[Federal Register Volume 78, Number 73 (Tuesday, April 16, 2013)]
[Notices]
[Pages 22563-22576]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-08756]


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NUCLEAR REGULATORY COMMISSION

[NRC-2013-0069]


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 (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 March 21 to April 3, 2013. The last biweekly 
notice was published on April 2, 2013 (78 FR 19746).

ADDRESSES: You may access information and comment submissions related 
to this document, which the NRC possesses and is publicly-available, by 
searching on http://www.regulations.gov under Docket ID NRC-2013-0069. 
You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket NRC-2013-0069. 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-2013-0069 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 http://www.regulations.gov and search for Docket ID NRC-2013-0069.
     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, 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-2013-0069 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 in comment submissions that you do not want to be publicly 
disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into 
ADAMS, and the NRC does not 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 in their comment submissions 
that they do not want to be publicly disclosed. Your request should 
state that the NRC will not edit comment submissions to remove such 
information before making the comment submissions available to the 
public or entering the comment submissions into ADAMS.

[[Page 22564]]

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 Sec.  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 ``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 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 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC'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 [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

[[Page 22565]]

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's 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. 
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 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].

Detroit Edison, Docket No. 50-341, Fermi 2, Monroe County, Michigan

    Date of amendment request: January 11, 2013.
    Description of amendment request: The proposed amendment would 
revise Fermi 2 Technical Specifications (TS) to

[[Page 22566]]

incorporate the NRC-approved TSTF-423, Revision 1. The proposed 
amendment would modify TS to risk-inform requirements regarding 
selected Required Action end states by incorporating the boiling water 
reactor (BWR) owner's group (BWROG) approved Topical Report NEDC-32988-
A, Revision 2, ``Technical Justification to Support Risk-Informed 
Modification to Selected Required Action End States for BWR Plants.'' 
Additionally, the proposed amendment would modify the TS Required 
Actions with a Note prohibiting the use of limiting condition for 
operation (LCO) 3.0.4.a when entering the preferred end state (Mode 3) 
on startup.
    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 change to certain required end 
states when the TS Completion Times for remaining in power operation 
will be exceeded. Most of the requested technical specification (TS) 
changes are to permit an end state of hot shutdown (Mode 3) rather 
than an end state of cold shutdown (Mode 4) contained in the current 
TS. The request was limited to: (1) Those end states where entry 
into the shutdown mode is for a short interval, (2) entry is 
initiated by inoperability of a single train of equipment or a 
restriction on a plant operational parameter, unless otherwise 
stated in the applicable TS, and (3) the primary purpose is to 
correct the initiating condition and return to power operation as 
soon as is practical. Risk insights from both the qualitative and 
quantitative risk assessments were used in specific TS assessments. 
Such assessments are documented in Section 6 of topical report NEDC-
32988-A, Revision 2, ``Technical Justification to Support Risk 
Informed Modification to Selected Required Action End States for BWR 
Plants.'' They provide an integrated discussion of deterministic and 
probabilistic issues, focusing on specific TSs, which are used to 
support the proposed TS end state and associated restrictions. The 
NRC staff finds that the risk insights support the conclusions of 
the specific TS assessments. Therefore, the probability of an 
accident previously evaluated is not significantly increased, if at 
all. The consequences of an accident after adopting TSTF-423 are no 
different than the consequences of an accident prior to adopting 
TSTF-423. 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, 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 does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
If risk is assessed and managed, allowing a change to certain 
required end states when the TS Completion Times for remaining in 
power operation are exceeded (i.e., entry into hot shutdown rather 
than cold shutdown to repair equipment) 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 and the commitment by the licensee to adhere to the guidance 
in TSTF-IG-05-02, ``Implementation Guidance for TSTF-423, Revision 
1, `Technical Specifications End States, NEDC-32988-A,'' will 
further minimize possible concerns.
    Thus, based on the above, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change allows, for some systems, entry into hot 
shutdown rather than cold shutdown to repair equipment, if risk is 
assessed and managed. The BWROG's risk assessment approach is 
comprehensive and follows NRC staff guidance as documented in 
Regulatory Guides (RG) 1.174 and 1.177. In addition, the analyses 
show that the criteria of the three-tiered approach for allowing TS 
changes are met. The risk impact of the proposed TS changes was 
assessed following the three-tiered approach recommended in RG 
1.177. A risk assessment was performed to justify the proposed TS 
changes. The net change to the margin of safety is insignificant.
    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: Bruce R. Masters, DTE Energy, General 
Counsel--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
    NRC Branch Chief: Robert D. Carlson.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3 (ONS1, ONS2, and ONS3), 
Oconee County, South Carolina

    Date of amendment request: October 30, 2012.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) to allow operation of a 
reverse osmosis system during normal plant operation to purify the 
water in the borated water storage tanks and the spent fuel pools.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee 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 requests NRC's approval of design features 
and controls that will be used to ensure that periodic limited 
operation of a Reverse Osmosis (RO) System during Unit operation 
does not significantly impact the Borated Water Storage Tank (BWST) 
or Spent Fuel Pool (SFP) function or other plant equipment. The 
proposed change also requests NRC to approve proposed Technical 
Specification (TS) requirements that will impose operating 
restrictions and isolation requirements on the RO System. Duke 
Energy evaluated the effect of potential failures, identified 
precautionary measures that must be taken before and during RO 
System operation, and identified specific required operator actions 
to protect affected structures, systems, and components (SSCs) 
important to safety.
    The new high energy piping and non-seismic piping being 
installed for the RO System is non-QA1 and is postulated to fail and 
cause an Auxiliary Building flood. Duke Energy determined that 
adequate time is available to isolate the flood source (BWST or SFP) 
prior to affecting SSCs important to safety.
    The existing Auxiliary Building Flood evaluation postulates a 
single break in the non-seismic piping occurring in a seismic event. 
The addition of the RO System will not increase the probability of a 
seismic event. The existing postulated source of the pipe break in 
the Auxiliary Building is due to the piping not being seismically 
designed. The new RO System piping is considered a potential source 
of a single pipe break for the same reason. The new non-seismic RO 
System piping is of similar quality as the existing non-seismic 
piping and is no more likely to fail than the existing piping. As 
such, the addition of new non-seismic piping does not significantly 
increase the probability of occurrence of an Auxiliary Building 
flood due to a single pipe break. An Auxiliary Building flood due to 
a non-seismic RO

