[Federal Register Volume 79, Number 140 (Tuesday, July 22, 2014)]
[Notices]
[Pages 42539-42557]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2014-17257]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0169]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: 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 to 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 June 26, 2014, to July 9, 2014.

DATES: Comments must be filed by August 21, 2014. A request for a 
hearing must be filed by September 22, 2014.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0169. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: 3WFN-06-A44M, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Mable Henderson, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-3760, email: Mable.Henderson@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0169 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0169.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then

[[Page 42540]]

select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section
     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-2014-0169 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 in your comment submission. The NRC will post all comment 
submissions at http://www.regulations.gov as well as enter the comment 
submissions into ADAMS, and the NRC does not routinely edit comment 
submissions to remove identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

II. Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses and Combined Licenses and Proposed No Significant 
Hazards Consideration Determination

    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.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

    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's 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.

[[Page 42541]]

    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.

B. Electronic Submissions (E-Filing)

    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 
ten 10 days prior to the filing deadline, the participant should 
contact the Office of the Secretary by email at hearing.docket@nrc.gov, 
or by telephone at 301-415-1677, to request (1) a digital 
identification (ID) certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.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 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 NRC'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 public 
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
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

[[Page 42542]]

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 three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan
    Date of amendment request: April 23, 2014, as supplemented by 
letter dated June 19, 2014. Publicly available versions are in ADAMS 
under Accession Nos. ML14113A445 and ML14170B201, respectively.
    Description of amendment request: The proposed amendment would 
revise the technical specification (TS) surveillance requirements (SRs) 
associated with TS 3.8.4, ``DC Sources--Operating'' and TS 3.8.6, 
``Battery Cell Parameters.''
    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.
    Performing the proposed changes in battery parameter surveillance 
testing and verification is not a precursor of any accident previously 
evaluated. Furthermore, these changes will help to ensure that the 
voltage and capacity of the batteries is such that they will provide 
the power assumed in calculations of design basis accident mitigation.
    Therefore, DTE concludes that the proposed changes do not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve any modification of the plant 
or how the plant is operated; they only involve surveillance testing 
and verification activities.
    Therefore, DTE concludes that these proposed changes do not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of the 
fission product barriers to perform their design functions during and 
following an accident situation. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. The 
performance of the fuel cladding, reactor coolant, and containment 
systems will not be impacted by the proposed changes.
    The proposed Fermi 2 revisions of the SRs ensure the continued 
availability and operability of the batteries. As such, sufficient 
[direct current] capacity to support operation of mitigation equipment 
remains within the design basis.
    Therefore, DTE concludes that the proposed changes 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 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. Maters, DTE Energy, General 
Counsel--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
    NRC Branch Chief: Robert D. Carlson.
Duke Energy Progress Inc., Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina
    Date of amendment request: February 10, 2014, as supplemented by 
letter dated April 4, 2014. Publicly-available versions are in ADAMS 
under Accession Nos. ML14052A065 and ML14107A339, respectively.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.3.1 for the Reactor Protection System 
Instrumentation Turbine Trip function on Low Auto Stop Oil (ASO) 
Pressure to a Turbine Trip function on Low Electro-Hydraulic (EH) Fluid 
Oil Pressure. The amendment would revise the Allowable Value and 
Nominal Trip Setpoint and revise the TS by applying additional testing 
requirements listed in Technical Specifications Task Force Traveler 
493-A Revision 4, ``Clarify Application of Setpoint Methodology for 
Limiting Safety System Setting Functions,'' for Low EH Fluid Oil 
Pressure trip only.
    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 reflects a design change to the turbine control 
system that results in the use of an increased control oil pressure 
system, necessitating a change to the value at which a low EH fluid oil 
pressure initiates a reactor trip on turbine trip. The EH oil pressure 
is an input to the reactor trip instrumentation in response to a 
turbine trip event. The value at which the low Electro-Hydraulic fluid 
oil initiates a reactor trip is not an accident initiator. A change in 
the nominal control oil pressure does not introduce any mechanisms that 
would increase the probability of an accident previously analyzed. The 
reactor trip on turbine trip function is initiated by the same 
protective signal as used for the ASO System trip signal. There is no 
change in form or function of this signal and the probability or 
consequences of previously analyzed accidents are not impacted.
    The proposed change also adds test requirements to a TS instrument 
function related to those variables that have a significant safety 
function to ensure that instruments will function as required to 
initiate protective systems or actuate mitigating systems at the point 
assumed in the applicable setpoint calculation. Surveillance tests are 
not an initiator to any accident previously evaluated. As a result, the 
probability of any accident previously evaluated is not significantly 
increased. The systems and components required by the TSs for which 
surveillance tests are added are still required to be operable, meet 
the acceptance criteria for the surveillance requirements, and be 
capable of performing any mitigation function.

[[Page 42543]]

    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 EH fluid oil pressure decreases in response to a turbine trip. 
The value at which the low EH fluid oil initiates a reactor trip is not 
an accident initiator. The proposed TS change reflects the higher 
pressure that will be sensed after the pressure switches are relocated 
from the ASO System to the AST [Auto Stop Trip] high pressure header. 
Failure of the new switches would not result in a different outcome 
than is considered in the current design basis. Further, the change 
does not alter assumptions made in the safety analysis but ensures that 
the instruments perform as assumed in the accident analysis.
    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 change involves a parameter that initiates an anticipatory 
reactor trip following a turbine trip. The safety analyses do not 
credit this anticipatory trip for reactor core protection. The original 
pressure switch configuration and the new pressure switch configuration 
both generate the same reactor trip signal. The difference is that the 
initiation of the trip will now be adjusted to a different system of 
higher pressure. This system function of sensing and transmitting a 
reactor trip signal on turbine trip remains the same. Also, the 
proposed change adds test requirements that will assure that (1) 
technical specifications instrumentation Allowable Values will be 
limiting settings for assessing instrument channel operability and (2) 
will be conservatively determined so that evaluation of instrument 
performance history and the as left tolerance requirements of the 
calibration procedures will not have an adverse effect on equipment 
operability. The testing methods and acceptance criteria for systems, 
structures, and components, specified in applicable codes and standards 
(or alternatives approved for use by the NRC) will continue to be met 
as described in the plant licensing basis including the updated Final 
Safety Analysis Report. There is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis because 
no change is made to the accident analysis assumptions.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, 
Charlotte, NC 28202.
    NRC Acting Branch Chief: Lisa M. Regner.
Duke Energy Progress Inc., Docket No. 50-400, Shearon Harris Nuclear 
Power Plant, Unit No. 1, New Hill, North Carolina
    Date of amendment request: April 24, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14114A743.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.4.5, ``Steam Generator Tube 
Integrity,'' TS 6.8.4.I, ``Steam Generator Program,'' and TS 6.9.1.7, 
``Steam Generator Tube Inspection Report'' to address implementation 
associated with the inspections and reporting requirements as described 
in Technical Specifications Task Force (TSTF) TSTF-510-A, Revision 2, 
``Revision to Steam Generator Program Inspection Frequencies and Tube 
Sample Selection.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG tube 
sample selection. A steam generator tube rupture (SGTR) event is one of 
the design basis accidents that are analyzed as part of a plant's 
licensing basis. The proposed SG tube inspection frequency and sample 
selection criteria will continue to ensure that the SG tubes are 
inspected such that the probability of a SGTR is not increased. The 
consequences of a SGTR are bounded by the conservative assumptions in 
the design basis accident analysis. The proposed change will not cause 
the consequences of a SGTR to exceed those assumptions.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the Steam Generator Program will not 
introduce any adverse changes to the plant design basis or postulated 
accidents resulting from potential tube degradation. The proposed 
change does not affect the design of the SGs or their method of 
operation. In addition, the proposed change does not impact any other 
plant system or component.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part of 
the reactor coolant pressure boundary and, as such, are relied upon to 
maintain the primary system's pressure and inventory. As part of the 
reactor coolant pressure boundary, the SG tubes are unique in that they 
are also relied upon as a heat transfer surface between the primary and 
secondary systems such that residual heat can be removed from the 
primary system. In addition, the SG tubes also isolate the radioactive 
fission products in the primary coolant from the secondary system. In 
summary, the safety function of a SG is maintained by ensuring the 
integrity of its tubes. Steam generator tube integrity is a function of 
the design, environment, and the physical condition of the tube. The 
proposed change does not affect tube design or operating environment. 
The proposed change will continue to require monitoring of the physical 
condition of the SG tubes such that there will not be a reduction in 
the margin of safety compared to the current requirements.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 42544]]

