[Federal Register Volume 77, Number 200 (Tuesday, October 16, 2012)]
[Notices]
[Pages 63343-63355]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2012-25240]


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

NUCLEAR REGULATORY COMMISSION

[NRC-2012-0236]


Biweekly Notice: Applications and Amendments to Facility 
Operating Licenses and Combined 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) 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 or 
combined license, as applicable, 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 September 20, 2012 to October 3, 2012. The 
last biweekly notice was published on October 2, 2012 (77 FR 60146-
60160).
    Addresses: You may access information and comment submissions 
related to this document, which the NRC possesses and are publicly 
available, by searching on http://www.regulations.gov under Docket ID 
NRC-2012-0236. You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0236. 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.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
Supplementary Information section of this document.
    Supplementary Information:

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2012-0236 when contacting the NRC 
about the availability of information regarding this document. You may 
access information related to this document, which the NRC possesses 
and are publicly available, by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0236.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the 
search,

[[Page 63344]]

select ``ADAMS Public Documents'' and then select ``Begin Web-based 
ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in 
ADAMS by performing a search on the document date and docket number.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2012-0236 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses and Combined 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 section 50.92 of Title 10 of the Code 
of Federal Regulations (10 CFR), 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.
    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 or 
combined 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. The NRC regulations are accessible electronically from the NRC 
Library on the NRC's 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 
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

[[Page 63345]]

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 the 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's 
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's 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's 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 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. Requests for hearing, 
petitions for leave

[[Page 63346]]

to intervene, and motions for leave to file new or amended contentions 
that are filed after the 60-day deadline will not be entertained absent 
a determination by the presiding officer that the filing demonstrates 
good cause by satisfying the following three factors in 10 CFR 
2.309(c)(1): (i) The information upon which the filing is based was not 
previously available; (ii) the information upon which the filing is 
based is materially different from information previously available; 
and (iii) the filing has been submitted in a timely fashion based on 
the availability of the subsequent information.
    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].

Carolina Power and Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina.

    Date of amendment request: June 19, 2012.
    Description of amendment request: The proposed license amendments 
would revise the Technical Specifications (TS) for the Brunswick Steam 
Electric Plant (BSEP), Units 1 and 2. The TS change proposes to extend 
the Completion Time (CT) of TS 3.8.1 Required Action D.4 for an 
inoperable diesel generator (DG). A commensurate change is also 
proposed to extend the maximum CT of TS 3.8.1 Required Actions C.3 and 
D.4. The licensee stated that it will add a supplemental alternating 
current power source (i.e., a supplemental diesel generator) with the 
capability to power any E-bus within one hour from the Station Blackout 
(SBO) event, and with the capacity to bring the affected unit to cold 
shutdown, to support 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. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The DGs are safety related components which provide backup 
electrical power supply to the onsite emergency power distribution 
system. The proposed changes do not affect the design of the DGs, 
the operational characteristics or function of the DGs, the 
interfaces between the DGs and other plant systems, or the 
reliability of the DGs. The DGs are not accident initiators; the DGs 
are designed to mitigate the consequences of previously evaluated 
accidents including a loss of offsite power. Extending the CT for a 
single DG would not affect the previously evaluated accidents since 
the remaining DGs supporting the redundant ESF [engineered safety 
feature] systems would continue to be available to perform the 
accident mitigation functions. Thus, allowing a DG to be inoperable 
for an additional 7 days for performance of maintenance or testing 
does not increase the probability of a previously evaluated 
accident.
    Deterministic and probabilistic risk assessments evaluated the 
effect of the proposed TS changes on the availability of an 
electrical power supply to the plant emergency safeguards features 
systems. These assessments concluded that the proposed TS changes do 
not involve a significant increase in the risk of power supply 
unavailability.
    There is small incremental risk associated with continued 
operation for an additional 7 days with one DG inoperable; however, 
the calculated impact on risk provides risk metrics consistent with 
the acceptance guidelines contained in RG [Regulatory Guide] 1.177 
and 1.174 (References 7.2.1 and 7.2.2). This risk is judged to be 
reasonably consistent with the risk associated with operations for 7 
days with one DG inoperable as allowed by the current TS.
    Specifically, the remaining operable DGs and paths are adequate 
to supply electrical power to the onsite emergency power 
distribution system. A DG is required to operate only if both 
offsite power sources fail and there is an event which requires 
operation of the plant engineered safety features such as a design 
basis accident. The probability of a design basis accident occurring 
during this period is low.
    The consequences of previously evaluated accidents will remain 
the same during the proposed 14-day CT as during the current 7-day 
CT. The ability of the remaining TS required DG to mitigate the 
consequences of an accident will not be affected since no additional 
failures are postulated while equipment is inoperable within the TS 
CT. The standby AC [alternating current] power supply for each of 
the four safety-related load groups consists of one DG complete with 
its auxiliaries, which include the cooling water, starting air, 
lubrication, intake and exhaust, and fuel oil systems. The sizing of 
the DGs and the loads assigned among them is such that any 
combination of three out of four of these DGs is capable of shutting 
down the plant safely, maintaining the plant in a safe shutdown 
condition, and mitigating the consequences of accident conditions.
    Thus this change does not involve a significant increase in the 
probability or consequences of a previously analyzed accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a change in the plant 
design, plant configuration, system operation, or procedures 
involved with the DGs. The proposed changes allow a DG to be 
inoperable for additional time. Equipment will be operated in the 
same configuration and manner that is currently allowed and designed 
for. The functional demands on credited equipment is unchanged. 
There are no new failure modes or mechanisms created due to plant 
operation for an extended period to perform DG maintenance or 
testing. Extended operation with an inoperable DG does not involve 
any modification in the operational limits or physical design of 
plant systems. There are no new accident precursors generated due to 
the extended CT.
    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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    Currently, if an inoperable DG is not restored to operable 
status within 7 days, TS 3.8.1, Condition H, requires the unit to be 
in MODE 3 (i.e., HOT SHUTDOWN) within a CT of 12 hours, and to be in 
MODE 4 (i.e., COLD SHUTDOWN) within a CT of 36 hours. This TS 
Condition is entered on both units resulting in a dual-unit 
shutdown. The proposed Technical Specification changes will allow 
steady state plant operation at 100 percent power for an additional 
7 days for performance of DG planned reliability improvements and 
preventive and corrective maintenance.
    Deterministic and probabilistic risk assessments evaluated the 
effect of the proposed TS changes on the availability of an 
electrical power supply to the plant ESF systems. These assessments 
concluded that the proposed TS changes do not involve a significant 
increase in the risk of power supply unavailability.
    The DGs continue to meet their design requirements; there is no 
reduction in capability or change in design configuration. The DG 
response to LOOP [loss of offsite power], LOCA [loss-of-coolant 
accident], SBO [station blackout], or fire is not changed by this 
proposed amendment; there is no change to the DG operating 
parameters. In the extended CT, as in the existing CT, the remaining 
operable DGs and paths are adequate to supply electrical power to 
the onsite emergency power distribution system. The proposed change 
does not alter a design basis or safety limit; therefore, it does 
not significantly reduce the margin of safety. The DGs will continue 
to operate per the existing design and regulatory requirements.

