Biweekly Notice, Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 38585-38597 [2014-15770]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 79, Number 130 (Tuesday, July 8, 2014)]
[Notices]
[Pages 38585-38597]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2014-15770]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0159]


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 12, to June 25, 2014. The last biweekly 
notice was published on June 24, 2014.

DATES: Comments must be filed by August 7, 2014. A request for a 
hearing must be filed by September 8, 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 https://www.regulations.gov and search for Docket ID NRC-2014-0159. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     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: Sandra Figueroa, Office, U.S. Nuclear 
Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-
1262, email: Sandra.Figueroa@nrc.gov.

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0159 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 https://www.regulations.gov and search for Docket ID NRC-2014-0159.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. 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-2014-0159 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 https://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

[[Page 38586]]

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 https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed by the above date, the Commission or a presiding officer 
designated by the Commission or by the Chief Administrative Judge of 
the Atomic Safety and Licensing Board Panel, will rule on the request 
and/or petition; and the Secretary or the Chief Administrative Judge of 
the Atomic Safety and Licensing Board will issue a notice of a hearing 
or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.

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 https://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 https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.

[[Page 38587]]

    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 https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the 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 https://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 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the 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 obtaining information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York
    Date of amendment request: February 4, 2014. A publicly available 
version is in ADAMS under Accession No. ML14050A383.
    Description of amendment request: The amendment would revise 
Technical Specification 5.5.15, ``Containment Leakage Rate Testing 
Program,'' to extend the frequency of the Type A, or the Containment 
Integrated Leak Rate Test, from 10 to 15 years on a permanent basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment involves changes to the IP3 [Indian Point 
Unit No. 3] containment leakage rate testing program. The proposed 
amendment does not involve a physical change to the plant or a 
change in the manner in which the plant is operated or controlled. 
The primary containment function is to provide an essentially leak 
tight barrier against the uncontrolled release of radioactivity to 
the environment for postulated accidents. As such, the containment 
itself and the testing requirements to periodically demonstrate the 
integrity of the containment exist to ensure the plant's ability to 
mitigate the consequences of an accident do not involve any accident 
precursors or initiators.
    Therefore, the probability of occurrence of an accident 
previously evaluated is not significantly increased by the proposed 
amendment.
    The proposed amendment adopts the NRC [Nuclear Regulatory 
Commission] accepted guidelines of [Nuclear Energy Institute] NEI 
94-01, Revision 3-A, for development of the IP3 performance-based 
testing program for the Type A testing. Implementation of these 
guidelines continues to provide adequate assurance that during 
design basis accidents, the primary containment and its components 
would limit leakage rates to less than the

[[Page 38588]]

values assumed in the plant safety analyses. The potential 
consequences of extending the ILRT [integrated leak rate test] 
interval to 15 years have been evaluated by analyzing the resulting 
changes in risk. The increase in risk in terms of person-rem per 
year within 50 miles resulting from design basis accidents was 
estimated to be acceptably small and determined to be within the 
guidelines published in [Regulatory Guide] RG 1.174. Additionally, 
the proposed change maintains defense-in-depth by preserving a 
reasonable balance among prevention of core damage, prevention of 
containment failure, and consequence mitigation. Entergy has 
determined that the increase in conditional containment failure 
probability due to the proposed change would be very small.
    Therefore, it is concluded that the proposed amendment does not 
significantly increase the consequences of an accident previously 
evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for the development of the IP3 performance-
based leakage testing program, and establishes a 15-year interval 
for the performance of the containment ILRT. The containment and the 
testing requirements to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident do not involve any accident precursors 
or initiators. The proposed change does not involve a physical 
change to the plant (i.e., no new or different type of equipment 
will be installed) or a change to the manner in which the plant is 
operated or controlled.
    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?
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for the development of the IP3 performance-
based leakage testing program, and establishes a 15-year interval 
for the performance of the containment ILRT. This amendment does not 
alter the manner in which safety limits, limiting safety system 
setpoints, or limiting conditions for operation are determined. The 
specific requirements and conditions of the containment leakage rate 
testing program, as defined in the TS [technical specifications], 
ensure that the degree of primary containment structural integrity 
and leak-tightness that is considered in the plant's safety analysis 
is maintained. The overall containment leakage rate limit specified 
by the TS is maintained, and the Type A, Type B, and Type C 
containment leakage tests would be performed at the frequencies 
established in accordance with the NRC-accepted guidelines of NEI 
94-01, Revision 3-A.
    Containment inspections performed in accordance with other plant 
programs serve to provide a high degree of assurance that the 
containment would not degrade in a manner that is not detectable by 
an ILRT. A risk assessment using the current IP3 PSA [probabilistic 
safety assessment] model concluded that extending the ILRT test 
interval from ten years to 15 years results in a very small change 
to the risk profile.
    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., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Benjamin G. Beasley.
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: April 1, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14099A227.
    Description of amendment request: The amendments would revise the 
technical specifications by implementing Technical Specification Task 
Force Traveler 510, Revision 2, ``Revision to Steam Generator Program 
Inspection Frequencies and Tube Sample Selection.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG 
tube sample selection. A steam generator tube rupture (SGTR) event 
is one of the design basis accidents that are analyzed as part of a 
plant's licensing basis. The proposed SG tube inspection frequency 
and sample selection criteria will continue to ensure that the SG 
tubes are inspected such that the probability of a SGTR is not 
increased. The consequences of a SGTR are bounded by the 
conservative assumptions in the design basis accident analysis. The 
proposed change will not cause the consequences of a SGTR to exceed 
those assumptions. The proposed change to reporting requirements and 
clarifications of the existing requirements have no affect on the 
probability or consequences of a SGTR.
    Therefore, this 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 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 changes do 
not affect the design of the SGs or their method of operation. In 
addition, the proposed changes do not impact any other plant system 
or component.
    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 amendment 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.
    Changes associated with inspection frequency and tube selection 
criteria are consistent with TSTF-510 Revision 2 and are based on 
recent industry experience and are more effective in managing the 
frequency of verification of tube integrity and sample selection 
than those required by current TSs [technical specifications].
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 38589]]