review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, 
Charlotte, NC 28202.
    NRC Acting Branch Chief: Lisa M. Regner.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of amendment request: March 18, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14086A389.
    Description of amendment request: The amendment would adopt 
Technical Specification (TS) Task Force (TSTF) change traveler TSTF-
535, Revision 0, ``Revise Shutdown Margin [SDM] Definition to Address 
Advanced Fuel Designs,'' at Columbia Generating Station. The notice of 
availability of TSTF-535, Revision 0, was announced in the Federal 
Register on February 26, 2013 (78 FR 13100).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the definition of SDM. SDM is not an 
initiator to any accident previously evaluated. Accordingly, the 
proposed change to the definition of SDM has no effect on the 
probability of any accident previously evaluated. SDM is an assumption 
in the analysis of some previously evaluated accidents and inadequate 
SDM could lead to an increase in consequences for those accidents. 
However, the proposed change revises the SDM definition to ensure that 
the correct SDM is determined for all fuel types at all times during 
the fuel cycle. As a result, the proposed change does not adversely 
affect the consequences of any accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the definition of SDM. The change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operations. The change does not alter 
assumptions made in the safety analysis regarding SDM.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the definition of SDM. The proposed 
change does not alter the manner in which safety limits, limiting 
safety system settings or limiting conditions for operation are 
determined. The proposed change ensures that the SDM assumed in 
determining safety limits, limiting safety system settings or limiting 
conditions for operation is correct for all BWR fuel types at all times 
during the fuel cycle.
    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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of amendment request: March 24, 2014, as supplemented by 
letter dated May 8, 2014. Publicly-available versions are in ADAMS 
under Accession Nos. ML14098A400 and ML14141A538, respectively.
    Description of amendment request: The amendment would revise 
Columbia Generating Station Technical Specification (TS) Table 3.3.1.1-
1 to update Scram Discharge Volume (SDV) instrumentation nomenclature, 
add a Surveillance Requirement (SR) which was previously omitted, and 
add footnotes to an SR consistent with TS Task Force (TSTF) change 
traveler TSTF-493, Revision 4, ``Clarify Application of Setpoint 
Methodology for LSSS [Limiting Safety System Settings] Functions,'' 
Option A. The notice of availability of the models for plant-specific 
adoption of TSTF-493, Revision 4, was announced in the Federal Register 
on May 11, 2010 (75 FR 26294).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to Function 7 names are administrative in 
nature and ensure that the description of SDV Water Level--High 
instrumentation in TS matches the plant configuration. The addition of 
a missing channel check SR and TSTF-493 footnotes for the new Function 
7.b instruments makes the TS more comprehensive by ensuring the 
appropriate surveillances and footnotes are applied to this 
instrumentation.
    The replacement instruments for Function 7.b meet the high 
functional reliability standard of GDC 21 [General Design Criteria 21, 
``Protection system reliability and testability,'' of 10 CFR Part 50, 
Appendix A] and all pertinent requirements of 10 CFR 50.55a(h)(2). The 
instrumentation modification was reviewed under 10 CFR 50.59(c)(1) and 
determined to not meet any of the criteria in 10 CFR 50.59(c)(2).
    The addition of a channel check to Function 7.a and addition of 
TSTF-493 notes (d) and (e) to SR 3.3.1.1.10 for the Function 7.b 
instrumentation do not change accident frequency or consequences. TS 
requirements that govern operability or routine testing of plant 
instruments are not assumed to be initiators of any analyzed event 
because these instruments are intended to prevent, detect, or mitigate 
accidents. Additionally, these proposed changes will not increase the 
consequences of an accident previously evaluated because the proposed 
changes do not adversely impact structures, systems, or components. The 
proposed TS changes establish requirements that ensure components are 
operable when necessary for the prevention or mitigation of accidents 
or transients. Furthermore, there will be no change in the types or 
significant increase in the amounts of any effluents released offsite.

[[Page 42545]]

