Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 67402-67418 [2013-27025]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 78, Number 218 (Tuesday, November 12, 2013)]
[Notices]
[Pages 67402-67418]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-27025]


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

NUCLEAR REGULATORY COMMISSION

[NRC-2013-0249]


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

Background

    Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission 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

[[Page 67403]]

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 October 17 to October 30, 2013. The last 
biweekly notice was published on October 29, 2013 (78 FR 64541).

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-2013-0249. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: 3WFN-06-A44MP, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2013-0249 when contacting the NRC 
about the availability of information regarding this document. You may 
access publicly-available information related to this action by the 
following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0249.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at https://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to pdr.resource@nrc.gov. Documents may be viewed in 
ADAMS by performing a search on the document date and docket number.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2013-0249 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information in comment submissions that you do not want to be publicly 
disclosed. The NRC posts all comment submissions at https://www.regulations.gov as well as entering the comment submissions into 
ADAMS, and the NRC does not edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information in their comment submissions 
that they do not want to be publicly disclosed. Your request should 
state that the NRC will not edit comment submissions to remove such 
information before making the comment submissions available to the 
public or entering the comment submissions into ADAMS.

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses and Combined Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in section 50.92 of Title 10 of the Code 
of Federal Regulations (10 CFR), this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Rules of 
Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested person(s) should consult a current copy of 10 CFR 2.309, 
which is available at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. NRC regulations are accessible electronically from the NRC 
Library on the NRC 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

[[Page 67404]]

should specifically explain the reasons why intervention should be 
permitted with particular reference to the following general 
requirements: (1) The name, address, and telephone number of the 
requestor or petitioner; (2) the nature of the requestor's/petitioner's 
right under the Act to be made a party to the proceeding; (3) the 
nature and extent of the requestor's/petitioner's property, financial, 
or other interest in the proceeding; and (4) the possible effect of any 
decision or order which may be entered in the proceeding on the 
requestor's/petitioner's interest. The petition must also identify the 
specific contentions which the requestor/petitioner seeks to have 
litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the Internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital information (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through Electronic Information Exchange System, users 
will be required to install a Web browser plug-in from the NRC 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 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 Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by email at 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an

[[Page 67405]]

exemption request, in accordance with 10 CFR 2.302(g), with their 
initial paper filing requesting authorization to continue to submit 
documents in paper format. Such filings must be submitted by: (1) First 
class mail addressed to the Office of the Secretary of the Commission, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
Attention: Rulemaking and Adjudications Staff; or (2) courier, express 
mail, or expedited delivery service to the Office of the Secretary, 
Sixteenth Floor, One White Flint North, 11555 Rockville Pike, 
Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications 
Staff. Participants filing a document in this manner are responsible 
for serving the document on all other participants. Filing is 
considered complete by first-class mail as of the time of deposit in 
the mail, or by courier, express mail, or expedited delivery service 
upon depositing the document with the provider of the service. A 
presiding officer, having granted an exemption request from using E-
Filing, may require a participant or party to use E-Filing if the 
presiding officer subsequently determines that the reason for granting 
the exemption from use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the following three 
factors in 10 CFR 2.309(c)(1): (i) The information upon which the 
filing is based was not previously available; (ii) the information upon 
which the filing is based is materially different from information 
previously available; and (iii) the filing has been submitted in a 
timely fashion based on the availability of the subsequent information.
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through ADAMS in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to pdr.resource@nrc.gov.

Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear 
Power Plant, Unit No. 1, Wake County, North Carolina

    Date of amendment request: November 29, 2012, as supplemented by 
letter dated January 3, 2013.
    Description of amendment request: This is being re-noticed in its 
entirety due to an error in the amendment description of the notice 
published in the Federal Register on February 19, 2013 (78 FR 11691). 
The proposed amendment would revise the degraded voltage time delay 
values in Technical Specification (TS) Table 3.3-4. In conjunction with 
planned plant modifications and reanalysis of the final safety analysis 
design basis large break loss of coolant accident (LOCA), the revisions 
would resolve a nonconservative TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the Technical Specifications (TS) 
Table 3.3-4, Functional Unit 9.b. Loss of Offsite Power, 6.9 kV 
(kilovolt) Emergency Bus Undervoltage--Secondary time delay values. 
The Loss of Offsite Power, 6.9 kV (kilovolt) Emergency Bus 
Undervoltage--Secondary instrumentation functions are not initiators 
to any accident previously evaluated. As such, the probability of an 
accident previously evaluated is not increased. The revised values 
continue to provide reasonable assurance that the Loss of Offsite 
Power, 6.9 kV (kilovolt) Emergency Bus Undervoltage--Secondary 
function will continue to perform its intended safety functions. As 
a result, the proposed change will not increase the consequences of 
an accident previously evaluated.
    Concurrent with this proposed change, the Harris Nuclear Plant 
is revising its large break loss of coolant accident analysis. The 
revised analysis will be evaluated in accordance with 10 CFR 50.59 
to confirm that a change to the technical specifications 
incorporated in the license is not required, and the change does not 
meet any of the criteria in Paragraph (c)(2) of that regulation. The 
revised analysis will employ the plant-specific methodology ANP-
3011(P), Harris Nuclear Plant, Unit 1, Realistic Large Break LOCA 
Analysis, Revision 1, as approved by NRC Safety Evaluation dated May 
30, 2012.
    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 revises the TS Table 3.3-4, Functional Unit 
9.b. Loss of Offsite Power, 6.9 kV (kilovolt) Emergency Bus 
Undervoltage--Secondary time delay values. No new operational 
conditions beyond those currently allowed are introduced. This 
change is consistent with the safety analyses assumptions and 
current plant operating practices. This simply corrects the setpoint 
consistent with the accident analyses and therefore cannot create 
the possibility of a new or different kind of accident from any 
previously evaluated accident.
    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 
the margin of safety?
    Response: No.
    The proposed change revises the TS Table 3.3-4, Functional Unit 
9.b. Loss of Offsite Power, 6.9 kV (kilovolt) Emergency Bus 
Undervoltage--Secondary time delay values. This proposed change 
implements a reduced time delay to isolate safety buses from offsite 
power if a Loss of Coolant Accident were to occur coincident with a 
sustained degraded voltage condition. This provides improved margin 
to ensure that emergency core cooling system pumps inject water into 
the reactor vessel within the time assumed and evaluated in the 
accident analysis.
    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

[[Page 67406]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: David T. Conley, Manager--Senior Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Jessie F. Quichocho.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida.