    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 Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York
    Date of amendment request: October 8, 2013. A publicly-available 
version is in ADAMS under Accession No. ML13282A559.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) requirements to reduce the 
reactor pressure associated with the Reactor Core Safety Limit from 785 
psig to 685 psig in TS 2.1.1.1 and TS 2.1.1.2. The proposed amendment 
would address the potential to not meet the lower pressure TS safety 
limit associated with a Pressure Regulator Failure-Maximum Demand 
(Open) (PRFO) transient reported by General Electric (GE) in their 10 
CFR Part 21 Communication, Potential to Exceed Low Pressure Technical 
Specification Safety Limit, SC05-03, dated March 29, 2005.
    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.
    Decreasing the reactor pressure in TS Safety Limit 2.1.1.1 or 
2.1.1.2 for reactor rated thermal power ranges effectively expands 
the validity range for GEXL correlation and the calculation of 
Minimum Critical Power Ratio Safety Limit (MCPR). The [critical 
power ratio] CPR rises during the pressure reduction following the 
scram that terminates the PRFO transient. Since the change does not 
involve a modification of any plant hardware, the probability and 
consequence of the PRFO transient are essentially unchanged. The 
reduction in the reactor dome pressure value in the safety limit 
from 800 psia (785 psig) to 700 psia (685 psig) provides greater 
margin to accommodate the pressure reduction during the transient 
within the revised TS limit.
    The proposed change will continue to support the validity range 
for GEXL correlation and the calculation of MCPR as approved. The 
proposed TS revision involves no significant changes to the 
operation of any systems or components in normal or accident or 
transient operating conditions.
    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 reduction in the reactor pressure value in the 
safety limit from 800 psia (785 psig) to 700 psia (685 psig) 
reflects a wider range of applicability for the GEXL correlation for 
fuels in use at JAF and does not involve changes to the plant 
hardware or its operating characteristics. As a result, no new 
failure modes are being introduced.
    Therefore, the change does not introduce a new or different kind 
of accident from those previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety is established through the design of the 
plant structures, systems, and components, and through the 
parameters for safe operation and setpoints for the actuation of 
equipment relied upon to respond to transients and design basis 
accidents. The proposed change in the reactor pressure safety limit 
enhances the safety margin, which protects the fuel cladding 
integrity during a depressurization transient, but does not change 
the requirements governing operation or availability of safety 
equipment assumed to operate to preserve the margin of safety. The 
change does not alter the behavior of plant equipment, which remains 
unchanged. The available pressure range is expanded by the change, 
thus offering greater margin for pressure reduction during the 
transient.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    Based on the above, Entergy concludes that the proposed 
amendment presents no significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Benjamin G. Beasley.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas
    Date of amendment request: January 29, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14029A438.
    Description of amendment request: The amendment would revise the 
facility operating license and technical specifications to reflect 
adoption of a new fire protection licensing basis which complies with 
the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance 
in NRC Regulatory Guide (RG) 1.205, Revision 1, ``Risk-Informed 
Performance-Based Fire Protection for Existing Light-Water Nuclear 
Power Plants,'' December 2009 (ADAMS Accession No. ML092730314). The 
license amendment request follows Nuclear Energy Institute (NEI) 04-02, 
Revision 2, ``Guidance for Implementing a Risk-Informed, Performance-
Based Fire Protection Program under 10 CFR 50.48(c),'' April 2008. The 
submittal describes the methodology used to demonstrate compliance 
with, and transition to, National Fire Protection Association (NFPA) 
805, and includes regulatory evaluations, probabilistic risk 
assessment, change evaluations, proposed modifications for non-
compliances, and supporting attachments.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1: The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated
    Operation of Arkansas Nuclear One, Unit 1 (ANO-1) in accordance 
with the proposed amendment does not result in a significant 
increase in the probability or consequences of accidents previously 
evaluated. The proposed amendment does not affect accident 
initiators or precursors as described in the ANO-1 Safety Analysis 
Report (SAR), nor does it adversely alter design assumptions, 
conditions, or configurations of the facility, and it does not 
adversely impact the ability of structures, systems, or components 
(SSCs) to perform their intended function to mitigate the 
consequences of accidents described and evaluated in the SAR. The 
proposed changes do not physically alter safety-related systems nor 
affect the way in which safety-related systems perform their 
functions as required by the accident analysis. The SSCs required to 
safely shut down the reactor and to maintain it in a safe shutdown 
condition will remain capable of performing their design functions.
    The purpose of this amendment is to permit ANO-1 to adopt a new 
risk-informed, performance-based fire protection licensing basis 
that complies with the requirements in 10 CFR 50.48(a) and 10 CFR 
50.48(c), as well as the guidance contained in Regulatory Guide (RG) 
1.205. The NRC considers that NFPA 805 provides an acceptable 
methodology and performance criteria for licensees to identify fire 
protection requirements that are an acceptable

