the fuel of higher power operation proposed under 
the Extended Power Uprate license amendment request. Although the fuel 
storage has been analyzed at the higher fuel enrichment in the new 
criticality analysis, the fuel enrichment limit of 4.5 wt% U-235 
specified in TS 5.5.1 will not be changed with the issuance of these 
license amendments.
    Date of issuance: October 31, 2011.
    Effective date: As of the date of issuance and shall be implemented 
by the completion of the Cycle 26 refueling outage for Unit 3 and Cycle 
27 refueling outage for Unit 4.
    Amendment Nos.: Unit 3--246 and Unit 4--242.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: October 5, 2010 (75 FR 
61527). The supplements dated February 22, May 20, September 14, and 
September 22, 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 as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 31, 2011.
    No significant hazards consideration comments received: No.

[[Page 70777]]

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: October 29, 2010, as supplemented by 
letters dated June 10 and August 31, 2011.
    Brief description of amendment: The amendment revised the 
acceptance criteria in CNS Technical Specification (TS) 3.8.4, ``DC 
[Direct Current] Sources--Operating,'' Surveillance Requirement (SR) 
3.8.4.1, and TS 3.8.6, ``Battery Cell Parameters,'' Table 3.8.6-1, 
``Battery Cell Parameter Requirements.'' Specifically, amendment 
revised the acceptance criteria in TS SR 3.8.4.1 and TS Table 3.8.6-1 
by revising the battery terminal voltage on float charge and specific 
gravity acceptance criteria to ensure that the safety-related batteries 
can perform their safety functions and will remain operable during 
postulated design basis events.
    Date of issuance: October 28, 2011.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 239.
    Renewed Facility Operating License No. DPR-46: Amendment revised 
the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 25, 2011 (76 FR 
4386). The supplemental letters dated June 10 and August 31, 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 as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 28, 2011.
    No significant hazards consideration comments received: No.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station, Unit 1 (NMP1), Oswego County, New York

    Date of application for amendment: November 2, 2010, as 
supplemented on January 27, 2011.
    Brief description of amendment: The amendment revises the NMP1 
Technical Specification (TS) Section 3.6.2, ``Protective 
Instrumentation,'' by modifying the operability requirements for the 
average power range monitoring (APRM) instrumentation system. The 
amendment eliminates the requirements that the APRM ``Upscale'' and 
``Inoperative'' scram and control rod withdrawal block functions be 
operable when the reactor mode switch is in the Refuel position. The 
amendment also clarifies the operability requirements for the APRM 
``Downscale'' control rod withdrawal block function when the reactor 
mode switch is in the Startup and Refuel positions.
    Date of issuance: October 31, 2011.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No.: 211.
    Renewed Facility Operating License No. DPR-63: The amendment 
revises the License and TSs.
    Date of initial notice in Federal Register: March 22, 2011 (76 FR 
16007). The supplemental letter dated January 27, 2011, provided 
additional information that clarified the application and did not 
expand the scope of the application as originally noticed, and did not 
change the Nuclear Regulatory Commission staff's initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 31, 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: March 30, 2010, as supplemented 
on June 1 and December 29, 2010, and January 14, February 25, April 27, 
and July 25, 2011.
    Brief description of amendment: The amendment changes the NMP2 
Technical Specification (TS) 3.8.1, ``AC Sources--Operating,'' to 
extend the Completion Time (CT) for an inoperable Division 1 or 
Division 2 diesel generator (DG) from 72 hours to 14 days.
    Date of issuance: October 31, 2011.
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No.: 138.
    Renewed Facility Operating License No. NPF-069: The amendment 
revises the License and TSs.
    Date of initial notice in Federal Register: July 13, 2010 (75 FR 
39980). The supplemental letters dated June 1 and December 29, 2010, 
and January 14, February 25, April 27, and July 25, 2011, provided 
additional information that clarified the application and did not 
expand the scope of the application as originally noticed, and did not 
change the Nuclear Regulatory Commission staff's initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 31, 2011.
    No significant hazards consideration comments received: No.

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 application for amendments: December 22, 2009, as 
supplemented by letters dated July 23, 2010, August 20, 2010, October 
8, 2010, January 14, 2011, February 23, 2011, April 6, 2011, and August 
9, 2011.
    Brief description of amendments: The amendments approve the 
application of the leak-before-break methodology to certain piping 
systems attached to the reactor coolant system at the Prairie Island 
Nuclear Generating Plant, Units 1 and 2.
    Date of issuance: October 27, 2011.
    Effective date: As of the date of issuance. The amendment for Unit 
1 shall be implemented within 180 days. The amendment for Unit 2 shall 
be implemented before the end of the next scheduled Unit 2 refueling 
outage.
    Amendment Nos.: 204, 191.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: May 11, 2010 (75 FR 
26290). The supplemental letters contained clarifying information and 
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 amendments is contained 
in a Safety Evaluation dated October 27, 2011.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 7th day of November 2011.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2011-29435 Filed 11-14-11; 8:45 am]
BILLING CODE 7590-01-P