    Date of amendment request: July 26, 2013, as supplemented by letter 
dated October 16, 2013.
    Description of amendment request: The proposed amendment would 
align St. Lucie TSs with NUREG-1432, Revision 4, Combustion Engineering 
Plants Standard Technical Specifications (STSs) describing the 
Administrative Controls requirements for the Responsibility and 
Organization, which includes Onsite and Offsite Organizations and the 
Unit Staff. The proposed amendment will revise TSs 6.1, Responsibility 
and 6.2, Organization to be consistent with STSs 5.1 Responsibility and 
5.2 Organization, which directly reference the requirements in 10 CFR 
50.54(m). The current Units 1 and 2 TSs 6.1 and 6.2 use custom language 
to define the requirements of the regulation.
    Basis for proposed no significant hazards consideration (NSHC) 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of NSHC, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes involve reformatting, renumbering, and 
rewording. The revisions have no technical implications with respect 
to the station organization, responsibilities, or unit staffing 
requirements. The changes do not affect the minimum shift complement 
in any mode of operation nor decrease the effectiveness of the shift 
personnel. The proposed changes are minor or editorial in nature and 
will not result in any significant increase in the probability of 
consequences of an accident as previously evaluated, as the proposed 
TS changes are consistent with the NUREG-1432, Combustion 
Engineering Plant Standard Technical Specifications. Further, the 
proposed changes do not introduce additional risk or greater 
potential for consequences of an accident that has not previously 
been evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are minor or editorial in nature. The 
proposed changes do not involve a physical modification of the plant 
or methods governing normal plant operation. No new or different 
type of equipment will be installed. The proposed changes will not 
introduce new failure modes/effects that could lead to an accident 
not previously analyzed. The proposed changes will not impose any 
new or change existing requirements that are not consistent with 
NUREG-1432, Combustion Engineering Plant Standard Technical 
Specifications.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes involve reformatting, renumbering, and 
rewording. The revisions have no technical implications with respect 
to the station organization, responsibilities, or unit staffing 
requirements. The changes do not affect the minimum shift complement 
in any mode of operation nor decrease the effectiveness of the shift 
personnel. The proposed changes will not involve a significant 
reduction in a margin of safety in that the changes are minor or 
editorial in nature. No plant equipment or accident analyses will be 
affected. Additionally, the proposed changes will not relax any 
criteria used to establish safety limits, safety system settings, or 
the bases for any limiting conditions for operation. Safety analysis 
acceptance criteria are not affected. Plant operation will continue 
within the design basis. The proposed changes do not adversely 
affect systems that respond to safely shutdown the plant, and 
maintain the plant in a safe shutdown condition. Consequently, the 
proposed changes do not involve a significant reduction in a margin 
of safety.

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

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

    Date of amendment request: April 25, 2013, as supplemented on 
September 4, 2013.
    Description of amendment request: The proposed license amendment 
request would revise certain requirements from Section 5, 
``Administrative Controls,'' of the CR-3 Improved Technical 
Specifications (ITSs). The revisions would revise and remove certain 
requirements in Section 5.1 ``Responsibility,'' 5.2 ``Organization,'' 
5.6 ``Procedures, Programs and Manuals,'' 5.7 ``Reporting 
Requirements,'' and 5.8 ``High Radiation Area,'' that are no longer 
applicable to CR-3 in the permanently defueled condition. The September 
4, 2013, supplement supersedes the April 25, 2013, application, and 
replaces it in its entirety. In addition, the proposed no significant 
hazards consideration determination in the basis section below corrects 
a typographical numbering error for TS 5.2.1.b (the section was 
incorrectly labeled ``5.1.2.b'' in Section 4.1 of Attachment B of the 
September 4, 2013, application).
    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 for each proposed change, which is presented below:

    A. ITS Section 5.1.1:
    This section defines the responsible position for overall unit 
operation and for approval of each proposed test, experiment, or 
modification to systems or equipment that affect stored nuclear fuel 
and fuel handling. The responsible position title is changed from 
the Plant General Manager to the Plant Manager.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The change reflects that the remaining credible accident is a 
fuel handling accident or loss of spent fuel cooling. The change in 
the position title of the responsible person is administrative and 
cannot increase the probability or consequences of a fuel handling 
accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change reflects an organizational change to transition from 
an operating plant to a permanently defueled plant. Such an 
administrative change cannot create a new or different kind of 
accident.

[[Page 67407]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The position title proposed here does not involve any physical 
plant limits or parameters and therefore cannot affect any margin of 
safety.
    B. ITS Section 5.1.2:
    This section identifies the responsibilities for the control 
room command function associated with Modes of plant operation, and 
is based on personnel positions and qualifications for an operating 
plant. It identifies the need for a delegation of authority for 
command in an operating plant when the principal assignee leaves the 
control room.
    This section is being changed to eliminate the MODE dependency 
for this function and personnel qualifications associated with an 
operating plant. The proposed change establishes the Shift 
Supervisor as having command of the shift.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This is a change to the requirements for control room staffing. 
In a permanently defueled plant, the fuel handling building accident 
is the only credible accident previously evaluated. This action 
cannot increase the probability or consequences of a fuel handling 
accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The changes proposed here for control room staffing cannot 
create a new or different kind of accident since they do not change 
the function of any plant structures, systems, or components.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The changes proposed here for control room staffing do not 
directly involve any limits or parameters and therefore cannot 
affect any margin of safety.
    C. ITS Section 5.2.1.a:
    The introduction to this section identifies that organizational 
positions are established that are responsible for the safety of the 
nuclear plant.
    This is changed to require that positions be established that 
are responsible for the safe storage and handling of nuclear fuel. 
This change removes the implication that CR-3 can return to 
operation.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change in the description of functional responsibility of 
organizational positions places emphasis on the safe storage and 
handling of nuclear fuel. This focus on their principal 
responsibility cannot increase the probability or consequences of a 
fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change in the description of functional responsibility of 
organizational positions cannot create a new or different kind of 
accident since they do not change the function of any plant 
structures, systems, or components.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any physical limits or 
parameters and therefore cannot affect any margin of safety.
    D. ITS Section 5.2.1.b:
    This section identifies the organizational position responsible 
for overall nuclear plant safety, for the safe operation of the 
plant, and for control of activities necessary for the safe 
operation and maintenance of the plant.
    This section is being changed to recognize that the safety 
concerns for a permanently defueled plant are for the safe storage 
and handling of nuclear fuel. It changes responsibility for overall 
safety for storage and handling of nuclear fuel to the 
Decommissioning Director. It changes responsibility for control over 
onsite activities necessary for safe handling and storage of nuclear 
fuel to the Plant Manager.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change in the description of functional responsibility of 
organizational positions places emphasis on the safe storage and 
handling of nuclear fuel. This focus on their principal 
responsibility cannot increase the probability or consequences of a 
fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change in the description of functional responsibility of 
organizational positions cannot create a new or different kind of 
accident since they do not change the function of any plant 
structures, systems, or components.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any physical limits or 
parameters and therefore cannot affect any margin of safety.
    E. ITS Section 5.2.1.c:
    This paragraph addresses the requirement for organizational 
independence of the operations, health physics, and quality 
assurance personnel from operating pressures.
    This is changed to replace ``operating staff'' with ``Certified 
Fuel Handlers,'' and to replace ``their independence from operating 
pressures'' to ``their ability to perform their assigned 
functions.''
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change continues to ensure that personnel in specifically 
identified positions retain independence from organizational 
pressures and will not increase the probability or occurrence of a 
fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    F. ITS Section 5.2.2.a:
    This paragraph addresses that one auxiliary nuclear operator 
must be assigned to the operating shift whenever fuel is in the 
reactor.
    Since this can never occur again at CR-3, the minimum 
requirement is changed to a minimum crew compliment of one Shift 
Supervisor and one Non-certified Operator.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change, in conjunction with new paragraph 5.2.2.f, 
continues to ensure that personnel trained and qualified for the 
safe handling and storage of nuclear fuel are onsite. This cannot 
increase the probability or consequences of a fuel handling 
accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    G. ITS Section 5.2.2.b:
    This paragraph addresses the conditions under which the minimum 
shift compliment may be reduced. It contains a reference to 10 CFR 
50.54(m) which establishes the minimum requirements for a licensed 
operating staff for facility operation.
    This reference is removed since CR-3 will not return to 
operation in the future, and the requirement for licensed operating 
personnel will no longer be required to protect public health and 
safety.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change continues to ensure that the minimum shift 
compliment of qualified