[[Page 63347]]

    The proposed TS changes do not alter the plant design nor does 
it change the assumptions contained in the safety analyses. The 
standby AC power system is designed with sufficient redundancy such 
that a DG may be removed from service for maintenance or testing. 
The remaining DGs are capable of carrying sufficient electrical 
loads to satisfy the UFSAR [updated final safety analysis report] 
requirements for accident mitigation or unit safe shutdown. The 
proposed changes do not impact the redundancy or availability 
requirements of offsite power circuits or change the ability of the 
plant to cope with a SBO.
    Therefore, based on the considerations given above, the proposed 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Senior Counsel--Manager 
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, NC 27602.
    NRC Acting Branch Chief: Jessie Quichocho.

Carolina Power and Light Company, Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina; Florida Power Corporation, et al., Docket No. 50-302, Crystal 
River Unit 3 Nuclear Generating Plant, Citrus County, Florida

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina; Carolina Power & Light Company, Docket No. 50-261, H. B. 
Robinson Steam Electric Plant, Unit 2, Darlington County, South 
Carolina

    Date of amendment request: September 12, 2012.
    Description of amendment request: The proposed license amendments 
would revise the Facility Operating Licenses for the Brunswick Steam 
Electric Plant, Units 1 and 2, H. B. Robinson Steam Electric Plant, 
Unit 2, Shearon Harris Nuclear Power Plant, Unit 1, and Crystal River 
Unit No. 3 Nuclear Generating Plant. The NRC issued license amendments, 
dated July 29, 2011, that approved the licensees' cyber security plan 
and associated implementation milestone schedule. Milestone 6 requires 
the identification, documentation, and implementation of cyber security 
controls for critical digital assets that could adversely impact the 
design function of physical security target set equipment by no later 
than December 31, 2012. The license amendment request would change the 
existing facility operating licenses for the Physical Protection/
Security license condition for these plants to reference the change to 
an implementation schedule milestone and a proposed Revised Cyber 
Security Plan Implementation Schedule for the scope of Milestone 6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the Cyber Security Plan Implementation 
Schedule is administrative in nature. This change does not alter 
accident analysis assumptions, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected. The proposed 
change does not require any plant modifications which affect the 
performance capability of the structures, systems, and components 
relied upon to mitigate the consequences of postulated accidents and 
has no impact on the probability or consequences of an accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the Cyber Security Plan Implementation 
Schedule is administrative in nature. This proposed change does not 
alter accident analysis assumptions, add any initiators, or affect 
the function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected. The proposed 
change does not require any plant modifications which affect the 
performance capability of the structures, systems, and components 
relied upon to mitigate the consequences of postulated accidents and 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed change to 
the Cyber Security Plan Implementation Schedule is administrative in 
nature. Because there is no change to these established safety 
margins as result of this change, the proposed change does not 
involve a significant reduction in a 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: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, 
Charlotte, NC 28202.
    NRC Acting Branch Chief: Jessie Quichocho.

Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit 2, (HBRSEP) Darlington County, South 
Carolina

    Date of amendment request: August 6, 2012.
    Description of amendment request: The proposed change would revise 
the Technical Specification (TS) requirements for inoperable snubbers 
by adding Limiting Condition for Operation (LCO) 3.0.8. The change is 
consistent with NRC approved Revision 4 to Technical Specification Task 
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
372, ``Addition of LCO 3.0.8, Inoperability of Snubbers.''
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on May 4, 2005 (70 FR 23252). The licensee affirmed 
the applicability of the model NSHC determination in its application 
dated August 6, 2012.
    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 allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable

[[Page 63348]]

snubber if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. Therefore, the probability 
of an accident previously evaluated is not significantly increased, 
if at all. The consequences of an accident while relying on 
allowance provided by proposed LCO 3.0.8 are no different than the 
consequences of an accident while relying on the TS required actions 
in effect without the allowance provided by proposed LCO 3.0.8. 
Therefore, the consequences of an accident previously evaluated are 
not significantly affected by this change. The addition of a 
requirement to assess and manage the risk introduced by this change 
will further minimize possible concerns.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    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 does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to inoperable snubbers, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns.
    Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an inoperable 
snubber, if risk is assessed and managed. The postulated seismic 
event requiring snubbers is a low-probability occurrence and the 
overall TS system safety function would still be available for the 
vast majority of anticipated challenges. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in RG [Regulatory Guide] 1.177. A bounding risk 
assessment was performed to justify the proposed TS changes. This 
application of LCO 3.0.8 is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The net change to the margin of safety is insignificant.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Manager--Senior Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Acting Branch Chief: Jessie F. Quichocho.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit 2, (HBRSEP) Darlington County, South Carolina

    Date of amendment request: August 29, 2012.
    Description of amendment request: The proposed change combines two 
changes that affect the same Technical Specification (TS) sections into 
one license amendment. The first part proposes to implement revisions 
consistent with Technical Specification Task Force (TSTF)-510, Revision 
2, ``Revision to Steam Generator (SG) Program Inspection Frequencies 
and Tube Sample Selection.'' The second part proposes to permanently 
revise TS 5.5.9 ``Steam Generator Program'' to exclude portions of the 
SG tube below the top of the SG tubesheet from periodic inspections by 
implementing the permanent alternate repair criteria ``H.*'' References 
2, 3, 8, 23 and 32 referred to in the licensees analysis can be found 
in the license amendment request under ADAMS Accession No. ML12275A176.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The previously analyzed accidents are initiated by the failure 
of plant structures, systems, or components. The proposed change 
modifies steam generator tube inspection frequencies and tube 
selection consistent with TSTF-510 and excludes the lower portion of 
steam generator tubes from inspection by implementing the alternate 
repair criteria (H*) on a permanent basis and does not have a 
detrimental impact on the integrity of any plant structure, system, 
or component that initiates an analyzed event. The proposed change 
will not alter the operation of, or otherwise increase the failure 
probability of any plant equipment that initiates an analyzed 
accident.
    Of the applicable accidents previously evaluated, the limiting 
transients with consideration to the proposed change to the SG tube 
inspection and repair criteria are the SG tube rupture (SGTR) event 
and the main steam line break (MSLB) postulated accident.
    The proposed SG tube inspection frequency and sample selection 
criteria will continue to ensure that the SG tubes are inspected 
such that the probability of a SGTR is not increased. The 
consequences of a SGTR are bounded by the conservative assumptions 
in the design basis accident analysis. The proposed SG tube 
inspection frequency and sample selection criteria will not cause 
the consequences of a SGTR to exceed those assumptions.
    With respect to the SGTR event, the required structural 
integrity margins of the SG tubes and the tube-to-tubesheet joint 
over the H* distance will be maintained. Tube rupture in tubes with 
cracks within the tubesheet is precluded by the constraint provided 
by the presence of the tubesheet and the tube-to-tubesheet joint. 
Tube burst cannot occur within the thickness of the tubesheet. The 
tube-to-tubesheet joint constraint results from the hydraulic 
expansion process, thermal expansion mismatch between the tube and 
tube sheet, and from the differential pressure between the primary 
and secondary side, and tube sheet rotation. The structural margins 
against burst, as discussed in Regulatory Guide [RG] 1.121, ``Bases 
for Plugging Degraded PWR [Pressurized-Water Reactor] Steam 
Generator Tubes'' (Reference 32) and [Nuclear Energy Institute] NEI 
97-06, ``Steam Generator Program Guidelines,'' (Reference 8) are 
maintained for both normal and postulated accident conditions.
    For the portion of the tube outside of the tubesheet, the 
proposed change also has no impact on the structural or leakage 
integrity. Therefore, the proposed change does not result in a 
significant increase in the probability of the occurrence of a SGTR 
accident.
    At normal operating pressures, leakage from degradations below 
the proposed limited inspection depth is limited by the tube-to-
tubesheet crevice. Consequently, negligible normal operating leakage 
is expected from degradation below the inspected depth within the 
tubesheet region. The consequences of an SGTR event are affected by 
the primary to secondary leakage flow during the event. However, 
primary to secondary leakage flow through a postulated tube that has 
been pulled out of the tubesheet is not affected by the proposed 
changes since the tubesheet enhances the tube integrity in the 
region of the hydraulic expansion by precluding tube deformation 
beyond its initial hydraulically expanded outside diameter. 
Therefore, the proposed change does not result in a significant 
increase in the consequences of an SGTR. In addition, the selected 
H* value envelopes the depth within the tubesheet required to 
prevent a tube pullout.
    The probability of a MSLB event is unaffected by the potential 
failure of a SG tube as the failure of a tube is not an initiator 
for a MSLB event. Therefore the proposed SG tube inspection 
frequency and sample selection criteria and the structural integrity 
margins of the SG tubes and the tube-to-tubesheet joint over the H* 
distance do not increase the probability of a MSLB event.