[[Page 38590]]

alternative to the 10 CFR Part 50, Appendix R, fire protection 
features (69 FR 33536; June 16, 2004).
    The purpose of the fire protection program is to provide 
assurance, through defense-in-depth, that the NRC's fire protection 
objectives are satisfied. These objectives are: (1) preventing fires 
from starting; (2) rapidly detecting and controlling fires and 
promptly extinguishing those fires that do occur, thereby limiting 
fire damage; (3) providing an adequate level of fire protection for 
SSCs important to safety, so that a fire that is not promptly 
extinguished will not prevent essential plant safety functions from 
being performed; and (4) ensuring that fires will not significantly 
increase the risk of radioactive releases to the environment. In 
addition, fire protection systems must be designed such that their 
failure or inadvertent operation does not adversely impact the 
ability of the SSCs important to safety to perform their safety-
related functions.
    NFPA 805, taken as a whole, provides an acceptable alternative 
for satisfying General Design Criterion 3 (GDC 3) of Appendix A to 
10 CFR Part 50, meets the underlying intent of the NRC's existing 
fire protection regulations and guidance, and achieves defense-in-
depth along with the goals, performance objectives, and performance 
criteria specified in NFPA 805, Chapter 1. In addition, if there are 
any increases in core damage frequency (CDF) or risk as a result of 
the transition to NFPA 805, the increase will be small, bounded by 
the delta risk requirements of NFPA 805, and consistent with the 
intent of the Commission's Safety Goal Policy.
    Engineering analyses, which may include engineering evaluations, 
probabilistic risk assessments, and fire modeling calculations, have 
been performed to demonstrate that the performance-based 
requirements of NFPA 805 have been met. The SAR documents the 
analyses of design basis accidents (DBAs) at ANO-1. All accident 
analysis acceptance criteria will continue to be met with the 
proposed amendment. The proposed changes will not affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of any accident 
previously evaluated. The proposed changes will not alter any 
assumptions or change any mitigation actions for the radiological 
consequence evaluations in the ANO-1 SAR. In addition, the 
applicable radiological dose acceptance criteria will continue to be 
met.
    Based on the above, the implementation of this amendment to 
transition the Fire Protection Plan (FPP) at ANO-1 to one based on 
NFPA 805, in accordance with 10 CFR 50.48(c), does not result in a 
significant increase in the probability of any accident previously 
evaluated. In addition, all equipment required to mitigate an 
accident remains capable of performing the assumed function. 
Therefore, the consequences of any accident previously evaluated are 
not significantly increased with the implementation of this 
amendment.
    Criterion 2: The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from Any Accident Previously 
Evaluated
    Operation of ANO-1 in accordance with the proposed amendment 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated. Previously analyzed 
accidents with potential offsite dose consequences were included in 
the evaluation of the transition to NFPA 805. The proposed amendment 
does not impact these accident analyses. The proposed change does 
not alter the requirements or functions for systems required during 
accident conditions as assumed in the licensing basis analyses and/
or DBA radiological consequences evaluations.
    Implementation of the new risk-informed, performance-based fire 
protection licensing basis, which complies with the requirements in 
10 CFR 50.48(a) and 10 CFR 50.48(c), as well as the guidance 
contained in RG 1.205, will not result in new or different kinds of 
accidents. The NRC considers that NFPA 805 provides an acceptable 
methodology and performance criteria for licensees to identify fire 
protection systems and features that are an acceptable alternative 
to the 10 CFR 50, Appendix R fire protection features (69 FR 33536, 
June 16, 2004). No new modes of operation are introduced by the 
proposed amendment, nor will it create any failure mode not bounded 
by previously evaluated accidents. Further, the impacts of the 
proposed change are not directly assumed in any safety analysis to 
initiate an accident sequence.
    The requirements in NFPA 805 address only fire protection and 
the impacts of fire effects on the plant have been evaluated. The 
proposed fire protection program changes do not involve new failure 
mechanisms or malfunctions that could initiate a new or different 
kind of accident beyond those already analyzed in the SAR. Based on 
this, as well as the discussion above, the implementation of this 
amendment to transition the FPP at ANO-1 to one based on NFPA 805, 
in accordance with 10 CFR 50.48(c), does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Criterion 3: The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety
    Operation of ANO-1 in accordance with the proposed amendment 
does not involve a significant reduction in a margin of safety. The 
transition to a new risk-informed, performance-based fire protection 
licensing basis that complies with the requirements in 10 CFR 
50.48(a) and 10 CFR 50.48(c) does 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 affected by this change. The proposed 
amendment does not adversely affect existing plant safety margins or 
the reliability of equipment assumed in the SAR to mitigate 
accidents. The proposed change does not adversely impact systems 
that respond to safely shut down the plant and maintain the plant in 
a safe shutdown condition. In addition, the proposed amendment will 
not result in plant operation in a configuration outside the design 
basis for an unacceptable period of time without implementation of 
appropriate compensatory measures.
    The risk evaluations for plant changes, in part as they relate 
to the potential for reducing a safety margin, were measured 
quantitatively for acceptability using the delta risk (i.e., 
[Delta]CDF and [Delta]LERF [large early release frequency]) criteria 
from Section 5.3.5, ``Acceptance Criteria,'' of NEI 04-02, as well 
as the guidance contained in RG 1.205. Engineering analyses, which 
may include engineering evaluations, probabilistic safety 
assessments, and fire modeling calculations, have been performed to 
demonstrate that the performance-based methods of NFPA 805 do not 
result in a significant reduction in the margin of safety. As such, 
the proposed changes are evaluated to ensure that risk and safety 
margins are kept within acceptable limits. Based on the above, the 
implementation of this amendment to transition the FPP at ANO-1 to 
one based on NFPA 805, in accordance with 10 CFR 50.48(c), will not 
significantly reduce a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, LA 70113.
    NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey
    Date of amendment request: April 30, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14127A435.
    Description of amendment request: The proposed amendment would 
revise Oyster Creek Nuclear Generating Station (OCNGS) Technical 
Specification (TS) 4.5 M., ``Shock Suppressors (Snubbers),'' to conform 
the TS to the revised OCNGS Snubber Inspection Program.
    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 would revise TS 4.5.M to conform the TS to 
the revised Snubber Inspection Program. Snubber examination, testing 
and service life

