Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 25310-25320 [2013-10020]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 78, Number 83 (Tuesday, April 30, 2013)]
[Notices]
[Pages 25310-25320]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-10020]


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NUCLEAR REGULATORY COMMISSION

[NRC-2013-0074]


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 April 4, 2013 to April 17, 2013. The last 
biweekly notice was published on April 16, 2013 (78 FR 22563).

ADDRESSES: Please refer to Docket ID NRC-2013-0074 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 is publicly available, using any of the following 
methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0074. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: Carol.Gallagher@nrc.gov.
     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-2013-0074 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 is publicly-available, by the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0074.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly-available documents online in the NRC 
Library at https://www.nrc.gov/reading-rm/adams.html. To begin the 
search, 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 pdr.resource@nrc.gov. 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-2013-0074 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 posts all comment submissions at https://www.regulations.gov as well as entering 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

[[Page 25311]]

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 ``Agency 
Rules of Practice and Procedure'' 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 https://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 
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's 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 hearing.docket@nrc.gov, 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 https://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 https://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

[[Page 25312]]

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 https://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 NRC guidance 
available on the NRC's public Web site at https://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's Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by email at 
MSHD.Resource@nrc.gov, 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 
https://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. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. 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 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 https://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 pdr.resource@nrc.gov.

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: February 7, 2013.
    Description of amendment request: Entergy Operations, Inc. (the 
licensee), proposes to revise RBS Technical Specification (TS) 3.8.4, 
``DC [Direct Current] Sources--Operating,'' Surveillance Requirements 
(SRs) 3.8.4.2 and 3.8.4.5. The changes to the SRs will add new 
acceptance criteria to address possible non-conservative conditions 
when the battery connection resistances are at maximum values.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are to the surveillance requirements only. 
The ability of the TS surveillance to ensure that the batteries have 
the capacity to perform their specified safety functions with regard 
to accident mitigation or meeting their licensing design basis 
requirements is not reduced/diminished.
    There are no design changes associated with this TS amendment. 
The DC power system/batteries will retain adequate independency, 
redundancy, capacity and testability to permit the functioning 
required of the engineered safety features. The batteries will each 
continue to independently

[[Page 25313]]

provide this capacity assuming a failure of a single active 
component. The proposed change will not affect accident initiators 
or precursors, or adversely alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated. The proposed change will not alter the 
ability of structures, systems and components to perform their 
intended functions to mitigate the consequences of an initiating 
event. The proposed change does not physically alter safety related 
systems nor affect the way in which safety related systems perform 
their function.
    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 involves only surveillance test acceptance 
criteria. The ability of the TS surveillance to ensure that the 
batteries have the capacity to perform their specified safety 
functions with regard to accident mitigation or meeting their 
licensing design basis requirements is not reduced/diminished.
    There are no proposed design changes, nor are there any changes 
in the method by which any safety related plant structure, system, 
or component (SSC) performs its specified safety function. The 
proposed change will not affect the normal method of plant operation 
or change any operating parameters. Equipment performance necessary 
to fulfill safety analysis missions will be unaffected. The proposed 
change will not alter any assumptions required to meet the safety 
analysis acceptance criteria. No new accident scenarios, transient 
precursors, failure mechanisms, or limiting single failures will be 
introduced because of this amendment. There will be no adverse 
effect or challenges imposed on any safety related system because of 
this amendment.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not reduce the ability of the TS 
surveillance requirements to ensure that the station batteries have 
adequate capacity to perform their engineered safety features 
functions with regard to accident mitigation and meeting their 
licensing design basis requirements. The lower battery inter-cell 
connection resistance values are more restrictive, consistent with 
design basis calculations and appropriately identified in 
maintenance procedures. The proposed changes do not physically alter 
safety related systems. There will be no effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. The applicable radiological dose consequence acceptance 
criteria will continue to be met.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
1 (ANO-1), Pope County, Arkansas