    In summary, 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 changes to administratively revise instrument 
descriptions, incorporate a new SR, and add footnotes to an existing SR 
do not change the parameters within which Columbia is operated.
    The proposed changes do not adversely impact the manner in which 
the SDV Water Level--High RPS [Reactor Protection System] 
instrumentation will operate under normal and abnormal operating 
conditions. The instrumentation design changes were reviewed under 10 
CFR 50.59(c)(1) and determined to not meet any of the criteria of 10 
CFR 50.59(c)(2). The proposed changes will not alter the functional 
demands on credited equipment. No alteration in the procedures which 
ensure that Columbia remains within analyzed limits are proposed and no 
change is being made to procedures relied upon to respond to an off-
normal event.
    Therefore, these proposed changes provide an equivalent level of 
safety and will 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 changes to the function descriptions in TS Table 
3.3.1.1-1 Functions 7.a and 7.b are considered administrative in 
nature, and do not impact plant safety.
    Margins of safety are established in the design of components, the 
configuration of components to meet certain performance parameters, and 
in the establishment of setpoints to initiate alarms and actions. The 
proposed changes support a planned upgrade of the SDV instrumentation 
that preserves the reliability of the RPS system. The proposed changes 
do not adversely affect the probability of failure or availability of 
the affected instrumentation. The instrumentation design changes were 
evaluated under 10 CFR 50.59(c)(1) and determined not to meet any of 
the criteria of 10 CFR 50.59(c)(2).
    The addition of a Channel Check SR to TS Table 3.3.1.1-1 Function 
7.a and the addition of TSTF-493 notes (d) and (e) to SR 3.3.1.1.10 for 
the new scram discharge instrumentation in TS Table 3.3.1.1-1 Function 
7.b are conservative changes that align the SRs for proper 
determination of operability with that of similar instrumentation.
    On this basis, is concluded that the proposed changes do not result 
in a reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts
    Date of amendment request: November 26, 2013. A publicly-available 
version is in ADAMS under Accession No. ML13346A026.
    Description of amendment request: The amendment would revise 
Technical Specification 4.3.4, ``Heavy Loads'' limitation imposed on 
maximum weight that could travel over the irradiated fuel in the spent 
fuel pool.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (CFR) Section 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 Reactor Building crane is being upgraded to meet the applicable 
single-failure-proof criteria of NUREG 0554 and NUREG 0612 for the 
modification of the existing non single-failure-proof crane. While 
loads in excess of 2,000 lbs [pounds] shall continue to be prohibited 
from travel over irradiated fuel assemblies in the spent fuel pool by 
the PNPS [Pilgrim Nuclear Power Station] Technical Specifications, a 
Multi-Purpose Canister (MPC) lid will be permitted to travel over 
irradiated fuel assemblies in a transfer cask, using a single-failure-
proof handling system as described in NUREG-0800 Section 9.1.5 
Paragraph llI.4.C, to enable the conduct of dry cask storage loading 
and unloading operations. Specifically, this will enable the MPC lid 
and its associated lifting apparatus to travel over irradiated fuel 
assemblies in a MPC. The probability of dropping this load onto an 
irradiated fuel assembly in the canister is reduced as a result of the 
reliability of the single-failure-proof handling system.
    The proposed change does not affect the consequences of any 
accidents previously evaluated in the PNPS UFSAR [Updated Final Safety 
Analysis Report]. The change involves the travel of heavy loads over 
irradiated fuel assemblies in a transfer cask using a single-failure-
proof handling system. Under these circumstances, no new load drop 
accidents are postulated and no changes to the probabilities or 
consequences of accidents previously evaluated are involved.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Section 10.3 of the PNPS UFSAR evaluates fuel storage and handling 
operations. Section 14 of the PNPS UFSAR discusses the analysis of 
design basis fuel handling accidents involving drop of an irradiated 
assembly resulting in multiple fuel rod failures and consequent release 
of radioactivity. The change involves the travel of heavy loads over 
irradiated fuel assemblies in a transfer cask using a single-failure-
proof handling system. Under these circumstances, no new or different 
load drop accidents are postulated to occur and there are no changes in 
any of the load drop accidents previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The revised Technical Specification changes do not involve a 
reduction in any margin of safety. Technical Specification 4.3.4 
currently prohibits travel of heavy loads in excess of 2,000 lbs over 
irradiated fuel assemblies in the spent fuel pool. The proposed change 
will continue to restrict travel of heavy loads in excess of 2,000 lbs 
over irradiated fuel assemblies in the spent fuel pool, with the 
exception of the MPC lid over irradiated fuel assemblies in the 
canister to enable dry cask storage operations. This exception is only 
permitted when the heavy load is handled using a single-failure-proof 
handling system. Due to the reliability of this upgraded handling 
system that complies with the guidance of NUREG-0800 Section 9.1.5 for 
a single-failure-proof handling system, a load drop accident is not 
considered a credible event. Under these circumstances, no

[[Page 42546]]

new load drop accidents are postulated and no reductions in margins of 
safety are involved.
    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: Ms. Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Benjamin G. Beasley
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont
    Date of amendment request: March 24, 2014. A publicly available 
version is in ADAMS under Accession No. ML14085A257.
    Description of amendment request: The proposed amendment would 
revise the site emergency plan for the permanently defueled condition 
to reflect changes in the on-shift staffing and Emergency Response 
Organization staffing.
    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 changes to the SEP [Site Emergency Plan] do not impact 
the function of plant structures, systems, or components (SSCs). The 
proposed changes do not affect accident initiators or precursors, nor 
does it alter design assumptions. The proposed changes do not prevent 
the ability of the on-shift staff and ERO [Emergency Response 
Organization] to perform their intended functions to mitigate the 
consequences of any accident or event that will be credible in the 
permanently defueled condition. The proposed changes only remove 
positions that will no longer be credited in the SEP in the permanently 
defueled condition.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes reduce the number of on-shift and ERO 
positions commensurate with the hazards associated with a permanently 
shutdown and defueled facility. The proposed changes do not involve 
installation of new equipment or modification of existing equipment, so 
that no new equipment failure modes are introduced. Also, the proposed 
changes do not result in a change to the way that the equipment or 
facility is operated so that no new accident initiators are created.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the level 
of radiation dose to the public. The proposed changes are associated 
with the SEP staffing and do not impact operation of the plant or its 
response to transients or accidents. The change does not affect the 
Technical Specifications. The proposed changes do not involve a change 
in the method of plant operation, and no accident analyses will be 
affected by the proposed changes. Safety analysis acceptance criteria 
are not affected by the proposed changes. The revised SEP will continue 
to provide the necessary response staff with the proposed changes.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company (EGC), 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, Unit Nos. 1 and 2, Ogle 
County, Illinois
    Date of amendment request: April 17, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14111A257.
    Description of amendment request: The proposed amendment would 
revise required action notes in the Braidwood and Byron TS 3.3.1 and TS 
3.3.2 to reflect the specific functions in TS 3.3.1 and TS 3.3.2 that 
have bypass test capability installed and the specific functions that 
do not have bypass test capability installed. The current wording is no 
longer applicable because the installation and implementation of the 
bypass test instrumentation modifications for certain functions have 
been completed.
    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 is administrative in nature as it revises 
previously approved specific TS [Technical Specifications] Required 
Actions Notes that are no longer applicable following plant 
modification installation and implementation to reflect the applicable 
RTS [Reactor Trip System] and ESFAS [Engineered Safety Feature 
Actuation System] Functions with installed bypass test capability.
    The proposed change does not impact any accident initiators, 
analyzed events, or assumed mitigation of accident or transient events 
modeled in the safety analyses. The proposed change does not alter the 
design assumptions, conditions, or configuration of the facility, nor 
does it affect the structural and functional integrity of the RTS and 
ESFAS. The proposed change does not alter or prevent the ability of any 
structures, systems, and components from performing their intended 
design function to mitigate the consequences of an initiating event 
within the applicable acceptance criteria.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.

[[Page 42547]]