[[Page 22567]]

System pipe break does not increase the consequences of the flood 
since the new non-seismic pipe break is bounded by the Auxiliary 
Building flood caused by existing non-seismic pipe breaks.
    Procedural controls will ensure that the boron concentration 
does not go below the TS limit as a result of water returned from 
the RO System with lower boron concentration. Thus, no adverse 
effects from decreased boron concentration will occur.
    The RO System takes suction from the top of the SFP to protect 
SFP inventory. Plant procedures will prohibit the use of the RO 
System for the Units 1 & 2 SFP during the time period directly after 
an outage that requires the Units 1 & 2 SFP level to be maintained 
higher than the TS Limiting Condition for Operation (LCO) 3.7.11 
level requirement. The higher level is required to support TS LCO 
3.10.1 requirements for Standby Shutdown Facility (SSF) Reactor 
Coolant (RC) Makeup System operability (due to the additional decay 
heat from the recently offloaded spent fuel). Plant procedures will 
also specify the siphon be broken during this time period so the SFP 
water above the RO suction point cannot be siphoned off if the RO 
piping breaks. The proposed change does not impact the fuel 
assemblies, the movement of fuel, or the movement of fuel shipping 
casks. The SFP boron concentration, level, and temperature limits 
will not be outside of required parameters due to restrictions/
requirements on the system's operation. In addition, the proposed 
new TS will require the siphon be broken during movement of 
irradiated fuel assemblies in the SFP or movement of cask over the 
SFP. Therefore, RO System operation cannot occur during these 
activities, effectively eliminating a Fuel Handling Accidents (FHA) 
from occurring while the RO System is in operation.
    The BWST is used for mitigation of Steam Generator Tube Rupture 
(SGTR), Main Steam Line Break (MSLB), and Loss of Coolant Accidents 
(LOCAs). The SGTR and MSLB are bounded by the small break (SBLOCA) 
analyses with respect to the performance requirements for the High 
Pressure Injection (HPI) System. In the normal mode of Unit 
operation, the BWST is not an accident initiator. The SFP is 
evaluated to maintain acceptable criticality margin for all abnormal 
and accident conditions including FHAs and cask drop accidents. Both 
the BWST and SFP are specified by TS requirements to have minimum 
levels/volumes and boron concentrations. The BWST also has TS 
requirements for temperature. Prior to RO System operation, 
procedures will require the minimum required initial boron 
concentration and initial level/volume to be adjusted. Additionally, 
they will require the RO System to be operated for a specified 
maximum time period before readjusting volume and boron 
concentration prior to another RO session. This ensures that the TS 
specified boron concentration and level/volume limits for both the 
SFP and the BWST are not exceeded during RO System operation. Thus, 
the design functions of the BWST and the SFP will continue to be met 
during RO System operation.
    Since the BWST and SFP will still have TS boron concentration 
and level/volume requirements and the RO System will be isolated 
prior to increasing radiation levels preventing access to the 
isolation valve, the mitigation of a LOCA or FHA does not result in 
an increase in dose consequence. Since the design basis LOCA 
analysis for Oconee assumes 5 gpm back-leakage from the Reactor 
Building sump to the BWST, the Emergency Operating Procedure will 
require the RO System to be isolated from the BWST prior to switch 
over to the recirculation phase. The proposed TS will require the RO 
system to be isolated (by breaking the siphon) from the SFPs during 
fuel handling activities and will require the seismic boundary valve 
between the BWST and RO System to be OPERABLE in MODES 1, 2, 3, and 
4.
    The additional controls imposed by the proposed Technical 
Specifications (TSs) will provide additional assurance that 
isolation valves and operating restrictions credited to eliminate 
the need to analyze new release pathways introduced by the RO system 
will be in place.
    Therefore, installation and operation of the RO System during 
Unit operation and the proposed TS imposing operating restrictions 
do not significantly increase the probability or consequences of any 
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 RO System adds non-seismic piping in the Auxiliary Building. 
However, the break of a single non-seismic pipe in the Auxiliary 
Building has already been postulated as an event in the licensing 
basis. The RO System also does not create the possibility of a 
seismic event concurrent with a LOCA since a seismic event is a 
natural phenomena event. The RO System does not adversely affect the 
Reactor Coolant System pressure boundary. The suction to the RO 
System, when using the system for BWST purification, contains a 
normally closed manual seismic boundary valve so the seismic design 
criteria is met for separation of seismic/non-seismic piping 
boundaries.
    Duke Energy also evaluated potential releases of radioactive 
liquid to the environment due to RO System piping failures. Design 
features, controls imposed by the proposed TS, and procedural 
controls will preclude release of radioactive materials outside the 
Auxiliary Building by ensuring the RO System will be isolated when 
required.
    The SFP suction line is designed such that the SFP water level 
will not go below TS required levels, thus the fuel assemblies will 
have the TS required water level over them. Procedural controls will 
restrict the use of the RO System and require breaking vacuum on the 
Units 1 & 2 SFP suction line when the SSF conditions require the SFP 
level be raised to support SSF RC Makeup System operability. Thus, 
the SFP water level will not be reduced below required water levels 
for these conditions. RO System operating restrictions will prevent 
reducing the SFP boron concentration below TS limits.
    Since the BWST and SFP will still have TS boron concentration 
and level/volume requirements and the RO System will be isolated 
prior to increasing radiation levels preventing access to the 
isolation valve, the mitigation of a LOCA or FHA does not result in 
an increase in dose consequence. Since the design basis LOCA 
analysis for Oconee assumes 5 gpm back-leakage from the Reactor 
Building sump to the BWST, the Emergency Operating Procedure will 
require the RO System to be isolated from the BWST prior to switch 
over to the recirculation phase. The proposed TS will require the RO 
system to be isolated (by breaking the siphon) from the SFPs prior 
to movement of irradiated fuel assemblies in the SFP or movement of 
cask over the SFP and will require the seismic boundary valve 
between the BWST and RO System to be OPERABLE in MODES 1, 2, 3, and 
4.
    The additional controls imposed by the proposed TSs will provide 
additional assurance that isolation valves and operating 
restrictions credited to eliminate the need to analyze new release 
pathways introduced by the RO system will be in place.
    Therefore, operation of the RO System during Unit operation will 
not create the possibility of a new or different kind of accident 
from any kind of accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The RO System adds non-seismic piping in the Auxiliary Building. 
Duke Energy evaluated the impact of RO System operation on SSCs 
important to safety and determined that the proposed TS controls and 
procedural controls will ensure that TS limits for SFP and BWST 
volume, temperature, and boron concentration will continue to be met 
during RO operation. For the BWST, these controls will ensure the TS 
minimum BWST boron concentration and level are available to mitigate 
the consequences of a small break LOCA or a large break LOCA. For 
the SFP, these controls ensure the assumptions of the fuel handling 
and cask drop accident analyses are preserved. Additionally, the 
failure of non-seismic RO System piping will not significantly 
impact SSCs important to safety. Oconee's licensing basis does not 
assume a design basis event occurs simultaneously with a seismic 
event. The proposed change does not significantly impact the 
condition or performance of SSCs relied upon for accident 
mitigation. This change does not alter the existing TS allowable 
values or analytical limits. The existing operating margin between 
Unit conditions and actual Unit setpoints is not significantly 
reduced due to these changes. The assumptions and results in any 
safety analyses are not impacted. Therefore, operation of the RO 
System during Unit operation 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.