    Date of amendment request: January 28, 2013.
    Description of amendment request: The ANO-1 Technical Specification 
(TS) requirements are revised from requirements on battery cells to 
requirements on the battery. This focuses the requirements on the 
assumed safety function of the battery. The proposed amendment would 
revise TS requirements related to direct current (DC) electrical 
systems in TS Limiting Condition for Operation (LCO) 3.8.4, ``DC 
Sources--Operating,'' LCO 3.8.5, ``DC Sources--Shutdown,'' and LCO 
3.8.6, ``Battery Parameters.'' A new ``Battery Monitoring and 
Maintenance Program'' is being proposed for Section 5.5, 
``Administrative Controls--Programs and Manuals.''
    These changes are consistent with the NRC-approved Technical 
Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, ``DC 
Electrical Rewrite--Update to TSTF-360.'' The availability of this TS 
improvement was announced in the Federal Register on September 1, 2011 
(76 FR 54510), as part of the consolidated line item improvement 
process (CLIIP).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes restructure the Technical Specifications 
(TS) for the direct current (DC) electrical power system and are 
consistent with TSTF-500, Revision 2. The proposed changes modify TS 
Actions relating to battery and battery charger operability 
requirements. The DC electrical power system, including associated 
battery chargers, is not an initiator of any accident sequence 
analyzed in the Safety Analysis Report (SAR). Rather, the DC 
electrical power system supports equipment used to mitigate 
accidents. The proposed changes to restructure TS and change 
surveillances for batteries and chargers to incorporate the 
applicable updates included in TSTF-500, Revision 2, will maintain 
the same level of equipment performance required for mitigating 
accidents assumed in the SAR. Operation in accordance with the 
proposed TS would ensure that the DC electrical power system is 
capable of performing its specified safety function as described in 
the SAR. Therefore, the mitigating functions supported by the DC 
electrical power system will continue to provide the protection 
assumed by the analysis. A new licensee-controlled Battery 
Monitoring and Maintenance Program will ensure appropriate 
monitoring and maintenance that is consistent with industry 
standards. In addition, the DC electrical power system is within the 
scope of 10 CFR 50.65, ``Requirements for monitoring the 
effectiveness of maintenance at nuclear power plants,'' which will 
ensure the control of maintenance activities associated with the DC 
electrical power system.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the SAR will not be 
affected by the proposed changes. Therefore, the consequences of 
previously analyzed accidents will not increase by implementing 
these changes.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes involve restructuring the TS for the DC 
electrical power system. The DC electrical power system, including 
associated battery chargers, is not an initiator to any accident 
sequence analyzed in the SAR. Rather, the DC electrical power system 
supports equipment used to mitigate accidents. The proposed changes 
to restructure the TS and change surveillances for batteries and 
chargers to incorporate the applicable updates included in TSTF-500, 
Revision 2, will maintain the same level of equipment performance 
required for mitigating accidents assumed in the SAR. Administrative 
and mechanical controls are in place to ensure the design and 
operation of the DC systems continues to meet the plant design basis 
described in the SAR.
    Therefore, operation of the facility in accordance with this 
proposed change will 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 margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The equipment margins will be

[[Page 25314]]

maintained in accordance with the plant-specific design bases as a 
result of the proposed changes. The proposed changes will not 
adversely affect operation of plant equipment. These changes will 
not result in a change to the setpoints at which protective actions 
are initiated. Sufficient DC capacity to support operation of 
mitigation equipment is ensured. The changes associated with the new 
Battery Maintenance and Monitoring Program will ensure that the 
station batteries are maintained in a highly reliable manner. The 
equipment fed by the DC electrical sources will continue to provide 
adequate power to safety-related loads in accordance with analysis 
assumptions.
    TS changes made in accordance with TSTF-500, Revision 2, 
maintain the same level of equipment performance stated in the SAR 
and the current TSs.
    Therefore, the proposed changes do not involve a significant 
reduction 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: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
2 (ANO-2), Pope County, Arkansas