    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to revise previously approved specific TS 
Required Actions Notes that are no longer applicable to specific RTS 
and ESFAS Functions with installed bypass test capability is 
administrative in nature. The proposed change does not result in a 
change to any design function or the manner in which the RTS and ESFAS 
operates to provide plant protection. The RTS and ESFAS will continue 
to have the same setpoints after the proposed change is implemented. In 
addition, this change does not install or modify any plant equipment. 
Therefore, no new failure modes are being created nor does the change 
result in the creation of any changes to the existing accident 
scenarios or do they create any new or different accident scenarios. 
The types of 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.
    No safety analyses are changed or modified as a result of the 
proposed change to revise previously approved specific TS Required 
Actions Notes that are no longer applicable to RTS and ESFAS Functions 
with installed bypass test capability. The proposed change does not 
alter the manner in which the safety limits, limiting safety system 
settings, or limiting conditions for operation are determined. Margins 
associated with the current applicable safety analyses acceptance 
criteria are unaffected. The current safety analyses remain bounding 
since their conclusions are not affected by this change and the plant 
will continue to operate in a manner consistent with the safety 
analyses. The safety systems credited in the safety analyses will 
continue to be available to perform their mitigation functions.
    Therefore, 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 Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
    Date of amendment request: October 31, 2013. A publicly-available 
version is in the ADAMS System under Accession No. ML13308A387.
    Description of amendments request: The amendments would modify the 
Technical Specification requirements regarding steam generator tube 
inspections and reporting as described in Technical Specification Task 
Force 510-A, Revision 2, ``Revision to Steam Generator Program 
Inspection Frequencies and Tube Sample Selection.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated; or
    No.
    The proposed change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG tube 
sample selection. A steam generator tube rupture event (SGTR) is one of 
the design basis accidents that are analyzed as part of a plant's 
licensing basis. The proposed SG tube inspection frequency and sample 
selection criteria will continue to ensure that the SG tubes are 
inspected such that the probability of a SGTR is not increased. The 
consequences of a SGTR are bounded by the conservative assumptions in 
the design basis accident analysis. The proposed change will not cause 
the consequences of a SGTR to exceed these assumptions.
    Therefore, operation of the facility in accordance with the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different type of accident 
from any accident previously evaluated; or
    No.
    The proposed changes to the SG Program will not introduce any 
adverse changes to the plant design basis or postulated accidents 
resulting from potential tube degradation. The proposed change does not 
affect the design of the SGs or their method of operation. In addition, 
the proposed change does not impact any other plant system or 
component.
    Therefore, the proposed amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    No.
    The SG tubes in pressurized water reactors are an integral part of 
the reactor coolant pressure boundary and, as such, are relied upon to 
maintain the primary system's pressure and inventory. As part of the 
reactor coolant pressure boundary, the SG tubes are unique in that they 
are also relied upon as a heat transfer surface between the primary and 
secondary systems such that residual heat can be removed from the 
primary system. In addition, the SG tubes also isolate the radioactive 
fission products in the primary coolant from the secondary system. In 
summary, the safety function of a SG is maintained by ensuring the 
integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change will continue to require monitoring of the physical 
condition of the SG tubes such that there will not be a reduction in 
the margin of safety compared to the current requirements.
    Therefore, 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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 
Exelon Way, Kennett Square, PA 19348.
    NRC Branch Chief: Benjamin G. Beasley

[[Page 42548]]

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
    Date of amendment request: November 13, 2013. A publicly-available 
version is in the ADAMS System under Accession No. ML13318A892.
    Description of amendments request: The amendments would modify the 
Technical Specification requirements to adopt the changes described in 
Technical Specification Task Force 426-A, Revision 5, ``Revise or Add 
Actions to Preclude Entry into LCO 3.0.3--RITSTF Initiatives 6b and 
6c.''
    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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated; or
    No.
    The proposed change provides a short Completion Time to restore an 
inoperable system for conditions under which the existing Technical 
Specifications require a plant shutdown to begin within one hour in 
accordance with Limiting Condition for Operation 3.0.3. Entering into 
Technical Specification Actions is not an initiator of any accident 
previously evaluated. As a result, the probability of an accident 
previously evaluated is not significantly increased. The consequences 
of any accident previously evaluated that may occur during the proposed 
Completion Times are no different from the consequences of the same 
accident during the existing one hour allowance. As a result, the 
consequences of any accident previously evaluated are not significantly 
increased.
    Therefore, operation of the facility in accordance with the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different type of accident 
from any accident previously evaluated; or
    No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements. The 
changes do not alter assumptions made in the safety analysis.
    Therefore, the proposed amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    No.
    The proposed change increases the time the plant may operate 
without the ability to perform an assumed safety function. The analyses 
in WCAP-16125-NP-A, ``Justification for Risk-Informed Modifications to 
Selected Technical Specifications for Conditions Leading to Exigent 
Plant Shutdown,'' Revision 2, August 2010, demonstrated that there is 
an acceptably small increase in risk due to a limited period of 
continued operation in these conditions and that this risk is balanced 
by avoiding the risks associated with a plant shutdown. As a result, 
the change to the margin of safety provided by requiring a plant 
shutdown within one hour is not significant.
    Therefore, 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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 
Exelon Way, Kennett Square, PA 19348
    NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
    Date of amendment request: January 13, 2014. A publicly-available 
version is in the ADAMS under Accession No. ML14015A138.
    Description of amendments request: The amendments would add a 
Technical Specification (TS) for the atmospheric dump valves (ADVs).
    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 addition of a new TS to address the operability of the 
ADVs does not alter the assumed initiators to any analyzed event. The 
probability of an accident previously evaluated will not be increased 
by this proposed change. This proposed change will not affect 
radiological dose consequence analyses. The radiological dose 
consequence analyses assume a certain release of radioactive material 
through the ADVs following a steam generator tube rupture (SGTR), which 
is not affected by the addition of the ADVs to the TS. The addition of 
a Surveillance Requirement for the ADVs will continue to ensure that 
the ADVs can perform their specified function. The consequences of an 
accident previously evaluated will not be increased by this proposed 
change.
    Therefore, operation of the facility in accordance with the 
proposed TS for the ADVs will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed addition of a new TS to address the operability of the 
ADVs has been evaluated to determine the effect of adding the new TS to 
the operation of the plant. This change does not involve any alteration 
in the plant configuration (no new or different type of equipment will 
be installed) or make changes in the methods governing normal plant 
operation. The change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, operation of the facility in accordance with the 
proposed addition of a new TS to address the operability of the ADVs 
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 margin of safety is related to the ability of the ADV to 
release enough steam to cool the Reactor Coolant System down and be 
isolated when required to limit the radioactive release from a SGTR. 
The inclusion of the ADVs in the TS will provide limited time for 
continued operation without both ADVs available. This ensures that the 
margin of safety is maintained by ensuring that the ADV can meet the 
assumptions for its operation specified in the SGTR analysis. Since the 
radiological consequences of a SGTR are not affected

[[Page 42549]]

by the addition of the proposed TS, the margin of safety is not changed 
significantly.
    Therefore, the proposed addition of a new TS to address the 
operability of the ADVs does not involve a significant reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 
Exelon Way, Kennett Square, PA 19348.
    NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
    Date of amendment request: February 13, 2014. A publicly-available 
version is in the ADAMS under Accession No. ML14050A374.
    Description of amendments request: The amendments would modify the 
as-found lift tolerances in the surveillance requirement for the 
pressurizer safety valves (PSVs).
    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?
    No.
    The proposed change, modifying the as-found and as-left lift 
settings in the Surveillance Requirement of the PSVs, does not change 
the design function or operation of the PSVs and it does not change the 
way the PSVs are maintained, tested, or inspected. The PSVs are not 
accident initiators; they operate in response to the pressurization of 
the Reactor Coolant System (RCS). They limit the pressure of the RCS to 
less than the allowable American Society of Mechanical Engineers Boiler 
and Pressure Vessel, Section III Code during an accident or transient. 
Analyses were performed of peak pressure events, which are evaluated 
against the RCS limit. Action of the PSVs is required to mitigate the 
consequences of these events. The change in the setpoint tolerance and 
a change in one valve's nominal setpoint were explicitly considered in 
the analysis of these events. The RCS pressure remained below the 
required limits with these changes considered. Therefore, this change 
does not impact the ability of the PSVs to perform their safety 
function during evaluated accidents.
    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?
    No.
    The proposed change, modifying the as-found and as-left lift 
settings in the Surveillance Requirement of the PSVs, does not change 
the PSVs design function to maintain RCS pressure below the RCS 
pressure Safety Limit of 2750 psia [pounds per square inch absolute] 
during design basis accidents nor does it affect the PSVs ability to 
perform this design function. The proposed change does not require any 
modification to the plant (other than the setpoint change) or change 
equipment operation or testing. It also does not create any credible 
new failure mechanisms, malfunctions, or accident initiators that would 
cause an accident not previously considered.
    Therefore the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    No.
    The proposed change, modifying the as-found and as-left lift 
settings in the Surveillance Requirement of the PSVs, does not involve 
a significant reduction in the margin of safety in maintaining RCS 
pressure below Safety Limits of 2750 psia during design basis 
accidents. The analyses conducted in support of this proposed change 
evaluated the ability of the PSVs to maintain an adequate safety margin 
assuming the change in setpoint tolerances and a change in one valve's 
nominal setpoint. The analysis determined that the response of the PSVs 
would maintain an adequate safety margin to the reactor coolant Safety 
Limit of 2750 psia.
    Therefore the proposed change does not involve a significant 
reduction in the margin of safety of maintaining RCS pressure the below 
RCS pressure Safety Limit.
    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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 
Exelon Way, Kennett Square, PA 19348.
    NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
    Date of amendment request: May 1, 2014. A publicly-available 
version is in the ADAMS under Accession No. ML14125A015.
    Description of amendments request: The amendments would modify the 
Technical Specifications (TSs) by relocating specific surveillance 
frequencies to a licensee-controlled program with the implementation of 
Nuclear Energy Institute 04-10, ``Risk Informed Method for Control of 
Surveillance Frequencies.''
    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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated; or
    Response: No.
    The proposed change relocates the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, the 
probability of any accident previously evaluated is not significantly 
increased. The systems and components required by the Technical 
Specifications for which the surveillance frequencies are relocated are 
still required to be operable, meet the acceptance criteria for the 
surveillance requirements, and be capable of performing any mitigation 
function assumed in the accident analysis. As a result, the 
consequences of any accident previously evaluated are not significantly 
increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different type of accident 
from any accident previously evaluated; or
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical

[[Page 42550]]

alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or different 
requirements. The changes do not alter assumptions made in the safety 
analysis. The proposed changes are consistent with the safety analysis 
assumptions and current plant operating practice.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Response: No.
    The design, operation, testing methods, and acceptance criteria for 
systems, structures and components specified in applicable codes and 
standards (or alternatives approved for use by the NRC) will continue 
to be met as described in the plant licensing basis (including the 
updated final safety analysis report and the bases to the TS), since 
these are not affected by changes to the surveillance frequencies. 
Similarly, there is no impact to safety analysis acceptance criteria as 
described in the plant licensing basis. To evaluate a change in the 
relocated surveillance frequency, Calvert Cliffs will perform a 
probabilistic risk evaluation using the guidance contained in NRC 
approved NEI 04-10, Revision 1 in accordance with the TS Surveillance 
Frequency Control Program. Nuclear Energy Institute 04-10, Revision 1 
methodology provides reasonable acceptance guidelines and methods for 
evaluating the risk increase of proposed changes to surveillance 
frequencies consistent with Regulatory Guide 1.177.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment's request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 
Exelon Way, Kennett Square, PA 19348.
    NRC Branch Chief: Benjamin G. Beasley.
Florida Power and Light Company (FPL), et al., Docket Nos. 50-335 and 
50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
    Date of amendment request: February 20, 2014. Available in ADAMS 
under Accession No. ML14070A087.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) by relocating specific surveillance 
frequency requirements to a licensee-controlled program with 
implementation of Nuclear Energy Institute (NEI) 04-10, ``Risk Informed 
Technical Specification Initiative 5b, Risk Informed Method for Control 
of Surveillance Frequencies'' (ADAMS Accession No. ML071360456). The 
licensee stated that the NEI 04-10 methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase of 
proposed changes to surveillance frequencies, consistent with 
Regulatory Guide 1.177, ``An Approach for Plant-Specific Risk-Informed 
Decision-Making: Technical Specifications'' (ADAMS Accession No. 
ML003740176). The licensee stated that the changes are consistent with 
NRC-approved Technical Specification Task Force (TSTF) Standard 
Technical Specifications change TSTF-425, ``Relocate Surveillance 
Frequencies to Licensee Control--RITSTF [Risk Informed Technical 
Specifications Task Force] Initiative 5b,'' Revision 3 (ADAMS Accession 
No. ML090850642). The Federal Register notice published on July 6, 2009 
(74 FR 31996), announced the availability of TSTF-425, Revision 3.
    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 as follows:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change relocates the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, the 
probability of any accident previously evaluated is not significantly 
increased. The systems and components required by the Technical 
Specifications for which the surveillance frequencies are relocated are 
still required to be operable, meet the acceptance criteria for the 
surveillance requirements, and be [sic] capable of performing any 
mitigation function assumed in the accident analysis. As a result, the 
consequences of any accident previously evaluated are not significantly 
increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements. The 
changes do not alter assumptions made in the safety analysis 
assumptions and current plant operating practice.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction in 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria for 
systems, structures, and components (SSCs), specified in applicable 
codes and standards (or alternatives approved for use by the NRC) will 
continue to be met as described in the plant licensing basis (including 
the final safety analysis report and bases to TS), since these are not 
affected by changes to the surveillance frequencies. Similarly, there 
is no impact to safety analysis acceptance criteria as described in the 
plant licensing basis. To evaluate a change in the relocated 
surveillance frequency, FPL will perform a probabilistic risk 
evaluation using the guidance contained in NRC-approved NEI 04-10, 
Revision 1 in accordance with the TS Surveillance Frequency Control 
Program. NEI 04-10, Revision 1, methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase of 
proposed changes to surveillance frequencies consistent with Regulatory 
Guide (RG) 1.177.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff 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: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700

[[Page 42551]]