[[Page 22568]]

    Attorney for licensee: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street-EC07H, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Robert J. Pascarelli.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of amendment request: February 22, 2013.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification curves for pressure and temperature 
limits on the reactor coolant system, and limits on heatup and cooldown 
rates.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment replaces the current Oconee Nuclear 
Station (ONS) Units 1, 2, and 3 pressure/temperature (P-T) limit 
curves applicable to 33 effective full power years (EFPY) in 
Technical Specification (TS) 3.4.3 with new P-T limit curves 
applicable to 54 EFPY. The proposed changes also revise the Reactor 
Coolant System heatup and cooldown rates and allowable reactor 
coolant pump combinations of TS Tables 3.4.3-1 and 3.4.3-2. The 
pressure-temperature (P-T) limit curves in the TSs were 
conservatively generated in accordance with fracture toughness 
requirements of ASME Code Section XI, Appendix G, and the minimum 
pressure and temperature requirements as listed in Table 1 of 10 CFR 
Part 50, Appendix G. The proposed changes do not impact the 
capability of the reactor coolant pressure boundary (i.e., no change 
in operating pressure, materials, seismic loading, etc.).
    Therefore, the proposed changes do not increase the potential 
for the occurrence of a loss of coolant accident (LOCA). The changes 
do not modify the reactor coolant system pressure boundary, nor make 
any physical changes to the facility design, material, or 
construction standards. The probability of any design basis accident 
(DBA) is not affected by this change, nor are the consequences of 
any DBA affected by this change. The proposed P-T limits, heatup and 
cooldown rates and allowable operating reactor coolant pump 
combinations are not considered to be an initiator or contributor to 
any accident analysis addressed in the ONS Updated Final Safety 
Analyses Report (UFSAR).
    The proposed changes will not impact assumptions and conditions 
previously used in the radiological consequence evaluations nor 
affect the mitigation of these consequences due to an accident 
described in the UFSAR. Also, the proposed changes will not impact a 
plant system such that previously analyzed SSCs might be more likely 
to fail. The initiating conditions and assumptions for accidents 
described in the UFSAR remain as analyzed.
    Therefore, the probability or consequences of an accident 
previously evaluated is not significantly increased.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The requirements for P-T limit curves have been in place since 
the beginning of plant operation. The revised curves are based on a 
later edition to Section XI of the ASME Code that incorporates 
current industry standards for P-T curves. The revised curves are 
based on reactor vessel irradiation damage predictions using 
Regulatory Guide 1.99 methodology. No new failure modes are 
identified nor are any SSCs required to be operated outside the 
design bases.
    Therefore, the possibility of a new or different kind of 
accident from any kind of accident previously evaluated is not 
created.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed P-T curves continue to maintain the safety margins 
of 10 CFR Part 50, Appendix G, by defining the limits of operation 
which prevent non-ductile failure of the reactor pressure vessel. 
Analyses have demonstrated that the fracture toughness requirements 
are satisfied and that conservative operating restrictions are 
maintained for the purpose of low temperature overpressure 
protection. The P-T limit curves provide assurance that the RCS 
pressure boundary will behave in a ductile manner and that the 
probability of a rapidly propagating fracture is minimized.
    Therefore, this request 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, 526 South Church Street-EC07H, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Robert J. Pascarelli.

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

Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and 2, 
Ogle County, Illinois

    Date of amendment request: December 21, 2012.
    Description of amendment request: The proposed amendment would 
Revise Technical Specifications (TS) 3.3.6, ``Containment Ventilation 
Isolation Instrumentation.'' Specifically, this amendment request 
proposes to revise Footnote (b) of TS Table 3.3.6-1, ``Containment 
Ventilation Isolation Instrumentation,'' which specifies the 
``Containment Radiation--High'' trip setpoint for two containment area 
radiation monitors (i.e., 1(2) RE-AR011 and 1(2) RE-AR012). The 
proposed changes would revise the ``Containment Radiation--High'' trip 
setpoint from the current, overly conservative value (i.e., a 
submersion dose rate of less than or equal to 10 mRhr in the 
containment building), to less than or equal to 2 times the containment 
building background radiation reading at rated thermal power, which is 
consistent with NUREG-1431, ``Standard Technical Specifications, 
Westinghouse Plants.'' Upon reaching the ``Containment Radiation--
High'' setpoint, these area radiation monitors provide an isolation 
signal to the containment normal purge, mini purge and post-LOCA (Loss 
of Coolant Accident) systems' containment isolation valves.
    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 containment ventilation isolation radiation monitors serve 
two primary functions, they:
    a. act as backup to the SI [safety injection] signal to ensure 
closing of the purge valves; and
    b. are the primary means for automatically isolating containment 
in the event of a fuel handling accident in containment.
    Upon sensing a high radiation condition in containment, these 
area radiation monitors provide an isolation signal to the 
containment normal purge, mini purge and post- LOCA systems 
containment isolation valves (i.e., a containment ventilation 
isolation signal).
    The accidents that could potentially be impacted by the proposed 
change were evaluated; specifically the Loss of Coolant Accident 
(LOCA), Control Rod Ejection Accident (CREA) and Fuel Handling 
Accident (FHA) in Containment. The proposed change has no impact on 
probability of these accidents occurring as the subject containment 
radiation area monitors detect a high radiation condition resulting 
from these accidents. The radiation monitors do not initiate any 
accidents or transients. Changing the ``Containment