[[Page 38591]]

monitoring will continue to meet the requirements of 10 CFR 
50.55a(g). Snubber examination, testing and service life monitoring 
is not an initiator of any accident previously evaluated.
    Therefore, the probability of an accident previously evaluated 
is not significantly increased.
    Snubbers will continue to be demonstrated OPERABLE by 
performance of a program for examination, testing and service life 
monitoring in compliance with 10 CFR 50.55a or authorized 
alternatives. The proposed changes do not adversely affect plant 
operations, design functions or analyses that verify the capability 
of systems, structures, and components to perform their design 
functions.
    Therefore, the consequences of accidents previously evaluated 
are not significantly increased.
    Based on the above, these proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the amendment change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve any physical alteration of 
plant equipment. The proposed changes do not alter the method by 
which any safety-related system performs its function. As such, no 
new or different types of equipment will be installed, and the basic 
operation of installed equipment is unchanged. The methods governing 
plant operation and testing remain consistent with current safety 
analysis assumptions.
    Therefore, it is concluded that these 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 a margin of safety?
    Response: No.
    The proposed changes ensure snubber examination, testing and 
service life monitoring will continue to meet the requirements of 10 
CFR 50.55a(g). Snubbers will continue to be demonstrated OPERABLE by 
performance of a program for examination, testing and service life 
monitoring in compliance with 10 CFR 50.55a or authorized 
alternatives.
    Therefore, it is concluded that the proposed changes do not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, VP & Deputy General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Meena Khanna.
Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
    Date of amendment request: April 9, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14101A367.
    Description of amendment request: The proposed amendment would 
revise the Donald C. Cook Nuclear Plant, Units 1 and 2, technical 
specification (TS) 3.4.2, ``[Reactor Coolant System (RCS)] Pressure and 
Temperature (P/T) Limits,'' to address an issue regarding the 
applicability of TS Figures 3.4.3-1 ``Reactor Coolant System Pressure 
versus Temperature Limits--Heatup Limit, Criticality Limit, and Leak 
Test Limit (Applicable for service period up to 32 [Effective Full 
Power Years (EPFY)]'' and 3.4.3-2 ``Reactor Coolant System Pressure 
versus Temperature Limits--Various Cooldown Rates Limits (Applicable 
for service period up to 32 EFPY)'' during vacuum fill operations of 
the RCS.
    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 TS changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
There are no physical changes to the plant being introduced by the 
proposed changes to the heatup and cooldown limitation curves. The 
proposed changes do not modify the RCS pressure boundary. That is, 
there are no changes in operating pressure, materials, or seismic 
loading. The proposed changes do not adversely affect the integrity 
of the RCS pressure boundary such that its function in the control 
of radiological consequences is affected.
    Therefore, it is concluded that the proposed amendment does not 
involve a significant increase in the probability or the 
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 TS changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. No new modes of operation are introduced by the proposed 
changes. The proposed changes will not create any failure mode not 
bounded by previously evaluated accidents. Further, the proposed 
changes to the heatup and cooldown limitation curves do not affect 
any activities or equipment other than the RCS pressure boundary and 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Consequently, the proposed changes do not create the possibility 
of a new or different kind of accident, from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed TS changes do not involve a significant reduction 
in the margin of safety. The revised heatup and cooldown limitation 
curves and low-temperature overpressure protection limits are 
established in accordance with current regulations and the [American 
Society of Mechanical Engineers Boiler and Pressure Vessel (ASME 
B&PV)] Code 1995 edition with 1996 Addenda. These proposed changes 
are acceptable because the ASME B&PV Code maintains the margin of 
safety required by [Title 10 of the Code of Federal Regulations (10 
CFR)] 50.55(a). Because operation will be within these limits, the 
RCS materials will continue to behave in a non-brittle manner 
consistent with the original design bases.
    Therefore, the proposed amendment does not involve a significant 
reduction in 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: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: Robert D. Carlson.
Northern States Power Company--Minnesota (NSPM), Docket No. 50-263, 
Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota
    Date of amendment request: October 4, 2013, as supplemented by 
letter dated April 29, 2014. Publicly-available versions are in ADAMS 
under Accession Nos. ML13281A826 and ML14122A044, respectively.
    Description of amendment request: Following completion of an on-
site staffing analysis of the Emergency Response Organization, NSPM 
determined that the Radwaste Operator is no longer required to augment 
plant staff for performing repairs and corrective actions as prescribed 
in the MNGP Emergency Plan. The amendment proposes to remove the 
Radwaste Operator position as a 60-minute responder credited within the 
MNGP Emergency Plan.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 38592]]

issue of no significant hazards consideration, which is provided below.