[[Page 67408]]

personnel will not be decreased for more than a limited period. It 
removes the qualification requirements for personnel who are capable 
of responding to operating plant transients and accidents. This does 
not involve an increase in the probability or consequences of a fuel 
handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    H. ITS Section 5.2.2.c:
    This paragraph establishes the requirement for one licensed 
Reactor Operator to be in the control room when fuel is in the 
reactor and for one Senior Reactor Operator to be in the control 
room during operating Modes 1-4.
    The change establishes the requirements for either a Non-
certified operator or Certified Fuel handler to be in the control 
room when fuel is stored in the pools.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change continues to ensure that personnel trained and 
qualified for the handling and storage of nuclear fuel man the 
control room. This cannot increase the probability or consequences 
of a fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    I. ITS Section 5.2.2.d:
    This paragraph established the requirement for a person 
qualified in Radiation Protection procedures to be onsite when fuel 
is in the reactor.
    This paragraph is revised to require a person qualified in 
Radiation Protection procedures to be onsite during fuel handling 
operations and during movement of heavy loads over the fuel storage 
racks.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This is an administrative change that cannot affect the 
probability of a fuel handling accident. The consequences of a fuel 
handling accident are governed by the characteristics of the fuel 
element and are not affected by the presence or absence of radiation 
protection trained personnel.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    J. ITS Section 5.2.2.e (New):
    A new paragraph is added to establish the requirement for having 
oversight of fuel handling operations to be performed by a Certified 
Fuel Handler.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Certified Fuel Handlers are specifically trained and qualified 
to safely handle irradiated fuel. Applying these qualifications to 
fuel movement ensures that the probability or consequences of a fuel 
handling accident are not increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    K. ITS Section 5.2.2.f (New):
    A new paragraph is added to establish that the Shift Supervisor 
must be a Certified Fuel Handler.
    In the permanently defueled plant, the Certified Fuel Handler is 
the senior position on the operating crew. It is not necessary for 
the Shift Supervisor to hold a Senior Reactor Operator license if 
the plant cannot operate to generate power.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Certified Fuel Handlers are specifically trained and qualified 
to safely handle irradiated fuel. Applying these qualifications to 
the supervision of fuel movement ensures that the probability or 
consequences of a fuel handling accident are not increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    L. ITS Section 5.3.1:
    This paragraph is changed to remove the requirements for the 
Shift Technical Advisor since that position is only required for a 
plant authorized for power operations.
    The paragraph retains the previous requirements for the 
personnel filling unit staff positions meet or exceed the minimum 
qualifications of ANSI [American National Standard Institute] N18.1, 
1971, and the Radiation Protection Manager meet or exceed the 
qualifications of Regulatory Guide 1.8, September 1975.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Shift Technical Advisor position was established to assist 
the control room operating personnel to diagnose the cause and 
advise on the response to operating transients and accidents. The 
absence of a staff member with those qualifications does not change 
the probability or consequences of a fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any physical equipment 
limits or parameters and therefore cannot affect any margin of 
safety.
    M. ITS Section 5.3.2:
    This new paragraph is added to identify that responsibility for 
the training and retraining of Certified Fuel Handlers is assigned 
to the Plant Manager.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This section recognizes the importance of establishing and 
maintaining Certified Fuel Handler qualifications and assigns a 
manager responsibility for this program. Training and retraining 
Certified Fuel Handlers specifically trained to safely handle 
nuclear fuel will not increase the probability or consequences of a 
fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of

[[Page 67409]]

accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any physical limits or 
parameters and therefore cannot affect any margin of safety.
    N. ITS Section 5.6.1.1.a:
    This section states the requirement for procedures to be 
established, implemented and maintained covering various plant 
activities.
    The scope is reduced to procedures applicable to the safe 
handling and storage of nuclear fuel.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The procedures necessary for the safe handling of nuclear fuel 
are included in the group of procedures applicable to the safe 
storage of nuclear fuel. With these procedures in effect for fuel 
handling, the probability or consequences of a fuel handling 
accident will not be increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The applicable procedures for the safe storage of nuclear fuel 
will direct the correct use of fuel handling equipment. These 
procedures are currently in place and have been used effectively for 
the safe handling of fuel. These procedures will not direct the use 
of plant structures, systems, or components in a different manner, 
therefore, they cannot create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    O. ITS Section 5.6.2.3:
    In this section, the authority for approval of changes to the 
Offsite Dose Calculation Manual (ODCM) is changed from the Plant 
General Manager to the Plant Manager consistent with the position 
title change in 5.1.1.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This is a change to the requirements for the position 
responsible for approving ODCM changes. In a permanently defueled 
plant, the fuel handling accident is the only credible accident 
previously evaluated. This action cannot increase the probability or 
consequences of a fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The change proposed here, identifying a different position 
responsible for ODCM change approval, cannot create a new or 
different kind of accident since this does not change the function 
of any plant structures, systems, or components.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The changes proposed here for ODCM approval do not directly 
involve any limits or parameters for operating systems and therefore 
cannot affect any margin of safety.
    P. ITS Section 5.6.2.4: Primary Coolant Sources Outside 
Containment:
    This program was established to minimize leakage from portions 
of systems outside containment that could contain highly radioactive 
fluids during a serious transient or accident.
    The program is being eliminated.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The fuel handling accident is the only credible accident for a 
permanently defueled plant. This change eliminates an inspection 
program that is no longer necessary to limit the consequences of 
operating transients and accidents. This change cannot increase the 
probability or consequences of the fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    Q. ITS Section 5.6.2.5: Component Cyclic or Transient Limit:
    This program provided controls to track cyclic and transient 
occurrences to ensure that components were maintained within their 
design limits.
    This program is being eliminated.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Eliminating an administrative event tracking program cannot 
increase the probability of a fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Eliminating an administrative event tracking program cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    R. ITS Section 5.6.2.8: Inservice Inspection Program:
    This program required periodic inspections, examinations, and 
tests of plant pressure boundary components to ensure their 
continued integrity for power operation.
    This program is being eliminated.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Inservice Inspection Program does not apply to nuclear fuel 
or fuel handling equipment. Therefore eliminating this program 
cannot increase the probability or occurrence of a fuel handling 
accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    For an operating plant the Inservice Inspection Program provided 
confidence that plant systems that were either a potential source of 
an accident or transient or served to mitigate events continued to 
meet their physical requirements. For a permanently shutdown plant, 
no transient, or accident can occur, so ending this inspection 
program cannot affect any margin of safety.
    S. ITS Section 5.6.2.9: Inservice Testing Program:
    This program required periodic testing of ASME Code Class 1, 2, 
and 3, components including applicable supports in accordance with 
the ASME Operations and Maintenance (OM) Code.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Inservice Testing Program does not apply to nuclear fuel or 
fuel handling equipment. Therefore eliminating this program cannot 
increase the probability or occurrence of a fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    For an operating plant, the Inservice Testing Program provided 
confidence that plant components that were required for safe

[[Page 67410]]

shutdown would perform as expected. For a permanently shutdown 
plant, the transients or accidents that would require safe shutdown 
equipment cannot occur, so ending this testing program cannot affect 
any margin of safety.
    T. ITS Section 5.6.2.10: Steam Generator (OTSG) Program:
    The Steam Generator Program established and implemented 
practices to ensure that OTSG tube integrity was maintained.
    This program is being eliminated.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The condition of the steam generator tubes inside the 
containment has no effect on fuel handling in the auxiliary building 
within the spent fuel pools. Therefore, eliminating the program 
cannot increase the probability or occurrence of a fuel handling 
accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The CR-3 steam generators will remain out of service until 
removed from the plant. In this state, the condition of the steam 
generator tubes is immaterial and cannot create a new or different 
kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    U. ITS Section 5.6.2.11: Secondary Water Chemistry Program:
    This program provided controls for monitoring secondary water 
chemistry to inhibit steam generator tube degradation and low 
pressure turbine disc stress corrosion cracking.
    This program is being eliminated.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The secondary piping systems do not interconnect with the fuel 
cooling or fuel handling systems. Therefore, eliminating the 
Secondary Water Chemistry Program cannot increase the probability or 
occurrence of a fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The components this program was intended to protect will no 
longer function for power production. Therefore, eliminating this 
program cannot affect any margin of safety.
    V. ITS Section 5.6.2.13: Explosive Gas and Storage Tank 
Radioactivity Monitoring Program:
    This program provided controls for potentially explosive gas 
mixtures contained in the Radioactive Waste Disposal (WD) System, 
and the quantity of radioactivity contained in gas storage tanks or 
fed into the offgas treatment system.
    This program is being eliminated.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This program is required for an operating plant where hydrogen 
and radioactive gases are created and must be controlled. Controlled 
release of any gases currently in the tanks, in accordance with 
existing procedures, will ensure there will be no hazard to public 
health and safety. Therefore, elimination of this program cannot 
increase the probability or consequences of a fuel handling 
accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This program is required for an operating plant where hydrogen 
and radioactive gases are created and must be controlled. Controlled 
release of any gases currently in the tanks, in accordance with 
existing procedures, will ensure there will be no hazard to public 
health and safety. Therefore, elimination of this program cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margins of safety.
    W. ITS Section 5.6.2.18: Core Operating Limits Report (COLR):
    This program established that core operating limits be 
established prior to each reload cycle.
    This program is being eliminated.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This program for controlling the design and operation of the 
reactor core has no bearing on fuel storage after fuel has been 
moved into the spent fuel pools. Therefore, eliminating this program 
cannot increase the probability or occurrence of a fuel handling 
accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Since CR-3 can never load a core into the reactor again, 
eliminating this control program cannot create a new or different 
kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Since CR-3 can never load a core into the reactor again, 
eliminating this control program cannot affect any margin of safety.
    X. ITS 5.6.2.19: Reactor Coolant System (RCS) Pressure And 
Temperature Limits Report (PTLR):
    This program ensured that RCS pressure and temperature limits, 
including heatup and cooldown rates, criticality, and hydrostatic 
and leak test limits, be established and documented in the PTLR.
    This program is being eliminated.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This program contains no actions or limits that affect the 
storage or handling of nuclear fuel. Therefore, eliminating this 
program cannot increase the probability or occurrence of a fuel 
handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This report is no longer needed since the reactor coolant system 
is not subject to pressurization and the reactor contains no fuel. 
Therefore, eliminating this control program cannot create a new or 
different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The limits established in this report do not apply to nuclear 
fuel stored in the spent fuel pools. Therefore, eliminating this 
program cannot affect any margin of safety.
    Y. ITS Section 5.6.2.20: Containment Leakage Rate Testing 
Program:
    This program was established to implement the leakage rate 
testing of the containment.
    This program is being eliminated in accordance with Regulatory 
Guide 1.1.84.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Since fuel can never be returned to the CR-3 containment, ending 
containment leakage rate testing cannot increase the probability or 
occurrence of a fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This change does not introduce any changes to the function of 
any plant structures, systems, or components therefore it cannot 
create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This change does not directly involve any limits or parameters 
and therefore cannot affect any margin of safety.
    Z. ITS Section 5.7.2: Special Reports:
    This section is being revised to eliminate reporting 
requirements associated with programs that are being eliminated.