[[Page 22569]]

Radiation--High'' trip setpoint from ``<=10 mR/hr in the containment 
building,'' to ``<=2 times the containment building background 
radiation reading at rated thermal power'' only affects the point 
(i.e., the radiation level in containment) at which a containment 
ventilation isolation signal would be generated. The requested 
change does not involve any physical plant modifications or 
operational changes that could adversely affect system reliability 
or performance of the radiation monitors, or that could affect the 
probability of operator error.
    The requested change does not affect any postulated accident 
precursors and therefore, will not affect the probability of an 
accident previously evaluated.
    The proposed change was evaluated to determine the impact on the 
dose consequences of the LOCA, CREA, or FHA in containment. The 
evaluation assumed a containment purge was in progress at the onset 
of the subject accidents and showed that the proposed change in the 
containment radiation monitors' setpoint had no effect on the purge 
valve isolation time. With regard to the LOCA and CREA, the safety 
analysis assumes a prompt purge valve isolation time (i.e., 
approximately 60 seconds) that significantly bounds the radiation 
monitor sensing and response time, and actual valve design closure 
time (i.e., a total of approximately 7 seconds). The radiation 
monitor setpoint is not considered in the safety analysis and any 
dose contribution associated with the containment purge, due to the 
proposed change in setpoint, was shown to be unaffected; therefore, 
the proposed change has no impact on the already insignificant dose 
contribution attributed to a containment purge during an accident of 
less than one mrem.
    The dose consequences associated with the FHA in containment are 
also not impacted by the proposed change in containment radiation 
monitor setpoint. The existing dose consequences resulting from a 
FHA with moving non-RECENTLY IRRADIATED FUEL (i.e., fuel moved more 
than 48 hours after reactor shutdown) conservatively assume the 
containment purge valves remain open throughout the event; 
therefore, a change in the isolation setpoint does not impact the 
results of this analysis. With regard to movement of RECENTLY 
IRRADIATED FUEL (i.e., fuel moved less then 48 hours after reactor 
shutdown), EGC's [Exelon Generation Company] proposal deletes TS LCO 
[limiting condition for operation] 3.9.4.c.2 which allowed the 
containment purge valves to be open provided the containment 
radiation isolation system is OPERABLE. Deletion of TS LCO 3.9.4.c.2 
ensures that the containment purge valves are in the closed position 
when moving RECENTLY IRRADIATED FUEL, thus removing dependence on 
the containment radiation isolation system and associated radiation 
monitor setpoint from the FHA dose consequences.
    The four other additional TS changes associated with the 
deletion of LCO 3.9.4, Item c.2, proposed for consistency (i.e., 
deleting a NOTE regarding MODE applicability, deleting a CONDITION 
related only to LCO 3.9.4.c.2, deleting a footnote regarding MODE 
applicability; and deleting two surveillances related to LCO 
3.9.4.c.2), also have no affect on either the probability or 
consequences of an accident previously evaluated.
    Based on the above discussion, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not result in a change to the design of 
the Containment Ventilation Isolation System or the manner in which 
the system operates or provides plant protection. The containment 
radiation monitors will sense radiation levels in the same way and 
will respond in the same manner when the setpoint is exceeded. The 
change in the ``Containment Radiation--High'' setpoint does not 
create a new failure mode for the associated containment radiation 
monitors or for any other plant equipment. The deletion of LCO 
3.9.4, Item c.2, in support of the setpoint change during refueling 
operations, is more conservative than the current allowances and 
actually eliminates a potential failure mode for the assumed open 
containment ventilation isolation valves as the proposed deletion of 
LCO 3.9.4, Item c.2 would require the valves to be closed prior to 
moving RECENTLY IRRADIATED FUEL.
    The changes do not result in the creation of any new accident 
precursors, the creation of any changes to the existing accident 
scenarios, nor do they create any new or different accident 
scenarios. Subsequently, the accidents defined in the UFSAR [updated 
final safety analysis report] continue to represent the credible 
spectrum of events to be analyzed which determine safe plant 
operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The analysis methodologies used in the subject safety analyses 
are not modified as a result of the proposed TS changes to the 
``Containment Radiation--High'' trip setpoint or the deletion of LCO 
3.9.4, Item c.2, or any of the other four associated TS changes. 
Although the ``Containment Radiation--High'' trip setpoint is being 
increased, the increase in response time to a high radiation 
condition in containment, when compared to the current setpoint, is 
negligible due to the projected prompt rise in containment radiation 
level upon initiation of a LOCA. The dose consequences and resultant 
margin of safety to the regulatory acceptance limits, due to 
revising the ``Containment Radiation--High'' setpoint to <= 2 times 
the containment building background radiation reading at rated 
thermal power, was shown to be unaffected for normal at-power 
containment releases; have a negligible impact on the associated 
LOCA and CREA accident dose consequences; and have no impact on the 
FHA when moving RECENTLY IRRADIATED FUEL. Therefore, the proposed 
changes do not impact any analysis margins.
    The proposed changes do not alter the manner in which the safety 
limits, limiting safety system setpoints, or limiting conditions for 
operation are determined. The current safety analyses remain 
bounding since their conclusions are not affected by the proposed 
changes. The safety systems credited in the safety analyses will 
continue to be available to perform their mitigation functions. All 
protection signals credited as the primary or secondary accident 
mitigating functions, and all operator actions credited in the 
accident analyses remain the same. The proposed changes will not 
result in plant operation in a configuration outside the design 
basis.
    Based on the above information, the proposed change does not 
result in a significant reduction in the margin of safety.
    Based on the above evaluation, EGC concludes that the proposed 
amendments do not involve a significant hazards consideration under 
the standards set forth in 10 CFR 50.92, paragraph (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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Acting Branch Chief: Jeremy S. Bowen.