[[Page 63349]]

    The leak rate factor of 1.82 for HBRSEP, for a postulated MSLB, 
has been calculated as shown in References 2, 3 and 23. HBRSEP Unit 
No. 2 will apply the factor of 1.82 to the normal operating leakage 
associated with the tubesheet expansion region in the condition 
monitoring and operational assessment. Through application of the 
limited tube sheet inspection scope, the existing operating leakage 
limit provides assurance that excessive leakage (i.e., greater than 
accident analysis assumptions) will not occur.
    When the TS operational leak rate limit of 75 [gallons per day] 
gpd or about 0.052 gallons per minute (gpm) through any one SG is 
multiplied by the MSLB leak rate factor applicable to HBRSEP Unit 
No. 2 of 1.82 (Table 9-7 in [Westinghouse Commercial Atomic Power 
Report] WCAP-17091-P, Reference 3) the maximum primary to secondary 
accident induced leak rate is less than 0.095 gpm and is bounded by 
the value of 0.11 gpm through the faulted SG used in the MSLB 
accident analyses. Since the existing limit on operational leakage 
continues to ensure that the MSLB assumed accident induced leakage 
will not be exceeded, the consequences of a MSLB accident are not 
increased.
    For the condition monitoring assessment, the component of 
leakage from the prior cycle from below the H* distance will be 
multiplied by a factor of 1.82 and added to the total leakage from 
any other source and compared to the allowable accident induced leak 
rate. For the operational assessment, the difference in the leakage 
between the allowable leakage and the calculated accident induced 
leakage from sources other than the tubesheet expansion region will 
be divided by 1.82 and compared to the observed operational leakage.
    Based on the above, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change modifies steam generator tube inspection 
frequencies and tube selection consistent with TSTF-510 and excludes 
the lower portion of steam generator tubes from inspection by 
implementing the alternate repair criteria (H*) on a permanent 
basis. The proposed change does not introduce any new equipment, 
create new failure modes for existing equipment, or create any new 
limiting single failures resulting from tube degradation. The 
proposed change does not affect the design of the SGs or their 
method of operation. In addition, the proposed change does not 
impact any other plant system or component. Plant operation will not 
be altered, and all safety functions will continue to perform as 
previously assumed in accident analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the change involve a significant reduction in the margin 
of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes. Steam 
generator tube integrity is a function of the design, environment, 
and the physical condition of the tube. The proposed change does not 
affect tube design or operating environment. The proposed change 
will continue to require monitoring of the physical condition of the 
SG tubes but will limit inspection within the tubesheet to the 
portion of the tube from the top of the tubesheet to a distance H* 
below the top of the tubesheet.
    The proposed change modifies steam generator tube inspection 
frequencies and tube selection consistent with TSTF-510 and limits 
required inspection to the safety significant portion of the steam 
generator tubes. WCAP-17345, Rev. 2 (Reference 2) identifies the 
specific inspection depth (H*) below which any type of tube 
degradation is shown to have no impact on the performance criteria 
in NEI 97-06 Rev. 3, ``Steam Generator Program Guidelines'' 
(Reference 8) and TS 5.5.9, ``Steam Generator (SG) Program.'' 
Changes associated with inspection frequency and tube selection 
criteria are consistent with TSTF-510 and are based on recent 
industry experience and are more effective in managing the frequency 
of verification of tube integrity and sample selection than those 
required by current TSs.
    The proposed change maintains the required structural margins of 
the SG tubes for both normal and accident conditions. Nuclear Energy 
Institute 97-06, ``Steam Generator Program Guidelines'' (Reference 
8), and NRC Regulatory Guide 1.121 ``Bases for Plugging Degraded PWR 
Steam Generator Tubes'' (Reference 32), are used as the bases in the 
development of the limited tubesheet inspection depth methodology 
for determining that SG tube integrity considerations are maintained 
within acceptable limits. Regulatory Guide 1.121 describes a method 
acceptable to the NRC for meeting General Design Criteria (GDC) 14, 
``Reactor Coolant Pressure Boundary,'' GDC 15, ``Reactor Coolant 
System Design,'' GDC 31, ``Fracture Prevention of Reactor Coolant 
Pressure Boundary,'' and GDC 32, ``Inspection of Reactor Coolant 
Pressure Boundary,'' by reducing the probability and consequences of 
a SGTR. Regulatory Guide 1.121 concludes that by determining the 
limiting safe conditions for tube wall degradation, the probability 
and consequences of a SGTR are reduced. This Regulatory Guide uses 
safety factors on loads for tube burst that are consistent with the 
requirements of Section III of the American Society of Mechanical 
Engineers (ASME) Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, Westinghouse WCAP-17091-P, Rev. 
0 (Reference 3) and WCAP-17345, Rev. 2 (Reference 2) define a length 
of degradation-free expanded tubing that provides the necessary 
resistance to tube pullout due to the pressure induced forces, with 
applicable safety factors applied. Application of the limited hot 
and cold leg tubesheet inspection criteria will preclude 
unacceptable primary to secondary leakage during all plant 
conditions. When the TS operational leak rate limit of 75 gpd or 
about 0.052 gpm through any one SG is multiplied by the MSLB leak 
rate factor applicable to HBRSEP Unit No. 2 of 1.82 (Table 9-7 in 
WCAP-17091-P (Reference 3) the maximum primary to secondary accident 
induced leak rate is less than 0.095 gpm and is bounded by the value 
of 0.11 gpm through the faulted SG used in the MSLB accident 
analyses.
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the license's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Manager--Senior Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Acting Branch Chief: Jessie F. Quichocho.

Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin

    Date of amendment request: July 30, 2012.
    Description of amendment request: The proposed amendment would 
modify KPS Technical Specifications requirements regarding steam 
generator tube inspections and reporting as described in Technical 
Specification Task Force (TSTF)-510, Revision 2, ``Revision to Steam 
Generator Program Inspection Frequencies and Tube Sample Selection.''
    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.

[[Page 63350]]

    The proposed change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG 
tube sample selection. A steam generator tube rupture (SGTR) event 
is one of the design basis accidents that are analyzed as part of a 
plant's licensing basis. The proposed SG tube inspection frequency 
and sample selection criteria will continue to ensure that the SG 
tubes are inspected such that the probability of a SGTR is not 
increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design basis accident analysis. The 
proposed change will not cause the consequences of a SGTR to exceed 
those assumptions.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the Steam Generator Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
proposed change does not affect the design of the SGs or their 
method of operation. In addition, the proposed change does not 
impact any other plant system or component.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change will continue to require monitoring of the physical 
condition of the SG tubes such that there will not be a reduction in 
the margin of safety compared to the current requirements.
    Therefore, it is concluded that 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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 
120 Tredegar Street, Richmond, VA 23219.
    NRC Acting Branch Chief: Istvan Frankl.