[[Page 67411]]

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Eliminating reporting requirements for programs that are no 
longer required in a permanently defueled plant cannot increase the 
probability or occurrence of a fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Eliminating reporting requirements that are no longer required 
cannot create a new or different kind of accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Eliminating reporting requirements that are no longer required 
cannot affect any margin of safety.
    AA. ITS Section 5.8.2: High Radiation Area Controls:
    Changes one of the personnel responsible for locked high 
radiation area key control from the Control Room Supervisor to the 
Shift Supervisor.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This is a change to the requirements for the position title 
responsible for key control. In a permanently defueled plant, the 
fuel handling accident is the only credible accident previously 
evaluated. This action cannot increase the probability or 
consequences of a fuel handling accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The change proposed here, identifying a different position title 
responsible for key control, cannot create a new or different kind 
of accident since they do not change the function of any plant 
structures, systems, or components.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The changes proposed here for key control do not directly 
involve any limits or parameters and therefore cannot affect 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kathryn B. Nolan, 550 South Tryon Street, 
Charlotte, North Carolina, 28202.
    NRC Branch Chief: Jessie F. Quichocho.

Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220, and 50-410, 
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County, New 
York

    Date of amendment request: October 7, 2013.
    Description of amendment request: The proposed amendment modifies 
the Nine Mile point Units 1 and 2 TS definition of ``Shutdown Margin'' 
(SDM) to require calculation of the SDM at a reactor moderator 
temperature of 68[emsp14][deg]F or a higher temperature that represents 
the most reactive state throughout the operating cycle. This change is 
needed to address new Boiling Water Reactor (BWR) fuel designs which 
may be more reactive at shutdown temperatures above 68[emsp14][deg]F.
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
opportunity for comment in the Federal Register on November 19, 2012; 
77 FR 69507, on possible amendments to revise the plant specific TS, to 
modify the TS definition of ``Shutdown Margin'' (SDM) to require 
calculation of the SDM at a reactor moderator temperature of 
68[emsp14][deg]F or a higher temperature that represents the most 
reactive state throughout the operating cycle, including a model safety 
evaluation and model NSHC [no significant hazards consideration] 
determination, using the consolidated line-item improvement process. 
The NRC staff subsequently issued a notice of availability of the 
models for referencing in license amendment applications in the Federal 
Register on February 26, 2013 (78 FR 13100). The licensee affirmed the 
applicability of the model NSHC determination in its application dated 
October 7, 2013, which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC adopted by the licensee is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the definition of SDM. SDM is not an 
initiator to any accident previously evaluated. Accordingly, the 
proposed change to the definition of SDM has no effect on the 
probability of any accident previously evaluated. SDM is an 
assumption in the analysis of some previously evaluated accidents 
and inadequate SDM could lead to an increase in consequences for 
those accidents. However, the proposed change revises the SDM 
definition to ensure that the correct SDM is determined for all fuel 
types at all times during the fuel cycle. As a result, the proposed 
change does not adversely affect the consequences of any accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the definition of SDM. The change 
does not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operations. The change does not alter 
assumptions made in the safety analysis regarding SDM.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the definition of SDM. The proposed 
change does not alter the manner in which safety limits, limiting 
safety system settings or limiting conditions for operation are 
determined. The proposed change ensures that the SDM assumed in 
determining safety limits, limiting safety system settings or 
limiting conditions for operation is correct for all fuel types at 
all times during the fuel cycle.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the analysis adopted by the licensee 
and, based on this review, it appears that the standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the request for amendment involves NSHC.
    Attorney for licensee: Gautam Sen, Senior Counsel, Constellation 
Energy Nuclear Group, LLC, 100 Constellation Way, Suite 200C, 
Baltimore, MD 21202.
    NRC Branch Chief: Robert Beall.

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

    Date of amendment request: September 25, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer 
Nuclear Station (VCSNS) Units 2 and 3 by departing from VCSNS Units 2 
and 3 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by 
revising reference document APP-OCS-GEH-220, ``AP1000 Human Factors 
Engineering Task Support Verification Plan,'' from Revision B to 
Revision 1. APP-OCS-GEH-220 is incorporated by

[[Page 67412]]

reference in the UFSAR as a means to implement the activities 
associated with the human factors engineering verification and 
validation.
    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 HFE Task Support Verification Plan is one of several 
verification and validation (V&V) activities performed on human-
system interface (HSI) resources and the Operation and Control 
Centers System (OCS), where applicable. The Task Support 
Verification Plan is used to assess and verify displays and 
activities related to normal and emergency operation. The changes 
are to the Task Support Verification Plan to clarify the scope and 
amend the details of the methodology. The Task Support Verification 
Plan does not affect the plant itself. Changing the Plan does not 
affect prevention and mitigation of abnormal events, e.g., 
accidents, anticipated operational occurrences, earthquakes, floods 
and turbine missiles, or their safety or design analyses. The 
Probabilistic Risk Assessment is not affected. No safety-related 
structure, system, component (SSC) or function is adversely 
affected. The change does not involve nor interface with any SSC 
accident initiator or initiating sequence of events, and thus, the 
probabilities of the accidents evaluated in the UFSAR are not 
affected. Because the changes do not involve any safety-related SSC 
or function used to mitigate an accident, the consequences of the 
accidents evaluated in the UFSAR are not affected.
    Therefore, there is no 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 changes to the Task Support Verification Plan change 
information related to validation and verification on Human System 
Interface and Operational Control Centers. Therefore, the changes do 
not affect the safety-related equipment itself, nor do they affect 
equipment which, if it failed, could initiate an accident or a 
failure of a fission product barrier. No analysis is adversely 
affected. No system or design function or equipment qualification 
will be adversely affected by the changes. This activity will not 
allow for a new fission product release path, nor will it result in 
a new fission product barrier failure mode, nor create a new 
sequence of events that would result in significant fuel cladding 
failures. In addition, the changes do not result in a new failure 
mode, malfunction, or sequence of events that could affect safety or 
safety-related equipment.
    Therefore, this activity does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The changes to the Task Support Verification Plan affect the 
validation and verification on the Human System Interface and the 
Operational Control Centers. Therefore, the changes do not affect 
the plant itself. These changes do not affect the design or 
operation of safety-related equipment or equipment whose failure 
could initiate an accident, nor does it adversely interface with 
safety-related equipment or fission product barriers. No safety 
analysis or design basis acceptance limit/criterion is challenged or 
exceeded by the requested change.
    Therefore, the 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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart.