Universe Blvd. MS LAW/JB, Juno Beach, Florida 33408-0420.
    NRC Acting Branch Chief: Lisa M. Regner.
Florida Power and Light Company (FPL), Docket Nos. 50-250 and 50-251, 
Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, 
Florida
    Date of amendment request: April 9, 2014. Available in ADAMS under 
Accession No. ML14105A042.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) by relocating specific surveillance 
frequency requirements to a licensee-controlled program with 
implementation of Nuclear Energy Institute (NEI) 04-10, ``Risk Informed 
Technical Specification Initiative 5b, Risk Informed Method for Control 
of Surveillance Frequencies'' (ADAMS Accession No. ML071360456). The 
licensee stated that the NEI 04-10 methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase of 
proposed changes to surveillance frequencies, consistent with 
Regulatory Guide 1.177, ``An Approach for Plant-Specific Risk-Informed 
Decision-Making: Technical Specifications'' (ADAMS Accession No. 
ML003740176). The licensee stated that the changes are consistent with 
NRC-approved Technical Specification Task Force (TSTF) Standard 
Technical Specifications change TSTF-425, ``Relocate Surveillance 
Frequencies to Licensee Control--RITSTF [Risk Informed Technical 
Specifications Task Force] Initiative 5b,'' Revision 3 (ADAMS Accession 
No. ML090850642). The Federal Register notice published on July 6, 2009 
(74 FR 31996), announced the availability of TSTF-425, Revision 3.
    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 as follows:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change relocates the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, the 
probability of any accident previously evaluated is not significantly 
increased. The systems and components required by the Technical 
Specifications for which the surveillance frequencies are relocated are 
still required to be operable, meet the acceptance criteria for the 
surveillance requirements, and be [sic] capable of performing any 
mitigation function assumed in the accident analysis. As a result, the 
consequences of any accident previously evaluated are not significantly 
increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 changes relocate the surveillance frequencies for 
Surveillance Requirements that have a set periodicity from the TS to a 
licensee controlled Surveillance Frequency Control Program. This change 
does not alter any existing surveillance frequencies. Within the 
constraints of the Program, the licensee will be able to change the 
periodicity of these surveillance requirements. Relocating the 
surveillance frequencies does not impact the ability of structures, 
systems or components (SSCs) from performing there [sic] design 
functions, and thus, does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements. The 
changes do not alter assumptions made in the safety analysis 
assumptions and current plant operating practice.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction in 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria for 
structures, systems, and components (SSCs) specified in applicable 
codes and standards (or alternatives approved for use by the NRC) will 
continue to be met as described in the plant licensing basis (including 
the final safety analysis report and bases to TS), since these are not 
affected by changes to the surveillance frequencies. Similarly, there 
is no impact to safety analysis acceptance criteria as described in the 
plant licensing basis. To evaluate a change in the relocated 
surveillance frequency, FPL will perform a probabilistic risk 
evaluation using the guidance contained in NRC-approved NEI 04-10, 
Revision 1 in accordance with the TS Surveillance Frequency Control 
Program. NEI 04-10, Revision 1, methodology provides reasonable 
acceptance guidelines and methods for evaluating the risk increase of 
proposed changes to surveillance frequencies consistent with Regulatory 
Guide (RG) 1.177, An Approach for Plant-Specific Risk-Informed 
Decision-Making: Technical Specifications.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff 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: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd. MS LAW/JB, 
Juno Beach, Florida 33408-0420.
    NRC Acting Branch Chief: Lisa M. Regner.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama
    Date of amendment request: June 3, 2014. A publicly available 
version is in ADAMS under Accession No. ML14154A136.
    Description of amendment request: The amendments would revise the 
Technical Specification Limiting Condition for Operation 3.3.1 and 
Surveillance Requirement 3.2.4.2 regarding the reactor trip system 
instrumentation.
    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 changes do not adversely affect accident initiators or 
precursors nor alter the design assumptions, conditions, or 
configuration of the facility or the

[[Page 42552]]

manner in which the plant is operated and maintained. The proposed 
changes do not alter or prevent the ability of structures, systems, and 
components (SSCs) from performing their intended function to mitigate 
the consequences of an initiating event within the assumed acceptance 
limits. The proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating the 
radiological consequences of an accident previously evaluated. Further, 
the proposed changes do not increase the types or amounts of 
radioactive effluent that may be released offsite, nor significantly 
increase individual or cumulative occupational/public radiation 
exposures. The proposed changes are consistent with safety analysis 
assumptions and resultant consequences.
    Therefore, the proposed changes do not increase the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not result in a change in the manner in 
which the Reactor Trip System (RTS) and engineered safety features 
actuation system (ESFAS) provide plant protection. The RTS and ESFAS 
will continue to have the same setpoints after the proposed changes are 
implemented. There are no design changes associated with the license 
amendment.
    The changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. In addition, 
the changes do not impose any new or different requirements or 
eliminate any existing requirements. The changes do not alter 
assumptions made in the safety analysis. The proposed changes are 
consistent with the safety analysis assumptions and current plant 
operating practice.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria are 
not impacted by these changes. Redundant RTS and ESFAS trains are 
maintained, and diversity with regard to the signals that provide 
reactor trip and engineered safety features actuation is also 
maintained. All signals credited as primary or secondary, and all 
operator actions credited in the accident analyses will remain the 
same. The proposed changes will not result in plant operation in a 
configuration outside the design basis.
    Therefore, the proposed changes 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 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: Leigh D. Perry, SVP & General Counsel of 
Operations and Nuclear, Southern Nuclear Operating Company, 40 Iverness 
Center Parkway, Birmingham, AL 35201.
    NRC Branch Chief: Robert J. Pascarelli.
Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant (HNP), Unit Nos. 1 and 2, Appling County, Georgia
    Date of amendment request: March 17, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14076A141.
    Description of amendment request: The proposed amendments would 
modify Technical Specification (TS) definition of Shutdown Margin (SDM) 
to require calculation of the SDM at a reactor moderator temperature of 
68[emsp14][deg]F or a higher temperature that represents the most 
reactive state throughout the operating cycle. This change is needed to 
address new Boiling Water Reactor (BWR) fuel designs which may be more 
reactive at shutdown temperatures above 68 [deg]F.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:
    SNC has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment(s) by focusing on 
the three standards set forth in 10 CFR 50.92, Issuance of amendment, 
as discussed below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the definition of SDM. SDM is not an 
initiator to any accident previously evaluated. Accordingly, the 
proposed change to the definition of SDM has no effect on the 
probability of any accident previously evaluated. SDM is an assumption 
in the analysis of some previously evaluated accidents and inadequate 
SDM could lead to an increase in consequences for those accidents. 
However, the proposed change revises the SDM definition to ensure that 
the correct SDM is determined for all fuel types at all times during 
the fuel cycle. As a result, the proposed change does not adversely 
affect the consequences of any accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the definition of SDM. The change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operations. The change does not alter 
assumptions made in the safety analysis regarding SDM.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the definition of SDM. The proposed 
change does not alter the manner in which safety limits, limiting 
safety system settings or limiting conditions for operation are 
determined. The proposed change ensures that the SDM assumed in 
determining safety limits, limiting safety system settings or limiting 
conditions for operation is correct for all BWR fuel types at all times 
during the fuel cycle.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed change presents 
no significant hazards consideration under the standards set forth in 
10 CFR 50.92

[[Page 42553]]