Entergy Nuclear Operations, Inc., Docket No 50-286, Indian Point 
Nuclear Generating Unit 3, Westchester County, New York

    Date of amendment request: August 14, 2012.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.5.4, ``Refueling Water Storage Tank,'' 
such that the non-seismically qualified piping of the spent fuel pool 
purification system may be connected to the refueling water storage 
tank (RWST) seismic piping by manual operation of a RWST seismically 
qualified boundary valve under administrative controls for a limited 
period of time (i.e., 14 days per fuel cycle for filtration for removal 
of suspended solids from the RWST water). This change will only be 
applicable until Refueling Outage 18 (Spring 2015) ends.
    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 use of the SFP [spent fuel pool] Purification Loop to re-
circulate the RWST does not involve any changes or create any new 
interfaces with the reactor coolant system or main steam system 
piping. Therefore, the connection of the SFP Purification Loop to 
the RWST would not affect the probability of these accidents 
occurring. The SFP Purification Loop is not credited for safe 
shutdown of the plant or accident mitigation. Administrative 
controls ensure that the SFP Purification Loop can be isolated as 
necessary in sufficient time to assure that the RWST volume will be 
adequate to perform the safety function as designed. Since the RWST 
will continue to perform its safety function and overall system 
performance is not affected, the consequences of the accident are 
not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The design of the RWST and the SFP Purification Loop to allow 
recirculation and purification has not been altered. Procedures for 
the operation of the plant have not been revised to create the 
possibility of a new or different type of accident. Contingent upon 
manual operator action, a SFP Purification Loop line break will not 
result in a loss of the RWST safety function. Similarly, an active 
or passive failure in the SFP Purification Loop will not be 
significantly different whether aligned to the SFP or the RWST.
    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 SFP Purification Loop is not credited for safe shutdown of 
the plant or accident mitigation. Adequate RWST volume will be 
maximized prior to purification and timely operator action can be 
taken to isolate the non seismic system from the RWST to assure it 
can perform its function. This will result in no significant 
reduction in the margin of safety.
    Therefore, the proposed change does not significantly reduce the 
margin of safety.

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

Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit 3, Westchester County, New York

    Date of amendment request: August 14, 2012.
    Description of amendment request: The amendment changes the 
licensing basis regarding the emergency diesel generator (EDG) fuel oil 
storage requirements. The current licensing basis, which requires 
sufficient EDG fuel oil to operate two EDGs at minimum safeguards for 
seven days, will be revised to provide sufficient EDG fuel oil to 
operate three EDGs at modified rated capacity for seven days. The 
Conditions, Required Actions, and Completion Times of Technical 
Specification (TS) 3.8.3, ``Diesel Fuel Oil, and Starting Air,'' will 
be modified

[[Page 63351]]

to be more consistent with the Nuclear Regulatory Commission's improved 
Standard Technical Specifications. Finally, TS 3.8.3 will be modified 
to relocate specific numerical values for EDG fuel oil storage 
requirements from the TSs to the TS Bases in accordance with TS Task 
Force (TSTF) 501 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.
    The proposed change relocates the volume of diesel fuel oil 
required to support 7 day operation of the onsite diesel generators 
to licensee control, revises the action statement to reflect the 
volume equivalent to a 6 day supply and locates the volume in the TS 
Bases under licensee control, consolidates surveillance requirements 
and recalculates the fuel oil volume required for the EDG. The 
revised specific volume of fuel oil equivalent to a 7 and 6 day 
supply is calculated consistent with the NRC approved methodology 
described in Regulatory Guide 1.137, Revision 1, ``Fuel Oil Systems 
for Standby Diesel Generators'' and ANSI N195-1976, ``Fuel Oil 
Systems for Standby Diesel Generators.'' Because the requirement to 
maintain a 7 day supply of diesel fuel oil is not changed and is 
revised to be more consistent with the assumptions in the accident 
analyses, the consolidated surveillances are more conservative, and 
the actions taken when the volume of fuel oil is less than a 6 day 
supply have not changed, neither the probability or the consequences 
of any accident previously evaluated will be affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The change does not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The change 
does not alter assumptions made in the safety analysis but ensures 
that the diesel generator operates as assumed in the accident 
analysis. The proposed change is consistent with the safety analysis 
assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or 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 relocates the volume of diesel fuel oil 
required to support 7 day operation of the onsite diesel generators, 
revises the action statement to reflect the volume equivalent to a 6 
day supply, locates the volume in the TS Bases under licensee 
control, consolidates surveillance requirements and recalculates the 
fuel oil volume required for the EDG. Although the bases for the 
existing limits on diesel fuel oil are changed, no change is made to 
the accident analysis assumptions and no margin of safety is reduced 
as part of this change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

South Carolina Electric and Gas Docket Nos.: 52-027 and 52-028, Virgil 
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County, 
South Carolina