    Date of amendment request: January 28, 2013.
    Description of amendment request: The ANO-2 Technical Specification 
(TS) requirements are revised from requirements on battery cells to 
requirements on the battery. This focuses the requirements on the 
assumed safety function of the battery. The proposed amendment would 
revise the TS requirements related to direct current (DC) electrical 
systems in TS Limiting Condition for Operation (LCO) 3.8.2.3, ``DC 
Distribution--Operating,'' and LCO 3.8.2.4, ``DC Distribution--
Shutdown.'' Because ANO-2 is a custom TS plant, a new TS 3.8.3, 
``Battery Parameters,'' would be created to capture the intent of 
Standard TS (STS) LCO 3.8.6, ``Battery Parameters,'' as modified by 
TSTF-500. A new ``Battery Monitoring and Maintenance Program'' is also 
being proposed for Section 6.5, ``Administrative Controls--Programs and 
Manuals.''
    These changes are consistent with the NRC-approved Technical 
Specifications Task Force (TSTF) Traveler TSTF-500, Revision 2, ``DC 
Electrical Rewrite--Update to TSTF-360.'' The availability of this TS 
improvement was announced in the Federal Register on September 1, 2011 
(76 FR 54510), as part of the consolidated line item improvement 
process (CLIIP).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes restructure the Technical Specifications 
(TS) for the direct current (DC) electrical power system and are 
consistent with TSTF-500, Revision 2. The proposed changes modify TS 
Actions relating to battery and battery charger operability. The DC 
electrical power system, including associated battery chargers, is 
not an initiator of any accident sequence analyzed in the Safety 
Analysis Report (SAR). Rather, the DC electrical power system 
supports equipment used to mitigate accidents. The proposed changes 
to restructure TS and change surveillances for batteries and 
chargers to incorporate the applicable updates included in TSTF-500, 
Revision 2, will maintain the same level of equipment performance 
required for mitigating accidents assumed in the SAR. Operation in 
accordance with the proposed TS would ensure that the DC electrical 
power system is capable of performing its specified safety function 
as described in the SAR. Therefore, the mitigating functions 
supported by the DC electrical power system will continue to provide 
the protection assumed by the analysis. A new licensee-controlled 
Battery Monitoring and Maintenance Program will ensure appropriate 
monitoring and maintenance that is consistent with industry 
standards. In addition, the DC electrical power system is within the 
scope of 10 CFR 50.65, ``Requirements for monitoring the 
effectiveness of maintenance at nuclear power plants,'' which will 
ensure the control of maintenance activities associated with the DC 
electrical power system.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the SAR will not be 
affected by the proposed changes. Therefore, the consequences of 
previously analyzed accidents will not increase by implementing 
these changes.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes involve restructuring the TS for the DC 
electrical power system. The DC electrical power system, including 
associated battery chargers, is not an initiator to any accident 
sequence analyzed in the SAR. Rather, the DC electrical power system 
supports equipment used to mitigate accidents. The proposed changes 
to restructure the TS and change surveillances for batteries and 
chargers to incorporate the applicable updates included in TSTF-500, 
Revision 2, will maintain the same level of equipment performance 
required for mitigating accidents assumed in the SAR. Administrative 
and mechanical controls are in place to ensure the design and 
operation of the DC systems continues to meet the plant design basis 
described in the SAR.
    Therefore, operation of the facility in accordance with this 
proposed change will 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 margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The equipment margins will be maintained in 
accordance with the plant-specific design bases as a result of the 
proposed changes. The proposed changes will not adversely affect 
operation of plant equipment. These changes will not result in a 
change to the setpoints at which protective actions are initiated. 
Sufficient DC capacity to support operation of mitigation equipment 
is ensured. The changes associated with the new Battery Maintenance 
and Monitoring Program will ensure that the station batteries are 
maintained in a highly reliable manner. The equipment fed by the DC 
electrical sources will continue to provide adequate power to 
safety-related loads in accordance with analysis assumptions.
    TS changes made in accordance with TSTF-500, Revision 2, 
maintain the same level of equipment performance stated in the SAR 
and the current TSs.
    Therefore, the proposed changes do not involve a significant 
reduction 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: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit 3, Nuclear Generating Plant (CR-3), Citrus County, Florida

    Date of amendment request: March 20, 2013.
    Description of amendment request: Due to the pending corporate name

[[Page 25315]]

change for the licensee of CR-3, the licensee is requesting that an 
amendment be made to this license to reflect the change in the name of 
the licensee from Florida Power Corporation to Duke Energy Florida, 
Inc.
    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?
    The proposed amendment involves a change of the corporate name 
from Florida Power Corporation to Duke Energy Florida, Inc. The 
proposed amendment does not involve any change in the technical 
qualifications of the licensee or the plant's design, configuration, 
or operation. All Limiting Conditions for Operation, Limiting Safety 
System Settings and Safety Limits specified in the CR-3 Improved 
Technical Specifications remain unchanged. Also, the Physical 
Security Plan and related plans, the Operator Training and 
Requalification Program, the Quality Assurance Program, and the 
Emergency Plan will not be materially changed by the proposed name 
change. The corporate name change amendment will not affect the 
executive oversight provided by the Chief Nuclear Officer and his 
staff.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed amendment does not involve any change in the 
plant's design, configuration, or operation. The current plant 
design, design bases, and plant safety analysis will remain the 
same.
    The Limiting Conditions for Operations, Limiting Safety System 
Settings and Safety Limits specified in the CR-3 Improved Technical 
Specifications are not affected by the proposed corporate name 
change. As such, the plant conditions for which the design basis 
accident analysis was performed remain valid.
    The proposed amendment does not introduce a new mode of plant 
operation or new accident precursors, does not involve any physical 
alterations to plant configuration, or make changes to system set 
points that could initiate a new or different kind of accident.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed amendment does not involve a change in the plant's 
design, configuration, or operation. The proposed amendment does not 
affect either the way in which the plant structures, systems, and 
components perform their safety function or its design and licensing 
bases.
    Plant safety margins are established through Limiting Conditions 
for Operation, Limiting Safety System Settings and Safety Limits 
specified in the Technical Specifications. Because there is no 
change to the physical design of the plant, there is no change to 
any of these margins.
    Therefore, the proposed amendment does not involve a significant 
reduction in 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 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, 550 South Tryon Street, 
Charlotte NC 28202.
    NRC Branch Chief: Jessie Quichocho.