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the Emergency Plan does not impact the 
function of plant structures, systems, or components (SSCs). The 
proposed change does not affect accident initiators or precursors, 
nor does it alter design assumptions. The proposed change does not 
alter or prevent the ability of the Emergency Response Organization 
to perform their intended functions to mitigate the consequences of 
an accident or event. This proposed change only removes a no longer 
credited position from the Emergency Plan.
    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 impact the accident analysis. The 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed), a change in 
the method of plant operation, or new operator actions. The proposed 
change does not introduce failure modes that could result in a new 
accident, and the change does not alter assumptions made in the 
safety analysis. This proposed change only removes a no longer 
credited position from the Emergency Plan. The proposed change, 
therefore, does not alter or prevent the ability of the Emergency 
Response Organization to perform their intended functions to 
mitigate the consequences of an accident or event.
    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.
    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 change is 
associated with the Emergency Plan staffing and does not impact 
operation of the plant or its response to transients or accidents. 
The change does not affect the Technical Specifications. The 
proposed change does not involve a change in the method of plant 
operation, and no accident analyses will be affected by the proposed 
change. Safety analysis acceptance criteria are not affected by this 
proposed change. The revised Emergency Plan will continue to provide 
the necessary response staff with the proposed change.
    Therefore, the proposed 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 request involves no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Robert D. Carlson.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska
    Date of amendment request: February 10, 2014, as supplemented by 
letter dated June 9, 2014. Publicly-available versions are in ADAMS 
under Accession Nos. ML14041A408 and ML14163A417, respectively.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) surveillance frequency for the 
pressurizer safety valves from a refueling frequency (i.e., 18 months 
+25 percent) to be consistent with the Inservice Testing Program. In 
addition, the proposed amendment would administratively change the 
format of the footnotes in TS Table 3-5, ``Minimum Frequencies for 
Equipment Tests.''
    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 requested change revises the performance interval of one TS 
surveillance requirement to be consistent with the Inservice Testing 
Program as stated in 10 CFR 50.55a(g)(5). The performance of the 
surveillance, or the failure to perform the surveillance, is not a 
precursor to an accident. Performing the surveillance or failing to 
perform the surveillance does not affect the probability of an 
accident. Even with the requested extension, the period during which 
the plant is in Modes 1 or 2 and the valves are required to be 
operable will be no longer than a typical operating cycle. Also, the 
proposed interval between tests will be consistent with the interval 
for this type of valve specified by the American Society of 
Mechanical Engineers (ASME) Code for Operation and Maintenance of 
Nuclear Power Plants (OM Code), 1998 Edition, through 2000 Addenda, 
Appendix I, frequency requirements for testing of pressure relief 
valves.
    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 change does not alter the physical design, safety 
limits, or safety analysis assumptions associated with the operation 
of the plant. Hence, the proposed change does not introduce any new 
accident initiators, nor does it reduce or adversely affect the 
capabilities of any plant structure or system in the performance of 
their safety function.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises the performance interval for one 
surveillance requirement to be consistent with the test interval for 
this type of valve specified by the ASME OM Code, 1998 Edition, 
through 2000 Addenda as required by 10 CFR 50.55a. This change does 
not alter any safety margins.
    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: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska
    Date of amendment request: March 31, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14090A417.
    Description of amendment request: The proposed amendment would 
change Technical Specification 2.5, Auxiliary Feedwater (AFW) system to 
allow a 7-day completion time for the turbine-driven AFW pump if the 
inoperability occurs following a refueling outage and if MODE 2 had not 
been entered.
    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

[[Page 38593]]

consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change to Technical Specification (TS) 2.5 would 
allow a seven day Completion Time for the turbine-driven Auxiliary 
Feedwater (AFW) pump if the inoperability occurs following a 
refueling outage, and if MODE 2 had not been entered. The note 
currently in TS 2.5 Applicability addresses the issue of allowing 
additional time to perform necessary testing to prove the 
operability of the turbine driven AFW pump following refueling as 
approved by the NRC in TS Amendment 127. This note does not 
specifically state that it is only allowed following refueling and 
does not restrict the time the plant can be in this condition. The 
proposed change will be more restrictive than the current TS since 
it will specifically state when it is allowed (following refueling) 
and for how long it is allowed.
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because: 1) the proposed amendment does not represent a change to 
the system design, 2) the proposed amendment does not prevent the 
safety function of the AFW system from being performed, since the 
other fully redundant essential train is required to be operable, 3) 
the proposed amendment does not alter, degrade, or prevent action 
described or assumed in any accident Updated Safety Analysis Report 
(USAR) from being performed since the other train of AFW is required 
to be operable, 4) the proposed amendment does not alter any 
assumptions previously made in evaluating radiological consequences, 
and 5) the proposed amendment does not affect the integrity of any 
fission product barrier. No other safety related equipment is 
affected by the proposed change.
    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 change does not alter the physical design, safety 
limits, or safety analysis assumptions associated with the operation 
of the plant. Hence, the proposed change does not introduce any new 
accident initiators, nor does it reduce or adversely affect the 
capabilities of any plant structure or system in the performance of 
their safety function.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not involve a significant reduction in a 
margin of safety. The proposed change to TS 2.5 would restrict for 
the turbine-driven AFW pump inoperability to a seven day Completion 
Time if the inoperability occurs following a refueling outage and 
prior to MODE 2 being entered. The current Note in TS 2.5 
Applicability does not require the turbine driven AFW pump to be 
operable until prior to entering MODE 2; therefore, the proposed 
change is more restrictive than current TS.
    The proposed change does not involve a significant reduction in 
a margin of safety because: (1) during a return to power operations 
following a refueling outage, decay heat is at its lowest levels, 
(2) the other AFW train is required to be operable, and (3) the 
motor-driven AFW train can provide sufficient flow to remove decay 
heat and cool the unit to shutdown cooling system entry conditions 
from power operations. This change does not alter any safety 
margins.
    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: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska
    Date of amendment request: April 25, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14118A435.
    Description of amendment request: The proposed amendment would 
revise Section 5.11, ``Structures Other Than Containment,'' and 
Appendix F, ``Classification of Structures and Equipment and Seismic 
Criteria,'' of the Fort Calhoun Station, Unit No. 1, Updated Safety 
Analysis Report. The changes would clarify the licensing and design 
basis to permit the use of seismic floor response spectra in analysis 
and design of seismic Class I structures and structural elements 
attached to structures.
    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.
    [T]his change to the Updated Safety Analysis Report (USAR) has 
no effect on the consequences of any accident, as it makes no 
physical changes to the plant. Since the Alternate Seismic Criteria 
and Methodologies (ASCM) floor response spectra (FRS) represent a 
refined version of the plant's original design basis, the design 
margins for any application utilizing the FRS will be maintained 
with respect to the design basis earthquake. Thus, the proposed 
amendment does not result in 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.
    [T]he change to the USAR does not change any accident analyses, 
does not make any physical changes to the plant, and does not change 
the way the plant is operated. The only change is to permit the 
utilization of the ASCM curves in the design and evaluation of 
structural applications. The curves themselves are based on the same 
earthquake as the plant's original design. Thus, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    [T]he ASCM FRS is based on the same earthquake as the plant's 
original design basis. The ASCM FRS are refined curves of the same 
design basis and thus, the design margins of any application or 
evaluation utilizing the ASCM FRS will be maintained with respect to 
the design basis earthquake. Thus, the proposed amendment does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska
    Date of amendment request: May 16, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14143A370.
    Description of amendment request: The proposed amendment would 
revise the Updated Safety Analysis Report (USAR) to allow pipe stress 
analysis of non-reactor coolant system safety-related piping to be 
performed in accordance with the American Society