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

    Date of amendment request: January 29, 2013.
    Description of amendment request: The license amendment request 
proposes to remove completed and satisfied license conditions and to 
correct inadvertent errors and incorrect references.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendments do not change or modify the fuel, fuel 
handling processes, fuel storage racks, number of fuel assemblies

[[Page 22570]]

that may be stored in the spent fuel pool (SFP), decay heat 
generation rate, or the spent fuel pool cooling and cleanup system. 
The proposed amendments only limit crediting of burnable absorbers 
in the spent fuel pool to Integrated Fuel Burnable Absorber (IFBA) 
rods that were specifically addressed in the currently approved 
criticality analysis ([Westinghouse Commercial Atomic Power report] 
WCAP-1 7094-P, Revision 3). The removal of the phrase ``or an 
equivalent amount of another burnable absorber'' eliminates the 
possibility of crediting a burnable absorber other than IFBA for 
storage of spent fuel assemblies in the spent fuel pool without 
prior NRC's approval. The deletion of the license condition 
associated with the Boraflex Remedy is editorial as it is no longer 
applicable. The proposed amendments do not affect the ability of the 
BAST [boric acid storage tank] to perform its function or the 
ability of the CREVS [control room emergency ventilation system] to 
perform its function. These latter proposed TS [technical 
specification] changes correct inadvertent errors and are consistent 
with the stated intent of original license submittals or delete 
license conditions that are no longer applicable or that have been 
fully satisfied.
    The proposed amendments do not cause any physical change to the 
existing spent fuel storage configuration, fuel makeup, RCS [reactor 
coolant system] pressure boundary, reactor containment, or plant 
systems. The proposed amendments do not affect any precursors to any 
accident previously evaluated or do not affect any known mitigation 
equipment or strategies.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendments do not change or modify the fuel, fuel 
handling processes, fuel racks, number of fuel assemblies that may 
be stored in the pool, decay heat generation rate, or the spent fuel 
pool cooling and cleanup system. The proposed amendments do not 
result in any changes to spent fuel or to fuel storage 
configurations. The removal of the phrase ``or an equivalent amount 
of another burnable absorber'' eliminates the possibility of 
crediting a burnable absorber other than IFBA for storage of spent 
fuel assemblies in the spent fuel pool without prior NRC approval. 
The proposed amendments do not affect the ability of the BAST to 
perform its function or the ability of the CREVS to perform its 
function. These latter proposed TS changes correct inadvertent 
errors and are consistent with the stated intent of the original 
license submittals, delete license conditions that are no longer 
applicable or have been fully satisfied.
    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 amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The proposed amendments do not change or modify the fuel, fuel 
handling processes, fuel racks, number of fuel assemblies that may 
be stored in the pool, decay heat generation rate, or the spent fuel 
pool cooling and cleanup system. Therefore, the proposed amendments 
have no impact to the existing margin of safety for subcriticality 
required by 10 CFR 50.68(b)(4). The other proposed OL [operating 
license] & TS changes correct inadvertent errors and are consistent 
with the stated intent of the original license submittals or delete 
license conditions that are no longer applicable or have been fully 
satisfied.
    Therefore, the proposed amendments do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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, Managing Attorney--Nuclear, 
Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Jessie F. Quichocho.

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

    Nemaha County, Nebraska
    Date of amendment request: June 25, 2012.
    Description of amendment request: The amendment would revise the 
description of the Fuel Handling Accident (FHA) in Section XIV-6.4 of 
the Cooper Nuclear Station (CNS) Updated Safety Analysis Report (USAR). 
The revised USAR FHA description is based on changes to the Design 
Basis Accident FHA dose calculation, to reflect a 24-month fuel cycle 
source term using a Global Nuclear Fuels (GNF) 10 x 10 fuel array, 
reduce the bounding Radial Peaking Factor, and revise the total 
effective dose equivalent (TEDE) contribution to consider the shine 
contribution from external sources.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The analyses changes described by this proposed change to the 
USAR are not initiators to events, and, therefore, do not involve 
the probability of an accident. The changes to the FHA calculation 
for radiological dose following a FHA incorporate the following:

--accounts for the increase to the source term owing to the use of 
Global Nuclear Fuels (GNF) 10 x 10 fuel exposed over a 24-month fuel 
cycle,
--reduces the Radial Peaking Factor from 2.0 to 1.95, and
--uses a calculated Control Room shine contribution that is added to 
the FHA dose consequences.

    The NRC computer code RADTRAD Version 3.03 is used for the 
offsite and Control Room dose calculation. The RADTRAD code was 
approved for use with the CNS FHA alternative source term (AST) dose 
calculation in License Amendment 222.
    Because the analysis affected by the changes are not considered 
to be an initiator to any previously analyzed accident, these 
changes cannot increase the probability of any previously evaluated 
accident. Therefore, these changes do not increase the probability 
of occurrence of an accident evaluated previously in the USAR.
    The changes in FHA dose consequences to the Control Room 
occupant resulting from the 24-month cycle/GNF 10 x 10 source term 
(without crediting the offset by a reduced Radial Peaking Factor), 
results in more than a minimal increase in the consequences of an 
accident previously evaluated in the USAR, as stated in 10 CFR 
50.59(c)(2)(iii). However, the resultant dose remains well within 
the regulatory limits of 10 CFR 50.67. When the reduced Radial 
Peaking Factor is applied, the dose consequences are minor. 
Therefore, this change does not significantly increase the 
consequences of an accident previously evaluated in the USAR.
    In summary, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not involve initiators to any events in the 
USAR, nor does the activity create the possibility for any new 
accidents. Rather, this change is a result of the evaluation of the 
most limiting FHA, which can occur at CNS. The changes to the FHA 
calculation for radiological dose following a FHA incorporate the 
following:

--accounts for the increase to the source term owing to the use of 
GNF 10 x 10 fuel exposed over a 24-month fuel cycle,
--reduces the Radial Peaking Factor from 2.0 to 1.95, and
--uses a calculated Control Room shine contribution that is added to 
the FHA dose consequences.