NextEra Energy Seabrook, LLC., Docket No. 50-443, Seabrook Station, 
Unit 1, Rockingham County, New Hampshire

    Date of amendment request: March 13, 2013.
    Description of amendment request: The proposed amendment will 
revise the Seabrook Technical Specifications (TSs). Specifically, the 
proposed amendment will modify the circuitry that initiates high-head 
safety injection (SI) by adding a new permissive, cold leg injection 
permissive. This permissive prevents opening of the high head SI valves 
until the reactor coolant system pressure decreases to the cold leg 
injection permissive set point.
    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, along with NRC edits in square 
brackets:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change adds an additional permissive before high 
head safety injection is initiated to assist the operators in 
mitigating the consequences of an inadvertent initiation of the 
emergency core cooling system (ECCS). This change in the ECCS 
actuation circuitry does not increase the probability of any 
accident previously evaluated because:
     There is no effect on any of the systems, structures, 
or components that are used for normal operation of the plant,
     There is no effect on any of the fission product 
barriers,
     This change will not affect the normal operating 
procedures,
    The revised circuitry will delay the initiation of high head SI 
until reactor coolant pressure is below the CLIP [cold leg injection 
permissive] setpoint; however, the proposed change does not 
significantly increase the consequences of accidents previously 
evaluated. The proposed change does not alter ECCS flow. The delayed 
opening of the high head SI valves has been evaluated for the effect 
on the consequences of the following:
     Mass and energy release for steam line break accidents,
     Steam line break--UFSAR section 15.1.5 (specifically 
hot zero-power conditions)
     Feedwater line break--UFSAR section 15.2.8
     Inadvertent operation of emergency core cooling system 
during power operation--UFSAR section 15.5.1
     Chemical and volume control system malfunction that 
increases reactor coolant inventory--UFSAR section 15.5.2
    For all of the above evaluated accidents, the analysis results 
continue to meet all the safety limits. For the inadvertent 
initiation of ECCS event, the proposed change assists the operators 
in mitigating the event by significantly extending the time for the 
pressurizer to fill. Additional evaluations of small break LOCA 
[loss-of-coolant accident], best estimate large break LOCA, long 
term cooling, LOCA forces, cold overpressure mitigation/low 
temperature over pressure protection, steam generator tube rupture, 
and LOCA mass and energy release were performed and it was concluded 
that they were not affected by this change.
    In addition evaluations were performed for the centrifugal 
charging pumps and reactor vessel internals; and for the NSSS 
[nuclear steam supply system] design transients to determine if the 
change in the timing of the high head injection would have an effect 
and it was concluded that these components and transients are not 
adversely affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change adds new components to the process 
protection racks and solid state protection system similar to the 
components and configurations that are already installed. The 
sequence of operation of equipment used to mitigate the consequences 
of an accident is changed; however, it does not add any different 
types of equipment. The proposed change is a change to the 
protection circuitry for the plant and not to the system or 
equipment used for normal operation of the plant. It does not alter 
any fluid flow paths or fission product barriers and does not change 
the method of control of any plant systems used for normal 
operations. The proposed change does not alter or prevent the 
ability of the ECCS to perform its specified function to mitigate 
the consequences of an initiating event within assumed acceptance 
limits. The evaluation of the centrifugal charging pumps, reactor 
internals, control systems and NSSS design transients confirmed that 
new failure modes were not created.