[[Page 38594]]

of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, 
Section III, 1980 Edition as an alternative to current Code of record.
    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 the current licensing basis (CLB) allows 
the use of American Society of Mechanical Engineers (ASME) Boiler 
and Pressure Vessel (BPV) Code, Section III, 1980 Edition (no 
Addenda) as an alternative to the original Code of Record (i.e., 
United States of America Standards (USAS) B31.7 1968 (DRAFT) 
Edition) for the design and analysis of non-reactor coolant system 
(RCS) piping. The American National Standards Institute (ANSI) B31 
Code Committee has determined that:
    ``. . . piping that has been designed and constructed in 
accordance with Section III of the ASME Boiler and Pressure Vessel 
Code including addenda and applicable cases may be accepted as 
complying with the requirements of B31.7, 1969 and applicable 
addenda for the respective class of construction.''
    Although the ANSI B31 Code Committee statement refers to the 
B31.7, 1969 Edition, there are no significant differences between it 
and the B31.7 1968 (DRAFT) Edition. The change involves the 
substitution of one accepted piping Code for another and not a 
physical plant change. The Updated Safety Analysis Report (USAR) 
accident analysis assumes the proper functioning of safety systems 
in demonstrating the adequacy of the plant's design. This change 
does not alter the intended function of any plant equipment nor does 
it degrade or increase challenges to the performance of safety 
systems assumed to function in the accident analysis.
    The use of ASME BPV Code, Section III, 1980 Edition (no Addenda) 
analytical methods provides acceptable design results with no 
reduction in radiological barrier safety margin. Hence, there is no 
change in radiological barrier performance that would increase the 
dose to personnel onsite (10 CFR 20) or to the public at the site 
boundary (10 CFR 100).
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated in the USAR.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment provides the basis for the use of ASME 
BPV Code, Section III, 1980 Edition (no Addenda) for stress analysis 
of non-RCS safety-related piping. This approach will not introduce 
any methods or analytical techniques that could create the 
possibility of a new or different kind of accident. Application of a 
Code methodology does not create the possibility of a different kind 
of accident.
    The application of the ASME BPV Code, Section III, 1980 Edition 
(no Addenda) does not create any new unanalyzed interactions between 
systems or components. Piping systems will be analyzed in accordance 
with the Code, which is one part of the framework to establish the 
necessary design, fabrication, construction, testing, and 
performance requirements for structures, systems, and components 
important to safety. The proposed change to the CLB does not create 
a new failure mechanism or new accident initiator. The proposed 
amendment does not involve a change in methods governing the 
operation of plant systems or components.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated in the USAR.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The Fort Calhoun Station Technical Specifications (TS) ensure 
that the plant operates in a manner that will ensure acceptable 
levels of protection for the health and safety of the public. The 
Technical Specifications ensure that the available equipment and 
initial conditions for a Design Basis Accident (DBA) as defined in 
the USAR meet the assumptions in the accident analysis contained in 
the USAR. The plant safety margins are addressed in the Technical 
Specification Bases and the USAR.
    This proposed amendment revises the CLB to allow the use of ASME 
BPV Code, Section III, 1980 Edition (no Addenda) for stress analysis 
of non-RCS safety-related piping. No changes are being made to the 
physical plant. The use of the ASME BPV Code, Section III, 1980 
Edition (no Addenda) does not change, revise, or otherwise affect 
the current Technical Specifications (TS) or TS Bases. Incorporation 
of the ASME BPV Code, Section III, 1980 Edition (no Addenda) into 
the FCS CLB will not affect the current plant design parameters or 
TS Limiting Conditions for Operation (LCO).
    The proposed change does not modify, change, revise, or 
otherwise affect any current calculations concerning the plant 
accident analysis or supporting basis for which the TSs, TS Bases, 
or USAR safety margins were established. Therefore, the proposed 
amendment does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.
ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power 
Station (ZNPS), Units 1 and 2, Lake County, Illinois
    Date of amendment request: March 17, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14078A049.
    Description of amendment request: The proposed amendments would 
amend licenses DPR-39 and DPR-48 and revise the Zion Technical 
Specifications (TS) to reflect the removal of all the spent fuel from 
the Zion spent fuel pool. The proposed changes to both Facility 
Operating Licenses modify Section 2.C.(6) to specify the ZNPS 
Independent Spent Fuel Storage Installation Physical Security Plan 
(ISFSI), eliminate Section 2.C.(7) Spent Fuel Pool Modification, and 
eliminate Section 2.C.(16), related to the single-failure proof fuel 
building crane. The proposed changes to the TS eliminate provisions of 
the specifications applicable to spent fuel stored in the spent fuel 
pool and relocate the remaining TS administrative requirements to the 
Quality Assurance Project Plan. These changes are proposed pursuant to 
the criteria contained in 10 CFR 50.36 and in accordance with the 
recommendations contained in the U.S. Nuclear Regulatory Commission's 
(NRC) Administrative Letter 95-06. The proposed changes will result in 
a TS that will be applicable to the ZNPS once the last spent fuel 
assembly has been removed from the spent fuel pool and placed at the 
ISFSI.
    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?
    The proposed changes (deletion of operational requirements and 
certain design requirements) reflect the complete transfer of the 
spent fuel from the spent fuel pool to the ISFSI. Design basis 
accidents related to the spent fuel pool are discussed in the ZNPS 
Defueled Safety Analysis Report (DSAR) Chapter 5. These postulated 
accidents are predicated on spent fuel being stored in the spent 
fuel pool. With the removal of the spent fuel from the spent fuel 
pool, there are no remaining spent fuel assemblies to be monitored 
and there are no credible accidents that require the actions of a 
Certified Fuel Handler, Shift Supervisor, or a Non-certified 
Operator to prevent occurrence or mitigate the consequences of an 
accident.
    In addition, the ZNPS DSAR Chapter 5 also provides analyses of 
accidents as result of