    The RADTRAD code accommodates the use of GNF 10 x 10 fuel 
exposed over a 24-month fuel cycle in calculating the FHA dose 
consequences. The reduction in Radial Peaking Factor remains 
bounding over the

[[Page 22571]]

CNS core design. The calculated Control Room shine contribution 
replaces the previously approved qualitative assessment. The 
proposed change does not introduce any new modes of plant operation 
and does not involve physical modifications to the plant. As a 
result, no new failure modes are being introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The dose consequences are calculated in accordance with the 
regulatory guidance found in RG 1.183. The RADTRAD code was used, as 
approved for application at CNS with License Amendment 222. With the 
reduced Radial Peaking Factor applied to the GNF 10 x 10 fuel that 
has been exposed over a 24-month fuel cycle, the dose consequences 
are minor. The calculated shine contribution being added to the 
total Control Room occupant FHA dose results are less than the 
previous qualitative assessment results that are being replaced. 
Accordingly, the safety margins to the regulatory dose limits are 
preserved.
    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. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: November 13, 2012.
    Description of amendment request: The proposed amendment would 
revise Renewed Facility Operating License (RFOL) Condition C.12 to 
clarify that the programs and activities, identified in Appendix A of 
NUREG-1955 and the Updated Final Safety Analysis Report (UFSAR) 
supplement are to be completed before the period of extended operation.
    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 amendment does not significantly increase the probability of 
an accident since it does not involve a change to any plant 
equipment that initiates a plant accident. The change clarifies 
RFOLC [RFOL Condition] C.12. The license conditions deal with 
administrative controls over information contained in the Updated 
Final Safety Analysis Repo[r]t (UFSAR) supplement. The proposed 
changes are administrative and the license conditions are not an 
initiator or mitigator of any previously evaluated accidents.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
since it does not involve any physical alteration of plant equipment 
and does not change the method by which any safety-related system 
performs its function. The license conditions deal with 
administrative controls over information contained in the UFSAR 
supplement. No new or different types of equipment will be installed 
and the basic operation of installed equipment is unchanged.
    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?
    Response: No.
    The proposed amendment does not affect design codes or design 
margins. The change that clarifies RFOLC C.12 is administrative in 
nature and does not have the ability to affect analyzed safety 
margins.
    Therefore, operation of DAEC in accordance with the proposed 
amendment will not involve a significant reduction in the margin to 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. James Petro, P. O. Box 14000, Juno 
Beach, FL 33408-0420.
    NRC Branch Chief: Robert D. Carlson.

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

    Date of amendment request: December 21, 2012.
    Description of amendment request: The proposed amendment would 
modify the current DAEC Technical Specifications (TS) requirement to 
operate the Standby Gas Treatment System for 10 hours on a frequency 
specified in the Surveillance Frequency Control Program in accordance 
with TSTF-522, Revision 0, ``Revise Ventilation System Surveillance 
Requirements to Operate for 10 hours per Month.
    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 replaces an existing Surveillance 
Requirement to operate the SGT System equipped with electric heaters 
for a continuous 10 hour period with a requirement to operate the 
SGT System for 15 continuous minutes without the heaters operating. 
In addition, the electrical heater output test in the VFTP 
(Specification 5.5.7.e) is proposed to be deleted and a 
corresponding change in the charcoal filter testing (Specification 
5.5.7.c) be made to require the testing be conducted at a humidity 
of at least 95% RH, which is more stringent than the current testing 
requirement of 70% RH.
    These systems are not accident initiators and therefore, these 
changes do not involve a significant increase in the probability of 
an accident. The proposed system and filter testing changes are 
consistent with current regulatory guidance for these systems and 
will continue to assure that these systems perform their design 
function which may include mitigating accidents. Thus the change 
does not involve a significant increase in the consequences of an 
accident.
    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 change replaces an existing Surveillance 
Requirement to operate the SGT System equipped with electric heaters 
for a continuous 10 hour period with a requirement to operate the 
systems for 15 continuous minutes without the heaters operating. In 
addition, the electrical heater output test in the VFTP 
(Specification 5.5.7.e) is proposed to be deleted and a 
corresponding change in the charcoal filter testing (Specification 
5.5.7.c) be made to require the testing be conducted at a humidity 
of at least 95% RH, which is more stringent than the current testing 
requirement of 70% RH.
    The change proposed for this ventilation system does not change 
any system

[[Page 22572]]

operations or maintenance activities. Testing requirements will be 
revised and will continue to demonstrate that the Limiting 
Conditions for Operation are met and the system components are 
capable of performing their intended safety functions. The change 
does not create new failure modes or mechanisms and no new accident 
precursors are generated.
    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 proposed change replaces an existing Surveillance 
Requirement to operate the SGT System equipped with electric heaters 
for a continuous 10 hour period with a requirement to operate the 
systems for 15 continuous minutes without heaters operating. In 
addition, the electrical heater output test in the VFTP is proposed 
to be deleted and a corresponding change in the charcoal filter 
testing be made to require the testing be conducted at a humidity of 
at least 95% RH, which is more stringent than the current testing 
requirement of 70% RH.
    The proposed increase to 95% RH in the required testing of the 
charcoal filters compensates for the function of the heaters, which 
was to reduce the humidity of the incoming air to below the 
currently-specified value of 70% RH for the charcoal. The proposed 
change is consistent with regulatory guidance and continues to 
ensure that the performance of the charcoal filters is acceptable.
    Therefore, it is concluded that 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: Mr. James Petro, P.O. Box 14000, Juno Beach, 
FL 33408-0420.
    NRC Branch Chief: Robert D. Carlson.