[[Page 25316]]

    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed changes will not 
relax any criteria used to establish safety limits and will not 
relax any safety system settings. The safety analysis acceptance 
criteria are not affected by this change. The proposed change will 
not result in plant operation in a configuration outside the design 
basis.
    The proposed change does involve a change in the timing of the 
mitigation of inadvertent ECCS actuation and steam line break.
    This change provides additional time for mitigating the 
inadvertent operation of emergency core cooling system during power 
operation event prior to filling the pressurizer water solid, by 
preventing the injection of high head safety injection when it is 
not required.
    This change delays the injection of high head safety injection 
on a steam line break. The delay has no effect on the steamline 
break mass and energy releases and the limiting analysis of record 
hot zero power steam line break as discussed below.
    An evaluation was performed to address the impact of the CLIP 
modification on the steamline break (SLB) mass and energy release 
stretch power uprate (SPU) analyses, the current analysis of record. 
For the steamline break mass and energy analyses, the CLIP 
modification has the potential to delay initiation of ECCS injection 
by inhibiting auto-open of the cold leg injection valves until both 
an S-signal and a CLIP signal are present. There are three parts to 
the evaluation: Part 1 addresses the licensing-basis cases for 
steamline break mass and energy release inside containment, part 2 
addresses the licensing-basis cases for steamline break mass and 
energy release outside containment, and part 3 addresses steamline 
break s smaller than those analyzed for the updated final safety 
analysis report (UFSAR) for which there may be an S-signal but no 
signal associated with the CLIP.
    Steamline break inside containment. For these breaks, two 
different break types are analyzed: double-ended ruptures and split 
breaks. All cases from the SPU analysis were reviewed with respect 
to the timing of SI flow actuation from the analysis of record and 
when SI flow delivery with CLIP occurs.
    In the SPU steamline break mass and energy release analysis for 
double-ended ruptures, the first signal is low steam pressure for 
all cases. Using the SPU analysis output results, the assumed time 
of SI flow delivery is compared to the time when SI flow delivery 
with CLIP occurs. The results are that for all of the double-ended 
ruptures, SI flow delivery with CLIP is not reached until after the 
time assumed for SI flow delivery in the SPU analysis. Although an 
increase in safety injection delay is considered nonconservative, a 
sensitivity calculation was specifically performed to evaluate the 
impact of safety injection and the results show that mass and energy 
releases are not impacted by the increased delay time for safety 
injection. These results were expected as the ECCS injection occurs 
at relatively low flow rates due to high reactor coolant system 
pressure, and boron injection occurs long after the return to power 
has been mitigated by increasing reactor coolant system temperature. 
Any delay in initiation of ECCS injection has a negligible effect on 
core cooling throughout the event and core reactivity during the 
initial return to power.
    In the SPU steamline break mass and energy release analysis for 
split breaks, the first signal is the time of the first high 
containment pressure setpoint. Using the SPU analysis output 
results, the assumed time of SI flow delivery is compared to the 
time when SI flow delivery with CLIP occurs. The results are that 
for all of the split breaks, SI flow delivery with CLIP is not 
reached until after the time assumed for SI flow delivery in the SPU 
analysis. Although an increase in safety injection delay is 
considered non-conservative, a sensitivity calculation was 
specifically performed to evaluate the impact of safety injection 
and the results show that mass and energy releases are not impacted 
by the increased delay time for Safety Injection. These results were 
expected as the ECCS injection occurs at relatively low flow rates 
due to high reactor coolant system pressure, and boron injection 
occurs long after the return to power has been mitigated by 
increasing reactor coolant system temperature. Any delay in 
initiation of ECCS injection has a negligible effect on core cooling 
throughout the event and core reactivity during the initial return 
to power.
    Steamline break outside containment. The SPU analysis for the 
steamline break mass and energy release outside containment was also 
evaluated for the CLIP modification. Each steamline break case 
actuated ECCS flow on a low-low pressurizer pressure S-signal. The 
CLIP modification requires an S-signal and a CLIP signal. The 
results show that the credited S-signal is much later than the CLIP 
signal. The results from the SPU analysis remain valid and bounding 
for the CLIP modification.
    Smaller Steamline breaks. For the condition involving an S-
signal actuation with pressurizer pressure above the CLIP setpoint, 
sensitivity cases varying the start time for ECCS injection, 
including no ECCS injection have concluded that the instantaneous 
and integrated mass and energy releases are insensitive to the 
injection start time. These results were expected as the ECCS 
injection occurs at relatively low flow rates due to high reactor 
coolant system pressure, and boron injection occurs long after the 
return to power has been mitigated by increasing reactor coolant 
system temperature. Any delay in initiation of ECCS injection has a 
negligible effect on core cooling throughout the event and core 
reactivity during the initial return to power.
    The hot zero-power steamline break event remains bounding for 
operation at the current uprate conditions. The CLIP modification 
does not impact the limiting case for hot zero-power steamline break 
results because the cold leg injection valves will be fully open 
before the as-modeled high head safety injection flow starts. In 
addition, sensitivity studies confirm that the maximum break size 
remains bounding for the hot zero-power steamline break event with 
the CLIP modification.
    The above evaluation shows that the installation of a CLIP would 
not impact the Seabrook steamline break mass and energy release 
licensing basis or the hot zero-power steam line break results.
    The feedline break (FLB) has been reanalyzed with the additional 
conservatism, with respect to the SPU FLB analysis, of assuming no 
safety injection flow. The results of the analysis show that all the 
safety limits continue to be met even with the additional 
conservatism of no safety injection assumed. The assumption that 
operator action is required to mitigate the consequences of a 
chemical and volume control malfunction is not changed by this 
modification. Before CLIP, the event was bounded by the inadvertent 
ECCS actuation event and its associated operator action. With CLIP, 
the event requires operator action to terminate charging and seal 
injection flows. As discussed above, the consequences of the other 
accidents evaluated remain bounded by the analyses of record. The 
results of analyses and evaluations supporting the proposed change 
demonstrate acceptance criteria continue to be met.
    Therefore, these 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 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. James Petro, Managing Attorney, Florida 
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Meena Khanna.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
1, Rockingham County, New Hampshire