[[Page 38595]]

decommissioning with the bounding consequences resulting from the 
failure of a High Integrity Container (HIC) containing dewatered 
radioactive demineralizer resin.
    The proposed changes do not have an adverse impact on the 
remaining decommissioning activities or any decommissioning related 
postulated accident consequences.
    The proposed changes related to the relocation of certain 
administrative requirements do not affect operating procedures or 
administrative controls that have the function of preventing or 
mitigating any remaining decommissioning design basis accidents. In 
addition, these proposed changes are consistent with the guidance of 
the NRC's Administrative Letter 95-06.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    (2) Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes eliminate the operational requirements and 
certain design requirements associated with the storage of the spent 
fuel in the spent fuel pool, and relocate certain administrative 
controls to the Quality Assurance Program Plan.
    With the complete removal of the spent fuel from the spent fuel 
pool and transfer to the ISFSI, there are no spent fuel assemblies 
that remain at the plant and the potential for fuel related 
accidents is removed. The proposed changes do not introduce any new 
failure modes. Therefore, the proposed changes will 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?
    The design basis and accident assumptions within the ZNPS DSAR 
and the TS relating to spent fuel are no longer applicable. The 
proposed changes do not affect remaining plant operations, systems, 
or components supporting decommissioning activities. In addition, 
the proposed changes do not result in a change in initial 
conditions, system response time, or in any other parameter 
affecting the course of the remaining decommissioning activity 
accident analysis. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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 
``Accessing Information and Submitting Comments'' section of this 
document.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 
1, 2, and 3, Maricopa County, Arizona
    Date of application for amendment: December 26, 2012, as 
supplemented by letter dated August 26, 2013.
    Brief description of amendment: The amendments adopt Technical 
Specifications Task Force (TSTF) change traveler TSTF-500, Revision 2, 
``DC Electrical Rewrite--Update to TSTF-360.'' The amendments revised 
TS requirements related to direct current (DC) electrical systems in TS 
limiting condition for operation (LCO) 3.8.4, ``DC Sources--
Operating,'' LCO 3.8.5, ``DC Sources--Shutdown,'' and LCO 3.8.6, 
``Battery Parameters.'' A new ``Battery Monitoring and Maintenance 
Program'' was added to Section 5.5, ``Programs and Manuals.''
    Date of issuance: June 25, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment No.: Unit 1--193; Unit 2--193; Unit 3--193. A publicly-
available version is in ADAMS under Accession No. ML14115A045; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
The amendments revised the Operating Licenses and Technical 
Specifications.
    Date of initial notice in Federal Register: March 4, 2013 (78 FR 
14129). The supplement dated August 26, 2013, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 25, 2014.
    No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc. (DEK), Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin
    Date of application for amendment request: April 16, 2013, as 
supplemented by letters dated September 5, 2013, October 14, 2013, and 
March 19, 2014.
    Brief description of amendment: The amendment revised the Renewed 
Facility Operating License by deleting a license condition associated 
with license renewal and adding a license condition related to spent 
fuel pool storage rack boron absorber surveillance.
    Date of issuance: June 23, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 213. A publicly-available version is in ADAMS under 
Accession No. ML14008A297; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-43: Amendment revised 
the Renewed Facility Operating License.

[[Page 38596]]