R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: April 20, 2012.
    Description of amendment request: The proposed amendment would 
revise the TS 3.1.7 to approve the use of an alternative method, other 
than the current method of the use of movable incore detectors system, 
to monitor the position of control rod or shutdown rod, in the event of 
a malfunction of the microprocessor rod position indication (MRPI) 
system. The use of this alternative method would reduce the required 
frequency of flux mapping using the movable incore detector system to 
determine the position of the control or shutdown rod position that is 
not being indicated. This will reduce the wear on the movable incore 
detector system that is also used to complete other required TS 
surveillances.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change provides an alternative method for verifying 
rod position of one rod. The proposed change meets the intent of the 
current specification in that it ensures verification of position of 
the rod once every 8 hours. The proposed change provides only an 
alternative method of monitoring rod position and does not change 
the assumptions or results of any previously evaluated accident.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change provides only an alternative method of 
determining the position of one rod. No new accident initiators are 
introduced by the proposed alternative manner of performing rod 
position verification. The proposed change does not affect the 
reactor protection system. Hence, no new failure modes are created 
that would cause a new or different kind of accidents from any 
accident previously evaluated.
    Therefore, operation of the facility in accordance with the 
proposed amendments would not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The basis of TS 3.1.7 states that the operability of the rod 
position indicators is required to determine control rod positions 
and thereby ensure compliance with the control rod alignment and 
insertion limits. The proposed change does not alter the requirement 
to determine rod position but provides an alternative method for 
determining the position of the affected rod. As a result, the 
initial conditions of the accident analysis are preserved and the 
consequences of previously analyzed accidents are unaffected.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear 
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor, 
Baltimore, MD 21202.
    NRC Acting Branch Chief: Sean Meighan.

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: March 26, 2013.
    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 by departing from the plant-specific 
design control document Tier 2* material contained within the Updated 
Safety Analysis Report (UFSAR) by revising the structural criteria code 
for anchoring of reinforcement bar within the nuclear island walls.
    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 design functions of the nuclear island structures are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the nuclear island. 
The nuclear island structures are structurally designed to meet 
seismic Category I requirements as defined in Regulatory Guide 1.29.
    The change of the requirements for anchoring headed 
reinforcement does not have an adverse impact on the response of the 
nuclear island structures to safe shutdown earthquake ground motions 
or loads due to anticipated transients or postulated accident 
conditions. The change of the requirements for anchoring headed 
reinforcement does not impact the support, design, or operation of 
mechanical and fluid systems. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to the predicted radioactive releases due to postulated 
accident conditions. The plant response to previously evaluated 
accidents or external events is not adversely affected, nor

[[Page 22573]]

does the change described create any new accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change is to provide the requirements for anchoring 
nuclear island headed reinforcement. The proposed change does not 
change the design of the nuclear island structures except to a 
limited extent to redistribute the shear reinforcement in the walls 
of the nuclear island. The proposed change does not impact the 
support, design, or operation of mechanical or fluid systems. The 
proposed change does not result in a new failure mechanism for the 
nuclear island structures or new accident precursors. As a result, 
the design functions of the nuclear island structures and the 
seismic Category I mechanical and electrical equipment located in 
the nuclear island are not adversely affected by the proposed 
change.
    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?
    Response: No.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed change, thus, no margin of 
safety is reduced. The limited application of alternative criteria 
for headed reinforcement does not reduce the margin of safety.
    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 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: Lawrence Burkhart, Acting.

Southern Nuclear Operating Company Docket Nos.: 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County, 
Georgia

    Date of amendment request: March 20, 2013.
    Description of amendment request: The proposed change would amend 
Combined Licenses Nos.: NPF-91 and NPF-92 for Vogtle Electric 
Generating Plant (VEGP) Units 3 and 4 by departing from the plant-
specific design control document Tier 2* material contained within the 
Updated Safety Analysis Report (UFSAR) by revising the structural 
criteria code for anchoring of reinforcement bar within the nuclear 
island walls.
    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 design functions of the nuclear island structures are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the nuclear island. 
The nuclear island structures are structurally designed to meet 
seismic Category I requirements as defined in Regulatory Guide 1.29.
    The change of the requirements for anchoring headed 
reinforcement does not have an adverse impact on the response of the 
nuclear island structures to safe shutdown earthquake ground motions 
or loads due to anticipated transients or postulated accident 
conditions. The change of the requirements for anchoring headed 
reinforcement does not impact the support, design, or operation of 
mechanical and fluid systems. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to the predicted radioactive releases due to postulated 
accident conditions. The plant response to previously evaluated 
accidents or external events is not adversely affected, nor does the 
change described create any new accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change is to provide the requirements for anchoring 
nuclear island headed reinforcement. The proposed change does not 
change the design of the nuclear island structures except to a 
limited extent to redistribute the shear reinforcement in the walls 
of the nuclear island. The proposed change does not impact the 
support, design, or operation of mechanical or fluid systems. The 
proposed change does not result in a new failure mechanism for the 
nuclear island structures or new accident precursors. As a result, 
the design functions of the nuclear island structures and the 
seismic Category I mechanical and electrical equipment located in 
the nuclear island are not adversely affected by the proposed 
change.
    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?
    Response: No.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed change, thus, no margin of 
safety is reduced. The limited application of alternative criteria 
for headed reinforcement does not reduce the margin of safety.
    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 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. M. Stanford Blanton, Blach & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Acting Branch Chief: Lawrence Burkhart.

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.

[[Page 22574]]

    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. 50-336, Millstone Power 
Station, Unit 2, New London County, Connecticut

    Date of amendment request: December 17, 2012, as supplemented by 
January 31, 2013.
    Description of amendment request: The amendment revised the 
Millstone Power Station, Unit 2 (MPS2) Technical Specification (TS) 
Surveillance Requirement 4.4.3.2 to remove the requirement to perform 
the quarterly surveillance for a pressurizer power-operated relief 
valve (PORV) block valve that is being maintained closed in accordance 
with TS 3.4.3 Action a. The proposed change is consistent with the 
requirements of the Standard Technical Specification--Combustion 
Engineering Plants (NUREG-1432, Revision 4).
    Date of issuance: March 26, 2013.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 314.
    Renewed Facility Operating License No. DPR-65: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: January 22, 2012 (78 FR 
4472). The supplemental letter dated January 31, 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 the amendment is contained 
in a Safety Evaluation dated March 26, 2013.
    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, West Feliciana Parish, 
Louisiana