    Date of amendment request: August 29, 2012.
    Description of amendment request: The proposed change would amend 
Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear 
Station (VCSNS) Units 2 and 3, respectively, by adding four non-Class 
1E containment electrical penetration assemblies (EPAs). Containment 
EPAs are a passive extension of containment which provide the passage 
of the electric conductors through a single aperture in the nuclear 
containment structure, while providing a pressure barrier between the 
inside and the outside of the containment structure.
    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 additional containment EPAs are a passive extension of 
containment and provide a pathway for communication [sic, passage] 
of non-Class 1E electrical signals [sic, conductors] between the 
Auxiliary Building and Containment. The proposed containment 
electrical penetration assemblies are similar in form, fit and 
function to the current non-Class 1E containment electrical 
penetration assemblies. The maximum allowable leakage rate allowed 
by Technical Specifications is also unchanged. The new EPAs will 
meet the same design function as current EPAs; therefore, the 
additional penetrations 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 containment electrical penetration assemblies are 
similar in form, fit, and function to the current non-Class 1E 
containment electrical penetration assemblies. The new EPAs will 
meet the same design function as current EPAs. Because the new EPAs 
are virtually identical in design and function to the current EPAs, 
no new type of failure modes exist.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed containment electrical penetration assemblies are 
similar in form, fit and function to the current non-Class 1E 
containment electrical penetration assemblies. The additional 
containment electrical penetration assemblies are an engineered 
passive extension of containment, and, therefore, do not affect 
containment or its ability to perform its design function. The 
addition of the new EPAs does not exceed or alter a design basis or 
safety limit and, therefore, does not significantly reduce the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Mark E. Tonacci.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Units 1 and 2, Surry County, Virginia

    Date of amendment request: July 31, 2012.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications requirements regarding steam 
generator tube inspections and reporting as described in TSTF-510, 
Revision 2, ``Revision to Steam Generator Program Inspection 
Frequencies and Tube Sample Selection.''

[[Page 63352]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG 
tube sample selection. A steam generator tube rupture (SGTR) event 
is one of the design basis accidents that are analyzed as part of a 
plant's licensing basis. The proposed SG tube inspection frequency 
and sample selection criteria will continue to ensure that the SG 
tubes are inspected such that the probability of a SGTR is not 
increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design basis accident analysis. The 
proposed change will not cause the consequences of a SGTR to exceed 
those assumptions. The proposed change to reporting requirements and 
clarifications of the existing requirements have no affect on the 
probability or consequences of SGTR.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the Steam Generator Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
proposed change does not affect the design of the SGs or their 
method of operation. In addition, the proposed change does not 
impact any other plant system or component.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a SG 
is maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change will continue to require monitoring of the physical 
condition of the SG tubes such that there will not be a reduction in 
the margin of safety compared to the current requirements.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety. Based on the 
above, Dominion concludes that the proposed change presents no 
significant hazards consideration under the standards set forth in 
10 CFR 50.92(c), and, accordingly, a finding of ``no significant 
hazards consideration'' is justified. The NRC staff has reviewed the 
licensee's analysis and, based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied.
    Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
    NRC Branch Chief: Robert J. Pascarelli.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined 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.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, 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 Public Document Room (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's Reference staff at 1-800-397-4209, 301-415-4737 or by email to 
[email protected].

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

    Date of amendment request: November 17, 2011.
    Description of amendment request: The proposed amendment would add 
Optimized ZIRLO\TM\ as an allowable fuel rod cladding material and add 
the Westinghouse topical report on Optimized ZIRLO\TM\ to the Millstone 
Power Station, Unit 3 Technical Specifications. In addition, a 
typographical error would be corrected.
    Date of issuance: September 24, 2012.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days. Amendment No.: 253.
    Renewed Facility Operating License No. NPF-69: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: May 15, 2012 (77 FR 
28629).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 24, 2012.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 23, 2012.
    Description of amendment request: The proposed amendment would 
conform the Millstone Power Station Unit 3 (MPS3) licenses to reflect a 
name change for Central Vermont Public Service Corporation (CVPS) 
resulting

[[Page 63353]]

from a subsequent restructuring in which CVPS will be consolidated with 
Gaz M[eacute]tro's other electric utility subsidiary in Vermont, Green 
Mountain Power Corporation.
    Date of issuance: October 3, 2012.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days. Amendment No.: 254.
    Renewed Facility Operating License No. NPF-49: Amendment revised 
the License.
    Date of initial notice in Federal Register: July 20, 2012 (77 FR 
42768).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 21, 2012.
    No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: September 26, 2011, as 
supplemented April 16, 2012.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) to incorporate aspects of NUREG-1431, 
``Standard Technical Specifications--Westinghouse [Electric Company] 
Plants,'' STS 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,'' 
Condition E, regarding Diesel Generator starting air receiver pressure 
limits.
    Date of issuance: September 28, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--267 and Unit 2--247.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: May 29, 2012 (77 FR 
31659).
    The supplement dated April 16, 2012, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 28, 2012.
    No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: December 5, 2011.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.3.1, Table 3.3.1-1, ``Reactor Trip System 
Instrumentation,'' Function 16(e) to replace the phrase ``Turbine 
Impulse Pressure'' with ``Turbine Inlet Pressure.''
    Date of issuance: October 1, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--268 and Unit 2--248.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses and the technical specifications.
    Date of initial notice in Federal Register: August 9, 2012 (77 FR 
47677).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 2012.
    No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket No. 50-269, Oconee Nuclear Station, 
Unit 1, Oconee County, South Carolina

    Date of application of amendments: April 3, 2012
    Brief description of amendments: The amendment revised the 
Technical Specifications related to the integrated leak rate test of 
the reactor containment building.
    Date of Issuance: October 1, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 381.
    Renewed Facility Operating License No. DPR-38: Amendment revised 
the license and the technical specifications.
    Date of initial notice in Federal Register: July 10, 2012, 77 FR 
40651.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 2012.
    No significant hazards consideration comments received: No.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana 
Parish, Louisiana