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

    Date of amendment request: September 25, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer 
Nuclear Station (VCSNS) Units 2 and 3 by departing from VCSNS Units 2 
and 3 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by 
revising reference document APP-OCS-GEH-120, ``AP1000 Human Factors 
Design Engineering Verification Plan,'' from Revision B to Revision 1. 
APP-OCS-GEH-120 is incorporated by reference in the updated UFSAR as a 
means to implement the activities associated with the human factors 
engineering verification and validation.
    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.
    Design verification provides a final check of the adequacy of 
the Human System Interface (HSI) Resources and Operation and Control 
Centers System (OCS) design. The changes do not affect the plant 
itself, and so there is no change to the probability or consequences 
of an accident previously evaluated. Changing the design 
verification plan does not affect prevention and mitigation of 
abnormal events, e.g., accidents, anticipated operational 
occurrences, earthquakes, floods and turbine missiles, or their 
safety or design analyses as the purpose of the plan is simply to 
verify implementation of design criteria. The Probabilistic Risk 
Assessment is not affected. No safety-related structure, system, 
component (SSC) or function is adversely affected. The change does 
not involve nor interface with any SSC accident initiator or 
initiating sequence of events, and thus, the probabilities of the 
accidents evaluated in the UFSAR are not affected. Because the 
changes do not involve any safety-related SSC or function used to 
mitigate an accident, the consequences of the accidents evaluated in 
the UFSAR are not affected.
    Therefore, there is no 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.
    Design verification provides a final check of the adequacy of 
the HSI Resources and Operation and Control Centers System design. 
The changes do not affect the plant itself, and so there is no new 
or different kind of accident from any accident previously 
evaluated. Therefore, the changes do not affect safety-related 
equipment, nor does it affect equipment which, if it failed, could 
initiate an accident or a failure of a fission product barrier. No 
analysis is adversely affected. No system or design function or 
equipment qualification is adversely affected by the changes. This 
activity will not allow for a new fission product release path, nor 
will it result in a new fission product barrier failure mode, nor 
create a new sequence of events that would result in significant 
fuel cladding failures. In addition, the changes do not result in a 
new failure mode, malfunction, or sequence of events that could 
affect safety or safety-related equipment.
    Therefore, this activity does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The changes to the design verification plan provide a final 
check of the adequacy of the HSI Resources and Operation and Control 
Centers System design. The changes do not affect the assessments or 
the plant itself. The changes do not affect safety-related equipment 
or equipment whose failure could initiate an accident, nor does it 
adversely interface with safety-related equipment or

[[Page 67413]]

fission product barriers. No safety analysis or design basis 
acceptance limit/criterion is challenged or exceeded by the 
requested change.
    Therefore, the 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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart.

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

    Date of amendment request: October 3, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer 
Nuclear Station (VCSNS) Units 2 and 3 by departing from VCSNS Units 2 
and 3 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by 
revising material by revising reference document APP-OCS-GEH-520, 
``AP1000 Plant Startup Human Factors Engineering Design Verification 
Plan,'' from Revision B to Revision 2. APP-OCS-GEH-520 is incorporated 
by reference in the UFSAR as a means to implement the activities 
associated with the human factors engineering verification and 
validation.
    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 APP-OCS-GEH-520, document confirms aspects of the human 
system interface (HSI) and Operation and Control Centers Systems 
(OCS) design features that could not be evaluated in other Human 
Factors Engineering (HFE) verification and validation (V&V) 
activities. It also confirms that the as-built in the plant HSIs, 
procedures, and training conform to the design that resulted from 
the HFE program. Additionally, it confirms that all HFE-related 
issues (including human error discrepancies (HEDs)) documented in 
the SmartPlant Foundation (SPF) Human Factors (HF) Tracking System 
are verified as adequately addressed or resolved. Finally, it 
confirms the HFE adequacy for risk-important human actions in the 
local plant, including the ability for the tasks to be completed 
within the time window according to the Probabilistic Risk 
Assessment (PRA). The changes to the plan are to clarify the scope 
and amend the details of the methodology. The plan does not affect 
the plant itself. Changing the plan does not affect prevention and 
mitigation of abnormal events, e.g., accidents, anticipated 
operational occurrences, earthquakes, floods and turbine missiles, 
or their safety or design analyses. The PRA is not affected. No 
safety-related Structure, System, or Component (SSC) or function is 
adversely affected. The document revision change does not involve 
nor interface with any SSC accident initiator or initiating sequence 
of events, and thus, the probabilities of the accidents evaluated in 
the Updated Final Safety Analysis Report (UFSAR) are not affected. 
Because the changes to the plan do not involve any safety-related 
SSC or function used to mitigate an accident, the consequences of 
the accidents evaluated in the UFSAR are not affected.
    Therefore, there is no 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.
    APP-OCS-GEH-520, ``AP1000 Plant Startup Human Factors 
Engineering Design Verification Plan'' is the plan to confirm 
aspects of the HSI and OCS design features that could not be 
evaluated in other HFE V&V activities. The plan also confirms that 
the as-built in the plant HSIs, procedures, and training conform to 
the design that resulted from the HFE program. Additionally, it 
confirms that all HFE-related issues (including HEDs) documented in 
the SPF HF Tracking System are verified as adequately addressed or 
resolved. Finally, it confirms the HFE adequacy for risk-important 
human actions in the local plant, including the ability for the 
tasks to be completed within the time window according to the PRA. 
These functions support evaluating the HSI and OCS. Therefore, the 
changes do not affect the safety-related equipment itself, nor do 
they affect equipment which, if it failed, could initiate an 
accident or a failure of a fission product barrier. No analysis is 
adversely affected. No system or design function or equipment 
qualification will be adversely affected by the changes. This 
activity will not allow for a new fission product release path, nor 
will it result in a new fission product barrier failure mode, nor 
create a new sequence of events that would result in significant 
fuel cladding failures. In addition, the changes do not result in a 
new failure mode, malfunction, or sequence of events that could 
affect safety or safety-related equipment.
    Therefore, this activity does not create the possibility of a 
new or different kind of accident than any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    APP-OCS-GEH-520, ``AP1000 Plant Startup Human Factors 
Engineering Design Verification Plan'' is the plan to confirm 
aspects of the HSI and OCS design features that could not be 
evaluated in other HFE V&V activities. The plan also confirms that 
the as-built in the plant HSIs, procedures, and training conform to 
the design that resulted from the HFE program. Additionally, it 
confirms that all HFE-related issues (including HEDs) documented in 
the SPF HF Tracking System are verified as adequately addressed or 
resolved. Finally, it confirms the HFE adequacy for risk-important 
human actions in the local plant, including the ability for the 
tasks to be completed within the time windows in the PRA. These 
functions support evaluating the HSI and OCS. The proposed changes 
to the plan do not affect the design or operation of safety-related 
equipment or equipment whose failure could initiate an accident, nor 
does the plan adversely affect the interfaces with safety-related 
equipment or fission product barriers. No safety analysis or design 
basis acceptance limit/criterion is challenged or exceeded by the 
requested changes.
    Therefore, the 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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart.