    Date of amendment request: December 8, 2011, as supplemented by 
letters dated April 11, May 2, and September 5, 2012, and January 9 and 
March 8, 2013.
    Brief description of amendment: The amendment revised Surveillance 
Requirement (SR) 3.3.8.1.3 (calibration of loss of power 
instrumentation) to extend the frequency of the SR from 18 to 24 
months, and revised certain Allowable Values in TS 3.3.8.1, ``Loss of 
Power Instrumentation.''
    Date of issuance: March 29, 2013.
    Effective date: As of the date of issuance and shall be implemented 
90 days from the date of issuance.
    Amendment No.: 179.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 17, 2012 (77 FR 
22811). The supplemental letters dated April 11, May 2, and September 
5, 2012, and January 9 and March 8, 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 March 29, 2013.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 6, 2012.
    Brief description of amendment: The amendment revised the technical 
specifications (TS) by adding a new Limiting Condition for Operation 
(LCO) 3.0.8 associated with the impact of inoperable snubbers. This LCO 
establishes conditions under which TS systems would remain operable 
when required snubbers are not capable of providing the related support 
function. The proposed amendment is consistent with NRC's approved 
Technical Specification Task Force (TSTF) Improved Standard Technical 
Specifications Change Traveler, TSTF-372, Revision 4, ``Addition of LCO 
3.0.8, Inoperability of Snubbers.''
    Date of issuance: March 29, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 251.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 2012 (77 
FR 70841).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 29, 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

    Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and 
2, Ogle County, Illinois
    Date of application for amendment: March 22, 2012, as supplemented 
by letter dated December 3, 2012.
    Brief description of amendment: The proposed amendment would modify 
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 25, 2013.
    Effective Date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 172 and 170.
    Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: 
The amendments revised the TS 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 25, 2013.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania

    Date of application for amendments: April 27, 2012, as supplemented 
on October 15, 2012.
    Brief description of amendments: The amendments: (1) Adopted a new 
methodology for preparation of the

[[Page 22575]]

reactor coolant system pressure-temperature (P-T) limits, (2) relocated 
the P-T limits in the Technical Specifications (TSs) to a new licensee-
controlled document, the Pressure and Temperature Limits Report (PTLR), 
and (3) modified the TSs to add references to the PTLR.
    Date of issuance: April 1, 2013.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendments Nos.: 286 and 289.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the License and TSs.
    Date of initial notice in Federal Register: July 3, 2012 (77 FR 
39525). The letter dated October 15, 2012, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the application beyond 
the scope of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 1, 2013.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 6, 2012, as 
supplemented by letter dated January 11, 2013.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3/4.6.2.3, ``Recirculation pH Control System and 
NaTB Basket Minimum Loading Requirement,'' to reduce the minimum 
loading requirement of sodium tetraborate.
    Date of issuance: April 2, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 257 and 253.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: January 25, 2013 (78 FR 
5505). The supplement dated January 11, 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 2, 2013.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of application for amendments: September 12, 2012
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) to adopt NRC-approved TS Task Force 
(TSTF) Traveler TSTF-510, Revision 2, ``Revision to Steam Generator 
Program Inspection Frequencies and Tube Sample Selection,'' using the 
consolidated line item improvement process. Specifically, the 
amendments revise TS 3.4.17, ``Steam Generator (SG) Tube Integrity,'' 
TS 5.5.7, ``Steam Generator (SG) Program,'' and TS 5.6.7, ``Steam 
Generator Tube Inspection Report,'' and include TS Bases changes that 
summarize and clarify the purpose of the TS.
    Date of issuance: March 22, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment Nos.: 320 and 304.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revise the Operating Licenses and the Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2012 (77 
FR 76080).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 22, 2013.
    No significant hazards consideration comments received: No.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit 1, Washington County, Nebraska

    Date of amendment request: March 9, 2012, as supplemented by letter 
dated October 31, 2012.
    Brief description of amendment: The amendment relocated the Fort 
Calhoun Station (FCS) Technical Specification (TS) Limiting Condition 
of Operation (LCO) 2.17, Miscellaneous Radioactive Material Sources, 
and the associated Surveillance Requirement (SR) 3.13, Radioactive 
Material Sources Surveillance, from the FCS TSs. NUREG-1432, Revision 
3, ``Standard Technical Specifications, Combustion Engineering 
Plants,'' does not contain a TS or SR for radioactive source 
surveillance. The operability and surveillance requirements for leak 
checking of miscellaneous radioactive material sources will be 
incorporated into the FCS Updated Safety Analysis Report and associated 
plant procedures.
    Date of issuance: March 29, 2013.
    Effective date: As of its date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment No.: 271.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the facility operating license and the Technical 
Specifications.
    Date of initial notice in Federal Register: November 13, 2012 (77 
FR 67684). The supplemental letter dated October 31, 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 March 29, 2013.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station, 
Unit 1, Salem County, New Jersey

    Date of application for amendment: May 8, 2012.
    Brief description of amendment: The amendment revised Salem Unit 1 
Technical Specification (TS) 6.8.4.i, ``Steam Generator (SG) Program,'' 
to permanently exclude portions of the tube below the top of the steam 
generator tubesheet from periodic steam generator tube inspections. In 
addition, this amendment also revises TS 6.9.1.10, ``Steam Generator 
Tube Inspection Report,'' to provide permanent reporting requirements 
that have been previously established on an interim basis. The 
amendment was submitted pursuant to 10 CFR 50.90, ``Application for 
amendment of license, construction permit, or early site permit.''
    Date of issuance: March 28, 2013.
    Effective date: The license amendment is effective as of the date 
of issuance and shall be implemented within 60 days.
    Amendment No.: 303.
    Renewed Facility Operating License No. DPR-70: The amendment 
revised the facility operating license and the Technical 
Specifications.
    Date of initial notice in Federal Register: January 22, 2013 (78 FR 
4474).
    The Commission' related evaluation of the amendments is contained 
in a Safety Evaluation dated March 28, 2013.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 5th day of April 2013.


[[Page 22576]]


    For the Nuclear Regulatory Commission.
John D. Monninger,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2013-08756 Filed 4-15-13; 8:45 am]
BILLING CODE 7590-01-P