    Date of amendment request: March 27, 2013.
    Description of amendment request: The proposed amendment will 
revise the Seabrook Technical Specifications (TSs). Specifically, the 
proposed amendment will modify TS requirements regarding steam 
generator tube inspections and reporting as described in TS Task Force 
(TSTF)-510, Revision 2, ``Revision to Steam Generator Program 
Inspection Frequencies and Tube Sample Selection,'' using the 
Consolidated Line Item Improvement Process (CLIIP). The

[[Page 25317]]

changes are consistent with Industry/TSTF Standard Technical 
Specification Change Traveler, TSTF-510. The availability of this TS 
improvement was announced in the Federal Register on October 27, 2011 
(76 FR 66763), as part of the CLIIP.
    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 is 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 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.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: James Petro, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Meena Khanna.

Virginia Electric and Power Company, Docket No. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia

    Date of amendment request: February 22, 2013.
    Description of amendment request: The proposed change will allow 
the sequence and overlap limits to be exceeded and TS 3.1.6.C Action 
entered if a failure is identified during the performance of 
Surveillance Requirement (SR) 3.1.4.2, which verifies control rod 
freedom of movement. This will align the sequence and overlap limit of 
Condition A with the control bank insertion limit Condition B. The 
control bank insertion limit of Condition B was modified with this same 
change in Amendments 179 and 160. The subsequent change to Improved 
Technical Specifications (ITS) added the Condition for sequence and 
overlap limits but failed to include the exception if a failure is 
identified during control rod freedom of movement testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    The proposed amendment would modify the North Anna Power Station 
current licensing basis by increasing the time that a single rod 
bank may be permitted to be outside of sequence and overlap limits. 
The new allowance only applies to minor sequence and overlap limit 
differences. The proposed change will result in a small increase in 
the probability that, at any given time, a control bank will be 
inserted outside of sequence and overlap limits. However, the 
probability of occurrence of previously evaluated accidents is not 
affected, since the existing TS already permit a similar deviation 
with respect to insertion limit. Only the allowed duration of the 
sequence and overlap limits' exceedance is being changed.
    The allowed misalignment is not a malfunction of equipment 
important to safety; therefore, the probability of such a 
malfunction is not increased. A single rod bank's position within 18 
steps of its sequence and overlap limits does not significantly 
increase the probability of a malfunction of a component important 
to safety. This change does not impact the requirement that the rod 
bank shall be operable (i.e., trippable); as such, it remains able 
to fulfill its safety function. Therefore, the proposed amendment 
does not involve a significant increase in the consequences of a 
previously evaluated accident.
    Therefore, neither the probability of occurrence nor the 
consequences of an accident previously evaluated is significantly 
increased.