    Date of amendment request: September 12, 2011, as supplemented by 
letters dated October 13, 2011, March 22, 2012, and April 3, 2012.
    Brief description of amendment: The amendment approved a change to 
the site Emergency Plan to relocate the existing backup emergency 
operations facility for RBS from its current location at the Entergy 
Operations-Baton Rouge Division Office, located at 1509 Government 
Street in Baton Rouge, Louisiana, approximately 23 miles southeast of 
RBS, to the Entergy Customer Service Center, located at 5564 Essen Lane 
in Baton Rouge, Louisiana, approximately 28 miles southeast of RBS.
    Date of issuance: September 24, 2012.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 175.
    Facility Operating License No. NPF-47: The amendment revised the 
RBS Emergency Plan.
    Date of initial notice in Federal Register: December 27, 2011 (76 
FR 80975). The supplemental letters dated October 13, 2011, March 22, 
2012, and April 3, 2012, 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 September 24, 2012.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 16, 2011, as supplemented 
by letters dated March 30, June 13, August 1, August 16, and September 
14, 2012.
    Brief description of amendment: The amendment revises the James A. 
FitzPatrick's (JAF's) current licensing basis, in the Updated Final 
Safety Analysis Report, to support installation of new reserve station 
service transformers (RSST) with on-load tap changers (OLTC). The new 
RSSTs with OLTCs will compensate for the wider range of offsite power 
voltage variations so that acceptable voltages at the safety-related 
equipment will be better maintained. The new RSSTs provided with OLTCs 
would facilitate operations in the automatic mode.
    The OLTCs are sub-components of two new RSSTs that will be 
installed at JAF during the refueling outage scheduled for September 
2012.
    Date of issuance: September 26, 2012.
    Effective date: As of the date of issuance, and shall be 
implemented

[[Page 63354]]

within 90 days The implementation of the amendment shall include 
revision of the Updated Final Safety Analysis Report as described in 
the licensee's application for this amendment.
    Amendment No.: 302.
    Renewed Facility Operating License No. DPR-59: The amendment 
revised the License and the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: November 15, 2011 (76 
FR 70768).
    The supplements dated March 30, June 13, August 1, August 16, and 
September 14, 2012 provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the NRC 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 September 26, 2012.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station (BVPS), Units 1 and 2, Beaver County, 
Pennsylvania; Docket No. 50-346, Davis-Besse Nuclear Power Station 
(DBNPS), Unit 1, Ottawa County, Ohio; Docket No. 50-440, Perry Nuclear 
Power Plant (PNPP), Unit 1, Lake County, Ohio

    Date of application for amendments: September 20, 2011, as 
supplemented by letter dated June 21, 2012.
    Brief description of amendments: The proposed amendments would 
change the licenses of BVPS, Units 1 and 2, DBNPS, and PNPP to reflect 
the name change of an owner licensee from ``FirstEnergy Nuclear 
Generation Corp.'' to ``FirstEnergy Nuclear Generation, LLC.'' The 
proposed amendment is administrative in nature. The proposed amendment 
will also correct errors regarding the name of FirstEnergy Nuclear 
Generation Corp in the DBNPS and PNNP Facility Operating Licenses.
    Date of issuance: October 2, 2012.
    Effective date: At date of issuance.
    Amendment Nos.: No. 290 to Facility Operating License No. DPR-66 
and No. 177 to Facility Operating License No. NPF-73 for BVPS, Units 1 
and 2, respectively, and No. 286 to Facility Operating License No. NPF-
3 for DBNPS, and No. 161 to Facility Operating License No. NPF-58 for 
PNPP.
    Facility Operating License Nos. DPR-66, NPF-73, NPF-3, and NPF-58: 
The amendments revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: October 18, 2011 (76 FR 
64391).
    The supplemental letter dated June 21, 2012, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed finding of no significant hazards 
consideration determination published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 2, 2012.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-389, St. Lucie 
Plant, Unit 2, St. Lucie County, Florida.

    Date of application for amendment: February 25, 2011, as 
supplemented by the letters dated November 4 and December 8, 2011, and 
April 30 and May 4 and 7, 2012.
    Brief description of amendment: This amendment raises the maximum 
fuel enrichment for fresh fuel storage from a maximum of 4.5 weight 
percent uranium-235 to a maximum lattice averaged value of 4.6 weight 
percent uranium-235. The Technical Specification changes associated 
with fuel stored in the spent fuel pool include increasing the maximum 
initial enrichment from 4.5 weight percent uranium-235 to a maximum 
planar average initial enrichment of 4.6 weight percent uranium-235, 
credit for empty storage locations, credit for use of 
METAMICTM inserts, credit for installation of full-length 
full-strength, five-fingered control element assemblies, and definition 
of three special configurations referred to in the nuclear criticality 
safety analysis as inspection and maintenance configurations.
    Date of issuance: September 18, 2012.
    Effective date: This license amendment is effective as of its date 
of issuance and shall be implemented within 60 days.
    Amendment No: 162.
    Renewed Facility Operating License No. NPF-16: Amendment revised 
the Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: September 1, 2011 (76 
FR 54503).
    The supplemental letters provided additional information that 
clarified the application and did not expand the scope of the 
application as originally noticed and published in the Federal 
Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 18, 2012.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-389, St. Lucie 
Plant, Unit 2, St. Lucie County, Florida.