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

    Date of amendment request: October 3, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer 
Nuclear Station (VCSNS) Units 2 and 3 by departing from VCSNS Units 2 
and 3 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by 
revising reference document APP-OCS-GEH-420, ``AP1000 Human Factors 
Engineering Discrepancy Resolution Process,'' from Revision B to 
Revision 1. APP-OCS-GEH-420 is incorporated by reference in the UFSAR 
as a means to implement the activities associated with the human 
factors engineering

[[Page 67414]]

verification and validation (TAC No. RQ0403) (LAR 13-18).
    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 HFE Discrepancy Resolution Process is used to capture and 
resolve Human Engineering Discrepancies (HEDs) identified during the 
Human Factors Engineering (HFE) verification and validation (V&V) 
activities. These discrepancy resolution process activities are used 
to support the final check of the adequacy of the HFE design of the 
Human-System Interface (HSI) resources and the Operation and Control 
Centers Systems (OCS) design. The discrepancy resolution process 
activities are performed as part of the V&V activities against the 
final configuration and control documentation, simulator or 
installed target system. The changes are to the Discrepancy 
Resolution Process to clarify the scope and amend the details of the 
methodology. The Discrepancy Resolution Process does not affect the 
plant itself. Changing the Discrepancy Resolution Process does not 
affect prevention and mitigation of abnormal events, e.g., 
accidents, anticipated operational occurrences, earthquakes, floods 
and turbine missiles, or their safety or design analyses. No safety-
related structure, system, component (SSC) or function is adversely 
affected. The document revision does not involve nor interface with 
any SSC accident initiator or initiating sequence of events, and 
thus the probabilities of the accidents evaluated in the Updated 
Final Safety Analysis Report (UFSAR) are not affected. Because the 
changes do not involve any safety-related SSC or function used to 
mitigate an accident, the consequences of the accidents evaluated in 
the UFSAR are not affected.
    Therefore, there is no 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 changes to the Discrepancy Resolution Process information 
are related to discrepancy resolution of HEDs during the HFE V&V 
activities on the HSI and the OCS. Therefore, the changes do not 
affect the safety-related equipment itself, nor do they affect 
equipment which, if it failed, could initiate an accident or a 
failure of a fission product barrier. No analysis is adversely 
affected. No system or design function or equipment qualification 
will be adversely affected by the changes. This activity will not 
allow for a new fission product release path, nor will it result in 
a new fission product barrier failure mode, nor create a new 
sequence of events that would result in significant fuel cladding 
failures. In addition, the changes do not result in a new failure 
mode, malfunction, or sequence of events that could affect safety or 
safety-related equipment.
    Therefore, this activity does not create the possibility of a 
new or different kind of accident than any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The changes to the Discrepancy Resolution Process affect 
discrepancy resolution of HEDs during the HFE V&V activities on the 
HSI and the OCS. Therefore, the changes do not affect the 
assessments or the plant itself. These changes do not affect the 
design or operation of safety-related equipment or equipment whose 
failure could initiate an accident, nor does it adversely interface 
with safety-related equipment or fission product barriers. No safety 
analysis or design basis acceptance limit/criterion is challenged or 
exceeded by the requested change.
    Therefore, the 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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room (PDR), located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available documents created or received at the 
NRC are accessible electronically through the Agencywide Documents 
Access and Management System (ADAMS) in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by 
email to pdr.resource@nrc.gov.

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

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

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

    Date of amendment request: April 20, 2013.
    Brief description of amendment: The amendments revised the 
corporate name of the licensee in each facility's operating license 
from Carolina Power & Light Company to Duke Energy Progress, Inc.
    Date of issuance: October 21, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 263, 291, 142, and 236.
    Renewed Facility Operating License Nos. DPR-71, DPR-62, NPF-63, and 
DPR-23: Amendments revised the Licenses and Appendix cover pages.

[[Page 67415]]