    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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Robert Pascarelli.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 And 50-
281 Surry Power Station, Units 1 and 2, Surry County, Virginia
    Date of amendment request: April 11, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14112A073.
    Description of amendment request: The proposed license amendment 
requests the changes to the Technical Specification (TS) TS 4.2, 
``Augmented Inspections,'' and TS 4.15, ``Augmented Inservice 
Inspection Program for High Energy Lines Outside of Containment,'' by 
relocating to the Surry Technical Requirements Manual (TRM). In 
addition, TS 6.4.U, ``Augmented Inspections and Examinations,'' will be 
added to the Administrative Controls Section 6.4, ``Unit Operating 
Procedures and Programs.'' The proposed relocation of the TS 4.2 and TS 
4.15 requirements to the TRM is appropriate since these requirements do 
not satisfy the categories and criteria of 10 CFR 50.36(c) for 
inclusion in the TS. Along with the relocation of the TS 4.2 and TS 
4.15 requirements to the TRM, the Bases for TS 4.2 and TS 4.15 are also 
being relocated to the TRM.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change relocates Technical Specification (TS) 4.2, 
``Augmented Inspections,'' TS 4.15, ``Augmented Inservice Inspection 
Program for High Energy Lines Outside of Containment,'' and the 
associated TS Bases to the Surry Technical Requirements Manual (TRM). 
In addition, TS 6.4.U, ``Augmented Inspections and Examinations,'' will 
be added to the Surry TS. The proposed relocation of the TS 4.2 and TS 
4.15 requirements to the TRM is appropriate since these requirements do 
not satisfy the categories and criteria of 10CFR50.36(c), which 
specifies what items qualify for inclusion in the TS.
    Specifically, the TS 4.2 augmented inspections of the low head 
safety injection piping located in the valve pit, the reactor coolant 
pump flywheel, the low pressure turbine rotor blades, sensitized 
stainless steel, and TS 4.15 augmented inspections of the welds in the 
main steam and main feedwater lines in the main steam valve house of 
each unit will be relocated to the TRM. The augmented inspections, 
which are performed in addition to required ASME Code Section Xl 
inspections/examinations, will continue to be performed as required by 
the TRM.
    The plant systems and components to which the augmented inspections 
apply will not be operated in a different manner. The proposed 
relocation of the augmented inspections does not involve a physical 
change to the plant or a change in the manner in which the plant is 
operated or controlled.
    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 any physical alteration of 
plant equipment. As such, no new or different types of equipment will 
be installed, and the basic operation of installed plant systems and 
components, to which the augmented inspections apply, is unchanged.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not reduce a margin of safety because the 
relocation of the augmented inspections to the TRM has no impact on any 
safety analysis assumptions, as indicated by the fact that the 
requirements do not meet the 10CFR50.36(c) criteria for inclusion in 
the TS. In addition, the augmented inspections will be moved to the TRM 
without change and will continue to be performed as required by the 
TRM.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Robert Pascarelli.
ZionSolutions LLC (ZS), Docket Nos. 50-295 and 50-304, Zion Nuclear 
Power Station (ZNPS), Units 1 and 2, Lake County, Illinois
    Date of amendment request: May 27, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14148A295.
    Description of amendment request: The license amendment request 
proposes changes to ZNPS Defueled Station Emergency Plan (DSEP) in 
accordance with 10 CFR 50.54(q). ZS proposes removal of the various 
emergency actions related to the former spent fuel pool, the transfer 
of responsibility for implementing the Emergency Plan to the 
Independent Spent Fuel Storage Installation (ISFSI) Shift Supervisor, a 
revised emergency plan organization, abandonment of the Control Room 
consistent with the current state of decommissioning, transition to NEI 
99-01 Revision 6 and reformatting consistent with current industry 
practice.
    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 change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. ZS has, in effect, a U.S. Nuclear Regulatory Commission-
approved (NRC) emergency plan. The remaining ZNPS accident (Radioactive 
Waste Handling Accident) and the credible accidents involving the ISFSI 
and the Modular, Advanced Generation, Nuclear All-purpose Storage 
(MAGNASTOR) system have been analyzed and determined that none result 
in doses to the public beyond the owner controlled area boundary that 
would exceed the U.S. Environmental Protection Agency's (EPA) 
Protective Action Guides (PAGs). These analyses have not changed. With 
spent fuel relocated to the ISFSI, the Spent Fuel Pool previously 
analyzed events (Loss of Spent Fuel Pool Cooling,

[[Page 42554]]

Loss of Spent Fuel Pool Inventory, and Fuel Handling Accident in the 
Fuel Building) are no longer credible.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    (2) Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. ZS has, in effect, an NRC-approved emergency plan. The 
remaining ZNPS accident (Radioactive Waste Handling Accident) and the 
credible accidents involving the ISFSI and MAGNASTOR system have been 
analyzed and determined that none result in doses to the public beyond 
the owner controlled area boundary that would exceed the EPA's PAGs. 
These analyses have not changed. With spent fuel relocated to the 
ISFSI, the Spent Fuel Pool previously analyzed events (Loss of Spent 
Fuel Pool Cooling, Loss of Spent Fuel Pool Inventory, and Fuel Handling 
Accident in the Fuel Building) are no longer credible. Accidents 
associated with the ISFSI are addressed in the MAGNASTOR Final Safety 
Analysis Report.
    Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    (3) Does the change involve a significant reduction in a margin of 
safety?
    No. Margin of safety is related to the ability of the fission 
product barriers (fuel cladding, reactor coolant system, and primary 
containment) to perform their design functions during and following 
postulated accidents. ZS has, in effect, an NRC-approved emergency 
plan. The remaining ZNPS accident (Radioactive Waste Handling Accident) 
and the credible accidents involving the ISFSI and MAGNASTOR system 
have been analyzed and determined that none result in doses to the 
public beyond the owner controlled area boundary that would exceed the 
EPA's PAGs These analyses have not changed. With spent fuel relocated 
to the ISFSI, the Spent Fuel Pool previously analyzed events (Loss of 
Spent Fuel Pool Cooling, Loss of Spent Fuel Pool Inventory, and Fuel 
Handling Accident in the Fuel Building) are no longer credible.
    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: Russ Workman, Deputy General Counsel, 
EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT 
84101.
    NRC Branch Chief: Bruce Watson.

III. Notice of Issuance of Amendments to Facility Operating Licenses 
and Combined Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New 
York
    Date of amendment request: May 23, 2013, as supplemented by letter 
dated October 11, 2013.
    Brief description of amendment(s): The amendments revise the 
Technical Specifications to risk-inform requirements regarding selected 
Required Action End States. Specifically, the changes permit an end 
state of Mode 4 rather than an end state of Mode 5 consistent with 
Technical Specification Task Force (TSTF) Traveler TSTF 432-A, Revision 
1, ``Change in Technical Specifications End States WCAP-16294.''
    Date of issuance: July 7, 2014.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: Unit 2-275; Unit 3-252. A publicly-available version 
is in ADAMS under Accession No. ML14122A303; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Facility Operating License Nos. DPR-26 and DPR-64: The amendment 
revised the Facility Operating License and the Technical 
Specifications.
    Date of initial notice in Federal Register: July 23, 2013 (78 FR 
44170). The supplemental letter provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 7, 2014.
    No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas
    Date of amendment request: June 11, 2013, as supplemented by letter 
dated December 11, 2013.
    Brief description of amendment: The amendment revised Technical 
Specification 2.1.1.1, to add a provision for the determination of the 
maximum local fuel pin centerline temperature using the NRC reviewed 
and approved COPERNIC fuel performance computer code.
    Date of issuance: July 9, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 249. A publicly-available version is in ADAMS under 
Accession No. ML14169A475; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-51: Amendment revised 
the

[[Page 42555]]