    Date of application for amendment: February 25, 2011 as 
supplemented by letters dated February 25, May 24, July 22, August 18 
(three letters), August 25 (three letters), August 29, September 2, 
September 8 (two letters), September 22, October 5, October 10, October 
12 (two letters), October 31, November 2, November 3, November 4, 
November 7, November 14 (three letters), November 23 (three letters), 
December 8, December 14, December 20, December 27, December 29, 2011, 
January 14, 2012 (two letters), January 18 (two letters), January 21 
(two letters), February 29, March 6 (two letters), March 8, March 15, 
March 16, March 17 (two letters), March 25, March 31 (two letters), 
April 5 (two letters), April 6, April 10, April 19 (seven letters), 
April 30, May 4, May 7, May 18, and July 23, 2012.
    Brief description of amendments: The proposed amendments would 
increase the licensed core power level for St. Lucie Unit 2 from 2070 
megawatts thermal (MWt) to 3020 MWt. This represents a net increase in 
the core thermal power of approximately 11.85 percent, including a 10-
percent power uprate and a 1.7 percent measurement uncertainty 
recapture, over the current licensed thermal power level and is defined 
as an extended power uprate. The proposed amendments would change the 
renewed facility operating license and the technical specifications 
(TSs) to support operation at the increased core thermal power level, 
including changes to the maximum licensed reactor core thermal power, 
reactor core safety limits, and reactor protection system and 
engineered safety feature actuation system limiting safety system 
settings. Additional TS changes include reactor coolant system heatup 
and cooldown limitations, accumulator and refueling water storage tank 
boron concentrations, main steam safety valve lift settings, emergency 
diesel generator fuel storage and core operating limits report 
references. A complete list of the proposed TS changes and the 
licensee's basis for change can be found in Attachment 1 of the 
licensee's application (ADAMS Accession No. ML110730116).
    Date of issuance: September 24, 2012.

[[Page 63355]]

    Effective date: This license amendment is effective as of its date 
of issuance and shall be implemented within 60 days.
    Amendment No.: 163.
    Renewed Facility Operating License No. NPF-16: Amendment revised 
the Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: September 1, 2011 (76 
FR 54503).
    The supplemental letters provided additional information that 
clarified the application and did not expand the scope of the 
application as originally noticed and published in the Federal 
Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 24, 2012.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 16, 2011, as supplemented by 
letters dated May 2, May 24, and September 17, 2012.
    Brief description of amendment: The amendment revised the Cooper 
Nuclear Station Technical Specifications (TS) and Operating License to 
implement a 24-month fuel cycle and adopt TS Task Force (TSTF) Traveler 
TSTF-493, Revision 4, ``Clarify Application of Setpoint Methodology for 
LSSS [Limiting Safety System Settings] Functions,'' Option A. 
Specifically, the amendment revised certain TS Surveillance Requirement 
frequencies that are specified as ``18 months'' by changing them to 
``24 months'' in accordance with the guidance provided in NRC Generic 
Letter 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate a 24-Month Fuel Cycle.''
    Date of issuance: September 28, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 242.
    Renewed Facility Operating License No. DPR-46: Amendment revised 
the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 6, 2012 (77 FR 
13371). The supplemental letters dated May 2, May 24, and September 17, 
2012, 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 September 28, 2012.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 1, 2012, as supplemented by 
letters dated June 27, 2012, and July 26, 2012.
    Brief description of amendments: The amendment revises existing TS 
3.3.5.1, on a one-time basis only, by adding a note to TS Table 
3.3.5.1-1, Function 1d, Modes 4 and 5. This one-time license amendment 
enables DAEC to re-coat the internal surface of the Suppression Chamber 
during Refueling Outage 23.
    Date of issuance: September 27, 2012.
    Effective date: This license amendment is effective as of the date 
of issuance and shall be implemented within 30 days from date of 
issuance.
    Amendment No.: 283.
    Renewed Facility Operating License No. DPR-49: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: July 10, 2012 (77 FR 
40654).
    The licensee's June 27, 2012, and July 26, 2012, supplemental 
letters contained clarifying information, did not change the scope of 
the original amendment request, did not change the NRC staff's initial 
proposed finding of 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 September 27, 2012.
    No significant hazards consideration comments received: Yes.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendment: June 17, 2011, supplemented by 
letters dated July 27, 2011, November 14, 2011, March 23, April 26, May 
15, May 24, and June 26, 2012 (TS-SQN-2011-07).
    Brief description of amendment: The amendments revised the 
licensing basis and the Technical Specifications (TSs) to permit the 
use of a more robust AREVA Advanced W17 high thermal performance fuel. 
This new fuel has been selected to address fuel assembly distortion and 
its resultant fuel-handling issues. The AREVA Advanced W17 HTP fuel 
assembly design consists of standard uranium dioxide fuel pellets with 
gadolinium oxide burnable poison and M5TM cladding. The new 
fuel design ensures mechanical compatibility with the existing fuel, 
reactor core, control rods, steam supply system, and fuel-handling 
system. The transition from the existing fuel (AREVA Mark-BW) to new 
fuel is planned to occur over two refueling cycles for each unit.
    Date of issuance: September 26, 2012.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from fall 2013 refueling outage (RFO) for Unit 1, and 
prior to startup from fall 2012 RFO for Unit 2.
    Amendment Nos.: 331 and 324.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revised the License and TSs.
    Date of initial notice in Federal Register: August 23, 2011 (76 FR 
52703). The supplement letters dated July 27, 2011, November 14, 2011, 
March 23, April 26, May 15, May 24, and June 26, 2012, 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 amendment is contained 
in a Safety Evaluation dated September 26, 2012.
    No significant hazards consideration comments received: No.

    For The Nuclear Regulatory Commission.

    Dated at Rockville, Maryland, this 5th day of October 2012.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2012-25240 Filed 10-15-12; 8:45 am]
BILLING CODE 7590-01-P