    Dates of initial notice in Federal Register: May 28, 2013 (78 FR 
31982) and correction to initial notice on June 21, 2013 (78 FR 37595).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 21, 2013.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 20, 2013.
    Brief description of amendment: The amendment changes the name of 
the Licensee in the Facility Operating License.
    Date of issuance: October 18, 2013.
    Effective date: Date of issuance, to be implemented within 60 days.
    Amendment No.: 243.
    Facility Operating License No. DPR-72: Amendment revises the 
Facility Operating License.
    Date of initial notice in Federal Register: April 30, 2013 (78 FR 
25314).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 18, 2013.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: July 1, 2011, as supplemented 
by letters dated September 2, 2011, April 27, 2012, June 29, 2012, 
August 9, 2012, October 15, 2012, November 9, 2012, January 14, 2013, 
February 1, 2013, May 1, 2013, June 21, 2013, and September 16, 2013.
    Brief description of amendments: The amendments revised the 
facility operating licenses and transitions the Donald C. Cook Nuclear 
Plant fire protection program to a new risk-informed, performance-based 
alternative in accordance with 10 CFR 50.48(c), which incorporates by 
reference the National Fire Protection Association (NFPA) Standard 805 
(NFPA 805), ``Performance-Based Standard for Fire Protection for Light 
Water Reactor Electric Generating Plants--2001.''
    Date of issuance: October 24, 2013.
    Effective date: As of the date of issuance and shall be implemented 
by October 24, 2014.
    Amendment Nos.: Unit 1--322; Unit 2--305.
    Facility Operating License No. DPR-58 and DPR-74: Amendments 
revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: October 4, 2011 (76 FR 
61396). The supplemental letters dated September 2, 2011, April 27, 
2012, June 29, 2012, August 9, 2012, October 15, 2012, November 9, 
2012, January 14, 2013, February 1, 2013, May 1, 2013, June 21, 2013, 
and September 16, 2013, provided additional information that clarified 
the application, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 24, 2013.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: March 27, 2013, as supplemented June 25, 
2013.
    Brief description of amendment: The amendment revised the Seabrook 
TS. The amendment modifies TS requirements regarding steam generator 
tube inspections and reporting as described in TS Task Force (TSTF)-
510, Revision 2, ``Revision to Steam Generator Program Inspection 
Frequencies and Tube Sample Selection,'' using the Consolidated Line 
Item Improvement Process (CLIIP). The changes are consistent with 
Industry/TSTF Standard Technical Specification Change Traveler, TSTF-
510. The availability of this TS improvement was announced in the 
Federal Register on October 27, 2011 (76 FR 66763), as part of the 
CLIIP.
    Date of issuance: October 25, 2013.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 138.
    Facility Operating License No. NPF-86: The amendment revised the 
Facility Operating License and TS.
    Date of initial notice in Federal Register: April 30, 2013 (78 FR 
25316). The supplement dated June 25, 2013, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 25, 2013.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 19, 2013, and revised by the letter 
dated August 27, 2013.
    Brief description of amendment: The proposed amendment would depart 
from VEGP Units 3 and 4 plant-specific Design Control Document (DCD) 
Tier 2* and associated Tier 2 material incorporated into the Updated 
Final Safety Analysis Report (UFSAR) by revising requirements for 
design spacing of shear studs and the design of structural elements in 
order to address interferences and obstructions other than wall 
openings.
    Date of issuance: October 8, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: Unit 3-14, and Unit 4-14.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: August 6, 2013 (78 FR 
47792).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 8, 2013.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses and Final Determination of No Significant Hazards 
Consideration and Opportunity for a Hearing (Exigent Public 
Announcement or Emergency Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish,

[[Page 67416]]

for public comment before issuance, its usual notice of consideration 
of issuance of amendment, proposed no significant hazards consideration 
determination, and opportunity for a hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 51.12(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 
application for amendment, (2) the amendment to Facility Operating 
License or Combined License, as applicable, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment, as 
indicated. All of these items are available for public inspection at 
the NRC's Public Document Room (PDR), located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through the Agencywide Documents Access and 
Management System (ADAMS) in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
PDR Reference staff at 1-800-397-4209, 301-415-4737 or by email to 
pdr.resource@nrc.gov.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. Within 60 days after the date 
of publication of this notice, any person(s) whose interest may be 
affected by this action may file a request for a hearing and a petition 
to intervene with respect to issuance of the amendment to the subject 
facility operating license or combined license. Requests for a hearing 
and a petition for leave to intervene shall be filed in accordance with 
the Commission's ``Rules of Practice for Domestic Licensing 
Proceedings'' in 10 CFR Part 2. Interested person(s) should consult a 
current copy of 10 CFR 2.309, which is available at the NRC's PDR, 
located at One White Flint North, Room O1-F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, and electronically on the 
Internet at the NRC Web site, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the document, 
contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737, or 
by email to pdr.resource@nrc.gov. 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 petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the 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 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. Since the Commission has made a final determination that the 
amendment involves no significant hazards consideration, if a hearing 
is

[[Page 67417]]

requested, it will not stay the effectiveness of the amendment. Any 
hearing held would take place while the amendment is in effect.
    All documents filed in the NRC adjudicatory proceedings, including 
a request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the Internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital information (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 NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in NRC's ``Guidance for Electronic 
Submission,'' which is available on the agency's public Web site at 
https://www.nrc.gov/site-help/e-submittals.html. Participants may 
attempt to use other software not listed on the Web site, but should 
note that the NRC's E-Filing system does not support unlisted software, 
and the NRC Meta System Help Desk will not be able to offer assistance 
in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through Electronic Information Exchange System, users 
will be required to install a Web browser plug-in from the NRC 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 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 Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by email at 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of amendment request: October 7, 2013, as supplemented by 
letters dated October 8 and October 9, 2013.
    Description of amendment request: This notice was previously 
published in the Federal Register on October 29, 2013 (78 FR 64550). 
This notice is being reissued in its entirety as it was

[[Page 67418]]

inadvertently placed in the incorrect section of the Biweekly report 
published on October 29, 2013. The amendment revised Technical 
Specification (TS) 3.6.9, ``Distributed Ignition System (DIS),'' to 
allow Train B of the DIS to be considered operable with two inoperable 
ignitors. The existing TS defines train operability as having no more 
than one ignitor inoperable. The amendment also allows one of five 
specific primary containment regions to have zero ignitors operable. 
The existing TS requires that at least one ignitor be operable in each 
region. The proposed TS revision is applicable until the fall 2014 
refueling outage, or until the unit enters a mode that allows 
replacement of the affected ignitors without exposing personnel to 
significant radiation and safety hazards.
    Date of issuance: October 9, 2013.
    Effective date: As of the date of issuance, to be implemented 
within 1 day.
    Amendment No.: 321.
    Renewed Facility Operating License No. DPR-58: Amendment revised 
the Technical Specifications and License.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated October 9, 
2013.
    Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    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: October 6, 2013, as supplemented by 
letters dated October 15, 21, and 22, 2013 and two letters dated 
October 23, 2013.
    Description of amendment request: The amendment revised the Updated 
Safety Analysis Report (USAR) for pipe break criteria for high energy 
piping outside of containment. Specifically, the proposed amendment 
would allow the use of NRC guidance provided in Branch Technical 
Position Mechanical Engineering Branch 3-1, Revision 2, which allows 
for the exemption of specific piping sections from postulated failures 
if certain criteria are met.
    Date of issuance: October 25, 2013.
    Effective date: As of its issuance date and shall be implemented 
upon approval.
    Amendment No.: 273.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the facility operating license.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (Omaha-World Herald, located in Omaha, 
Nebraska, from October 9 through October 15, 2013). The notice provided 
an opportunity to submit comments on the Commission's proposed NSHC 
determination. One comment was received and evaluated.
    The supplemental letters dated October 15, 21, and 22, 2013, and 
two letters dated October 23, 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 Omaha-World Herald from October 9 through 15, 2013.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
(including the comment received on the NSHC) are contained in a safety 
evaluation dated October 25, 2013.
    Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 
K Street, NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.

    Dated at Rockville, Maryland, this 1st day of November 2013.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-27025 Filed 11-8-13; 8:45 am]
BILLING CODE 7590-01-P