Technical Specifications and the renewed facility operating license.
    Date of initial notice in Federal Register: April 1, 2014 (79 FR 
18331). The supplemental letter dated December 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 amendment is contained 
in a Safety Evaluation dated July 9, 2014.
    No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi
    Date of application for amendment: November 8, 2013.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) definition of ``Shutdown Margin'' (SDM) to require 
calculation of the SDM at a reactor moderator temperature of 68 degrees 
Fahrenheit ([deg]F) or a higher temperature that represents the most 
reactive state throughout the operating cycle. This change is needed to 
address new Boiling Water Reactor (BWR) fuel designs which may be more 
reactive at shutdown temperatures above 68 [deg]F.
    This TS change is part of the Consolidated Line Item Improvement 
Process (CLIIP) TS Task Force (TSTF) Traveler TSTF-535, Revision 0, 
``Revise Shutdown Margin Definition to Address Advanced Fuel Designs.'' 
The licensee stated there are no variations or deviations from the NRC 
staff's model safety evaluation.
    Date of issuance: June 30, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 198. A publicly-available version is in ADAMS under 
Accession No. ML14106A133; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-29: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 4, 2014 (79 FR 
12244).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 30, 2014.
    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Nine Mile Point Nuclear Station, LLC, 
Docket No. 50-220, Nine Mile Point Nuclear Station, Unit No. 1, Oswego 
County, New York
    Date of application for amendment: June 11, 2012, as supplemented 
by letters dated February 27, March 27, April 30, and December 9, 2013; 
and January 22, March 14, April 15, May 9, and May 23, 2014.
    Brief description of amendment: The amendment authorizes the 
transition of the Nine Mile Point Nuclear Station, Unit 1, fire 
protection program to a risk-informed, performance-based program based 
on National Fire Protection Association (NFPA) 805, in accordance with 
10 CFR 50.48(c). NFPA 805 allows the use of performance-based methods 
such as fire modeling and risk-informed methods such as fire 
probabilistic risk assessment to demonstrate compliance with the 
nuclear safety performance criteria.
    Date of issuance: June 30, 2014.
    Effective date: As of its date of issuance and shall be implemented 
by 180 days from the date of issuance.
    Amendment No.: 215. A publicly available version is in ADAMS under 
Accession No. ML14126A003; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-63: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: September 11, 2012 (77 
FR 55874).
    The supplements dated February 27, March 27, April 30, and December 
9, 2013; and January 22, March 14, April 15, May 9, and May 23, 2014, 
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 June 30, 2014.
    No significant hazards consideration comments received: No.
    Amendment No.: 215. A publicly available version is in ADAMS under 
Accession No. ML14126A003; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-63: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: September 11, 2012 (77 
FR 55874).
    The supplements dated February 27, March 27, April 30, and December 
9, 2013; and January 22, March 14, April 15, May 9, and May 23, 2014, 
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 June 30, 2014.
    No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County, 
Texas
    Date of amendment request: March 28, 2013, as supplemented by 
letters dated July 16, October 22, November 26, and December 17, 2013, 
and January 16, April 17, and May 1, 2014.
    Brief description of amendment: The amendments revised Technical 
Specification (TS) 3.7.16, ``Fuel Storage Pool Boron Concentration,'' 
TS 3.7.17, ``Spent Fuel Assembly Storage,'' TS 4.3, ``Fuel Storage,'' 
and TS 5.5, ``Programs and Manuals,'' for storage of uprated fuel in 
Region II of the spent fuel pool. Changes to TS 3.7.16 reflect a change 
in the required fuel storage pool soluble boron concentration based on 
the results of a new criticality analysis. Changes to TS 3.7.17 include 
new spent fuel pool loading restrictions in terms of allowable storage 
patterns, and minimum burnup requirements as a function of enrichment, 
fuel type, and fuel reactivity category. The revised TS 4.3 section 
includes updates to the minimum soluble boron concentration, Region I 
fuel assembly spacing, specific new or partially spent fuel assembly 
storage restrictions in Region II consistent with TS 3.7.17, and 
general Region II storage restrictions consistent with TS 3.7.17. The 
change to TS 5.5 adds TS program 5.5.22, ``Neutron Absorber Monitoring 
Program.''
    Date of issuance: July 1, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: Unit 1-162; Unit 2-162. A publicly-available 
version is in ADAMS under Accession No. ML14160A035; documents related 
to these amendments are listed in the

[[Page 42556]]

Safety Evaluation enclosed with the amendments.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: November 5, 2013 (78 FR 
66391). The NRC staff's original proposed no significant hazards 
consideration determination was based on letters dated March 28, and 
July 16, 2013. The supplements dated October 22, November 26, and 
December 17, 2013, and January 16, April 17, and May 1, 2014, 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 July 1, 2014.
    No significant hazards consideration comments received: No.
NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin
    Date of amendment request: January 15, 2013, as supplemented on 
March 1, April 18, and September 12, 2013, and March 11, 2014.
    Description of amendment: The license amendment revised Technical 
Specifications 5.6.5, ``Reactor Coolant System (RCS) Pressure and 
Temperature Limits Report (PTLR),'' to allow the use of two new 
methodologies for determining RCS pressure and temperature limits at 
the Point Beach Nuclear Plant, Units 1 and 2.
    Date of issuance: June 30, 2014.
    Effective date: As of the date of issuance and shall be implemented 
with 180 days.
    Amendment Nos.: 250 (Unit 1) and 254 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML14126A378; documents related 
to this amendment are listed in the Safety Evaluation enclosed with the 
amendment.
    Renewed Facility Operating License Nos. DPR-24 and DPR-27: The 
amendment revised the Renewed Facility Operating License and the 
Technical Specifications.
    Date of initial notice in Federal Register: June 11, 2013 (78 FR 
35062). The supplemental letters dated March 1, April 18, and September 
12, 2013, and March 11, 2014, 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 June 30, 2014.
    No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
    Date of amendment requests: August 12, 2013, as supplemented by 
letters dated January 24, March 13, and March 25, 2014.
    Brief description of amendments: The licensee requested to revise 
the Technical Specifications to, in effect, extend the Type A primary 
containment Integrated Leak Rate Test intervals to fifteen years and 
the Type C local leak rate test intervals to 75 months, and incorporate 
the regulatory positions stated in RG 1.163.
    Date of issuance: July 3, 2014.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: Unit 1, 282; Unit 2, 282. A publicly-available 
version is in ADAMS under Accession No. ML14148A235; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-32 and DPR-37: The 
amendments revise the Renewed Facility Operating Licenses and the 
Technical Specifications.
    Date of initial notice in Federal Register: October 29, 2013 (78 FR 
64548). The supplemental letters dated January 24, March 13, and March 
25, 2014, 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 July 3, 2014.
    No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri
    Date of application for amendment: September 26, 2013.
    Brief description of amendment: The amendment revised Technical 
Specification Surveillance Requirement (SR) 3.7.10.1 and SR 3.7.13.1 to 
reduce the required run time for periodic operation of the control room 
pressurization system filter trains and emergency exhaust system filter 
trains, with heaters on, from 10 hours to 15 minutes. The amendment is 
consistent with plant-specific options provided in the NRC's model 
safety evaluation in Technical Specifications Task Force (TSTF) 
Traveler TSTF-522, Revision 0, ``Revise Ventilation System Surveillance 
Requirements to Operate for 10 hours per Month,'' as part of the 
consolidated line item improvement process.
    Date of issuance: July 1, 2014.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 209. A publicly-available version is in ADAMS under 
Accession No. ML14175A390; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-30: The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2014 (79 FR 
3418).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 1, 2014.
    No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas
    Date of amendment request: December 17, 2013.
    Brief description of amendment: The amendment revised Technical 
Specification Surveillance Requirement (SR) 3.7.10.1 and SR 3.7.13.1 to 
reduce the required run time for periodic operation of the control room 
pressurization system filter trains and emergency exhaust system filter 
trains, with heaters on, from 10 hours to 15 minutes. The amendment is 
consistent with plant-specific options provided in the NRC's model 
safety evaluation in Technical Specifications Task Force (TSTF) 
Traveler TSTF-522, Revision 0, ``Revise Ventilation System Surveillance 
Requirements to Operate for 10 hours per Month,'' as part of the 
consolidated line item improvement process.
    Date of issuance: July 1, 2014.

[[Page 42557]]

    Effective date: As of its date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 208. A publicly-available version is in ADAMS under 
Accession No. ML14157A082; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 18, 2014 (79 FR 
15151).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 1, 2014.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 14th day of July 2014.

    For the Nuclear Regulatory Commission.
Louise Lund,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2014-17257 Filed 7-21-14; 8:45 am]
BILLING CODE 7590-01-P