    Date of initial notice in Federal Register: August 20, 2013 (78 FR 
51223). The supplemental letters dated September 5, 2013, October 14, 
2013, and March 19, 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 23, 2014.
    No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc. (DEK), Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin
    Date of application for amendment request: May 29, 2013, as 
supplemented by letters dated September 23, October 15, October 17, 
October 31, and November 7, 2013, and letters dated January 7, 2014, 
and March 13, 2014.
    Brief description of amendment: The amendment revised the Renewed 
Facility Operating License Technical Specifications (TSs) to permit 
fuel handling activities consistent with the permanently shutdown and 
defueled condition of the facility. Specifically, in its March 13, 
2014, supplemental letter DEK stated that it had accelerated the 
schedule to transfer spent fuel from the spent fuel pool to the 
independent spent fuel storage installation (ISFSI). Under its new 
schedule, DEK plans to begin activities to support spent fuel transfer 
to the ISFSI by July 1, 2014. Based on its new schedule, DEK requested 
expedited review and partial approval of the deletion of certain TSs 
currently required for movement of irradiated fuel assemblies. If not 
amended, the affected TSs would require restoring operability of 
certain equipment during spent fuel handling activities that are no 
longer needed for accident mitigation.
    The NRC staff has issued a partial approval of the original May 29, 
2013, amendment request as supplemented, to permit fuel handling 
activities in accordance with DEK's request in its March 13, 2014, 
submittal. The staff continues to review the remaining license 
condition and technical specification changes requested in DEK's May 
29, 2013, submittal as supplemented, that were not addressed in this 
amendment.
    Date of issuance: June 9, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 212. A publicly-available version is in ADAMS under 
Accession No. ML14111A234; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-43: Amendment revised 
the Renewed Facility Operating License.
    Date of initial notice in Federal Register: August 20, 2013 (78 FR 
51224). The supplemental letters dated September 23, October 15, 
October 17, October 31, and November 7, 2013, January 7, 2014, and 
March 13, 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 9, 2014.
    No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana
    Date of amendment request: February 7, 2013, as supplemented by 
letter dated January 16, 2014.
    Brief description of amendment: The amendment revised the River 
Bend Station, Unit 1 (RBS) Technical Specification (TS) 3.8.4, ``DC 
[Direct Current] Sources--Operating,'' Surveillance Requirements 
3.8.4.2 and 3.8.4.5. The change is the result of the licensee's 
determination that the total battery capacity would possibly be 
insufficient to supply the required load to the DC system if each of 
the battery-to-battery connections were to reach the individual 
resistance limits. The changes to the Surveillance Requirements added 
new acceptance criteria to address the possible non-conservative 
conditions when the battery connection resistances are at maximum TS 
values.
    Date of issuance: June 18, 2014.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 181. A publicly-available version is in ADAMS under 
Accession No. ML14136A008; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 30, 2013 (78 FR 
25312). The supplemental letter dated January 16, 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 18, 2014.
    No significant hazards consideration comments received: No.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
    Date of application for amendment: September 14, 2012, as 
supplemented by letters dated January 29, February 14, May 30, and 
October 22, 2013, and March 11, 2014.
    Brief description of amendment: The amendments revised the 
operating licenses and Technical Specifications (TSs) to remove 
completed and satisfied license conditions, revised TS 5.5.1 to remove 
related conditions, corrected inadvertent errors, updated references to 
the Physical Security Plan, and made editorial changes to the operating 
licenses and TSs.
    Date of issuance: June 13, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 260 and 255. A publicly-available version is in 
ADAMS under Accession No. ML13329A092; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the licenses and the TSs.
    Date of initial notice in Federal Register: January 8, 2013 (78 FR 
1271), and April 16, 2013 78 FR 22569). The submittal dated January 29, 
2013, expanded the scope of the application dated September 14, 2012, 
and the application was renoticed April 16, 2013. The supplements dated 
February 14, May 30, and October 22, 2013, and March 11, 2014, provided 
additional information that clarified the application, did not expand 
the scope of the submittal dated January 29, 2013, as noticed, and did 
not change the staff's proposed no significant hazards consideration 
determinations published on January 8, 2013, and April 16, 2013. The 
supplement dated March 11, 2014, limited the scope of the supplement 
dated January 29, 2013, by deleting the

[[Page 38597]]

proposed change to TS Figure 3.1-2, ``Boric Acid Tank Minimum Volume.''
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 13, 2014.
    No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska
    Date of application for amendment: September 28, 2011, as 
supplemented by letters dated December 19 and December 22, 2011; March 
20, July 24, August 24, and September 27, 2012; April 23, May 21, July 
29, September 12, October 11, November 4, November 11, and December 18, 
2013; and January 24, February 28, April 10, and June 11, 2014.
    Brief description of amendment: The amendment transitions the Fort 
Calhoun Station 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 16, 2014.
    Effective date: As of its date of issuance and shall be implemented 
by 12 months from the date of issuance.
    Amendment No.: 275. A publicly-available version is in ADAMS under 
Accession No. ML14098A092; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 10, 2012 (77 FR 
21598). The supplements dated March 20, July 24, August 24, and 
September 27, 2012; April 23, May 21, July 29, September 12, October 
11, November 4, November 11, and December 18, 2013; and January 24, 
February 28, April 10, and June 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 16, 2014.
    No significant hazards consideration comments received: No.
Southern Nuclear Operating Company Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: December 20, 2013.
    Brief description of amendment: The amendment revises the plant's 
emergency plan. In conjunction with the new license condition, the 
amendment complies with the established regulatory changes set forth in 
``Enhancements to Emergency Preparedness Regulations,'' published in 
the Federal Register on November 23, 2011 (76 FR 72560). Specifically, 
the license amendment changes on-shift staffing analysis and the 
changes to the emergency plan address evacuation time estimates. The 
design, construction and operation of the plant are not affected by 
this license amendment and license condition.
    Date of issuance: May 30, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 20. A publicly-available version is in ADAMS under 
Accession No. ML14118A252; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: February 4, 2014, (79 
FR 6643).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 30, 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: December 13, 2012, as 
supplemented by letters dated June 11, 2013, and January 16 and April 
9, 2014.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.7.9, ``Ultimate Heat Sink (UHS),'' to incorporate 
more restrictive UHS level and pond temperature limits which are 
specified in Surveillance Requirements (SRs) 3.7.9.1 and 3.7.9.2, 
respectively. In addition, new SR 3.7.9.4 is added to verify that the 
UHS cooling tower fans respond appropriately to automatic start 
signals.
    Date of issuance: June 17, 2014.
    Effective date: As of its date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment No.: 208. A publicly-available version is in ADAMS under 
Accession No. ML14149A164; 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: March 4, 2013 (78 FR 
14138). The supplements dated June 11, 2013, and January 16 and April 
9, 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 17, 2014.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 27th day of June, 2014.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2014-15770 Filed 7-7-14; 8:45 am]
BILLING CODE 7590-01-P