Criterion 2--Does the change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

    The proposed amendment does not create any new allowances for 
operating the plant. Only the duration of an existing allowance is 
being lengthened, with additional restrictions being applied during 
the extended allowance. No physical changes are being made to any 
portion of the plant, so no new accident causal mechanisms are being 
introduced. The proposed change does not result in any new 
mechanisms that could initiate damage to the reactor or its 
principal safety barriers (i.e., fuel cladding, reactor coolant 
system, or primary containment).
    Therefore, the possibility for a new or different kind of 
accident from any accident previously evaluated is not created.

Criterion 3--Does this change involve a significant reduction in a 
margin of safety?

    The proposed amendment does not affect the inputs or assumptions 
of any of the design basis analyses that demonstrate the integrity 
of the fuel cladding, reactor coolant system, or containment during 
accident conditions. Operation within the proposed limits will not 
cause unacceptable core radial peaking factors that could result in 
exceeding departure from nucleate boiling (DNB) limits. Operation 
within the sequence and overlap limit differences will not result in 
shutdown margins lower than assumed in the accident analyses. 
Control and Shutdown rods will remain fully operable (i.e., 
trippable) during the duration of the proposed extended allowance.
    Therefore, it is concluded that this change does not involve a 
significant reduction in the margin of safety.


[[Page 25318]]


    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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 Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Robert Pascarelli.

Previously Published Notices 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 following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Southern California Edison, Docket No. 50-361, San Onofre Nuclear 
Generating Station, Unit 2, San Diego County, California

    Date of amendment request: April 5, 2013, as supplemented by letter 
dated April 9, 2013.
    Brief description of amendment request: The proposed amendment 
makes a temporary change to the steam generator management program and 
the license condition for maximum power. For the duration of Unit 2, 
Cycle 17, the proposed amendment would change the terms ``full range of 
normal operating conditions'' and ``normal steady state full power 
operation'' and restricts operation to 70 percent of the maximum 
authorized power level. ``Full range of normal operating conditions'' 
and ``normal steady state full power operation'' shall be based upon 
the steam generators being operated under conditions associated with 
reactor core power levels up to 70 percent Rated Thermal Power (2406.6 
megawatts thermal).
    Date of publication of individual notice in Federal Register: April 
16, 2013 (78 FR 22576).
    Expiration date of individual notice: May 16, 2013 (public 
comments) and June 17, 2013 (hearing requests).
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 online through the Agencywide Documents Access and 
Management System (ADAMS) in the NRC Library at https://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 
pdr.resource@nrc.gov.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of application for amendment: June 22, 2011, as supplemented 
by letters dated December 9, 2011, January 27, 2012, and January 30, 
2013.
    Brief description of amendment: The amendments revised Technical 
Specification (TS) 3.7.4, ``Atmospheric Dump Valves (ADVs).'' 
Specifically, the amendments revised the Limiting Condition for 
Operation for TS 3.7.4, with corresponding revisions to the TS 
Conditions, Required Actions, and Completion Times associated with one 
or more inoperable ADV lines.
    Date of issuance: April 11, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: Unit 1-191; Unit 2-191; Unit 3-191.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendment revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: October 4, 2011 (76 FR 
61394). The supplemental letters dated December 9, 2011, January 27, 
2012, and January 30, 2013, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 11, 2013.
    No significant hazards consideration comments received: No.

Carolina Power and Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: October 22, 2012.
    Brief Description of amendment: The amendment revised the Technical 
Specification (TS) surveillance requirements for addressing missed 
surveillances, and is consistent with the Nuclear Regulatory Commission 
approved Revision 6 of Technical Specification Task Force (TSTF) 
Standard TSs Change Traveler TSTF-358, ``Missed Surveillance 
Requirements.''
    Date of issuance: April 11, 2013.
    Effective date: As of date of issuance and shall be implemented 
within 90 days.

[[Page 25319]]

    Amendment No.: 141.
    Renewed Facility Operating License No. NPF-63: Amendment revised 
the TSs.
    Date of initial notice in Federal Register: November 27, 2012 (77 
FR 70839).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 11, 2013.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 23, 2012, as supplemented on 
August 3, 2012.
    Brief description of amendment: The amendment revises Technical 
Specification 3.7.4, ``Atmospheric Dump Valves (ADVs),'' Limiting 
Condition for Operation 3.7.4 to require four operable ADVs instead of 
three.
    Date of issuance: April 15, 2013.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 251.
    Facility Operating License Nos. DPR-26 and DPR-64: The amendment 
revised the License and the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 2012 (77 
FR 56880).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 2013.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station (VYNPS), 
Vernon, Vermont

    Date of amendment request: March 5, 2012, as supplemented on 
November 20, 2012, March 26, March 29, and April 5, 2013.
    Brief description of amendment: The amendment revised the VYNPS 
License Condition 3.P and 3.Q to clarify that the information in the 
updated final safety analysis report (UFSAR) supplement submitted 
pursuant to Section 54.21(d) of Title 10 of the Code of Federal 
Regulations (10 CFR), as revised during the license renewal application 
review process, and as supplemented by commitments of Appendix A of 
Supplement 2 of NUREG-1907, can be incorporated as part of the UFSAR 
and may be changed without prior NRC approval provided the requirements 
of 10 CFR 50.59 have been previously satisfied.
    Date of Issuance: April 17, 2013.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 256.
    Facility Operating License No. DPR-28: The amendment revised the 
License.
    Date of initial notice in Federal Register: April 3, 2012 (77 FR 
20074). The supplemental correspondence dated November 20, 2012, March 
26, March 29, and April 5, 2013, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 17, 2013.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

    Date of application for amendment: March 22, 2012, as supplemented 
by letter dated. December 3, 2012.
    Brief description of amendment: The amendment modifies technical 
specification (TS) requirements regarding steam generator tube 
inspections and reporting as described in Technical Specifications Task 
Force (TSTF)-510, ``Revision to Steam Generator Program Inspection 
Frequencies and Tube Sample Selection,'' with proposed variations and 
deviations.
    Date of issuance: March 21, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 172 and 172, respectively.
    Facility Operating License Nos. NPF-72 and NPF-77: The amendment 
revised the Technical Specifications and License.
    Date of initial notice in Federal Register: (77 FR 31660; May 29, 
2012).
    The December 3, 2012, supplement did not increase the scope of the 
application and did not change the NRC staff's initial proposed finding 
of no significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 21, 2013.
    No significant hazards consideration comments received: No.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
1, Rockingham County, New Hampshire

    Date of amendment request: December 20, 2012.
    Description of amendment request: The amendment revised the 
Seabrook Technical Specifications (TS) TS 6.7.6.m, ``Reactor Coolant 
Pump Flywheel Inspection Program.'' The amendment extends the reactor 
coolant pump (RCP) motor flywheel examination frequency from the 
currently approved 10-year inspection interval, to an interval not to 
exceed 20 years. The changes are consistent with Industry/Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-421, ``Revision to RCP Flywheel Inspection Program 
(WCAP-15666).'' The availability of this TS improvement was announced 
in the Federal Register on October 22, 2003, as part of the 
consolidated line item improvement process (CLIIP).
    Date of issuance: April 4, 2013.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 134.
    Facility Operating License No. NPF-86: The amendment revised the 
License and TS.
    Date of initial notice in Federal Register: January 22, 2013 (78 FR 
4473).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 4, 2013.
    No significant hazards consideration comments received: No.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
1, Rockingham County, New Hampshire

    Date of amendment request: November 17, 2011, as supplemented by 
letters dated December 3, 2012, and January 9, 2013.
    Description of amendment request: The change revised the 
applicability of the figures in the Technical Specifications for the 
reactor coolant system pressure-temperature limits and the cold 
overpressure protection setpoints. The change revised the applicability 
of the figures from 20 effective full-power years (EFPY) to 23.7 EFPY.
    Date of issuance: April 15, 2013.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 135.
    Facility Operating License No. NPF-86: The amendment revised the TS 
and the License.
    Date of initial notice in Federal Register: January 10, 2012 (77 FR 
1519). The supplements dated December 3, 2012, and January 9, 2013, 
provided additional information that clarified the application, did not 
expand

[[Page 25320]]

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 amendment is contained 
in a Safety Evaluation dated April 15, 2013.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: November 30, 2011, as supplemented by 
letters dated August 16 and December 7, 2012, and March 3, 2013.
    Brief description of amendment: The amendment revised the Technical 
Specification 3.8.1, ``AC [Alternating Current] Sources--Operating,'' 
Surveillance Requirements related to Diesel Generator test loads, 
voltage, and frequency. The changes correct non-conservative Diesel 
Generator load values that are currently under administrative controls.
    Date of issuance: April 11, 2013.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days of the date of issuance.
    Amendment No.: 204.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: June 12, 2012 (77 FR 
35078). The supplemental letters dated August 16 and December 7, 2012, 
and March 3, 2013, 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 April 11, 2013.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 22nd day of April 2013.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-10020 Filed 4-29-13; 8:45 am]
BILLING CODE 7590-01-P