Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 15144-15157 [2014-05645]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 79, Number 52 (Tuesday, March 18, 2014)]
[Notices]
[Pages 15144-15157]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2014-05645]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0045]


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 amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 5 to March 18, 2014. The last biweekly 
notice was published on March 4, 2014 (79 FR 12241).

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2014-0045. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: 3WFN-06-44M, U.S. Nuclear Regulatory

[[Page 15145]]

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-2014-0045 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-2014-0045.
     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. The ADAMS accession number 
for each document referenced in this document (if that document is 
available in ADAMS) is provided the first time that a document is 
referenced.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2014-0045 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at https://www.regulations.gov as well as entering the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of Title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated; or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR Part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC regulations are accessible electronically from the NRC Library on 
the NRC's Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed by the above date, the Commission or a presiding officer 
designated by the Commission or by the Chief Administrative Judge of 
the Atomic Safety and Licensing Board Panel, will rule on the request 
and/or petition; and the Secretary or the Chief Administrative Judge of 
the Atomic Safety and Licensing Board will issue a notice of a hearing 
or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) the name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert

[[Page 15146]]

opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having

[[Page 15147]]

granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/;, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through ADAMS in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to pdr.resource@nrc.gov.
    Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating, Unit 2, Westchester County, New York
    Date of amendment request: January 16, 2014.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.5.7, ``Steam Generator (SG) 
Program,'' to exclude portions of the SG tube below the top of the SG 
tubesheet from periodic inspections and plugging by implementing the H* 
alternate repair criteria. In addition, TS 5.6.7, ``Steam Generator 
Tube Inspection Report,'' would also be revised to include additional 
reporting requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change excludes the lower portion of steam 
generator tubes from inspection by implementing the alternate repair 
criteria H* and does not have a detrimental impact on the integrity 
of any plant structure, system, or component that initiates an 
analyzed event. The proposed change has no significant effect upon 
accident probabilities or consequences.
    Of the applicable accidents previously evaluated, the limiting 
transients with consideration to the proposed change to the steam 
generator tube inspection and repair criteria are the steam 
generator tube rupture (SGTR), the main steam line break (MSLB), 
Locked Rotor and Control Rod Ejection.
    At normal operating pressures, leakage from Primary Water Stress 
Corrosion Cracking (PWSCC) below the proposed limited inspection 
depth is limited by both the tube-to-tubesheet crevice and the 
limited crack opening permitted by the tubesheet constraint. 
Consequently, negligible normal operating leakage is expected from 
cracks within the tubesheet region.
    For the SGTR event, the required structural integrity margins of 
the steam generator tubes and the tube-to-tubesheet joint over the 
H* distance will be maintained. Tube rupture in tubes with cracks 
within the tubesheet is precluded by the constraint provided by the 
tube-to-tubesheet joint. This constraint results from the hydraulic 
expansion process, thermal expansion mismatch between the tube and 
tubesheet, and from the differential pressure between the primary 
and secondary side. The structural margins against burst, as 
discussed in Regulatory Guide (RG) 1.121, ``Bases for Plugging 
Degraded PWR Steam Generator Tubes,'' (Reference 11) and NEI 97-06, 
``Steam Generator Program Guidelines'' (Reference 3) are maintained 
for both normal and postulated accident conditions. Therefore, the 
proposed change results in no significant increase in the 
probability of the occurrence of a SGTR accident.
    The probability of a Steam Line Break, Locked Rotor, and Control 
Rod Ejection are not affected by the potential failure of a SG tube, 
as the failure of a tube is not an initiator for any of these 
events. In the supporting Westinghouse analyses, leakage is modeled 
as flow through a porous medium via the use of the Darcy equation. 
The leakage model is used to develop a relationship between 
allowable leakage and leakage at accident conditions that is based 
on differential pressure across the tubesheet and the viscosity of 
the fluid. A leak rate ratio was developed to relate the leakage at 
operating conditions to leakage at accident conditions. The fluid 
viscosity is based on fluid temperature and it has been shown that 
for the most limiting accident, the fluid temperature does not 
exceed the normal operating temperature. Therefore, the viscosity 
ratio is assumed to be 1.0 and the leak rate ratio is a function of 
the ratio of the accident differential pressure and the normal 
operating differential pressure.
    The leakage factor of 1.75 for IP2 for a-postulated MSLB, has 
been calculated as shown in the supporting Westinghouse analysis. 
IP2 [Indian Point Unit 2] will apply a factor of 1.75 to the normal 
operating leakage associated with the tubesheet expansion region in 
the Condition Monitoring Assessment and Operational Assessment. 
Through application of the limited tubesheet inspection scope, the 
administrative leakage limit of 75 gpd [gallons per day] provides 
assurance that excessive leakage (i.e., greater than accident 
analysis assumptions) will not occur. No leakage factor will be 
applied to the Locked Rotor or Control Rod Ejection due to their 
short duration, since the calculated leak rate ratio is less than 
1.0. Therefore, the proposed change does not result in a significant 
increase in the consequences of these accidents.
    For the Condition Monitoring Assessment, the component of 
leakage from the prior cycle from below the H* distance will be 
multiplied by a factor of 1.75 and added to the total leakage from 
any other source and compared to the allowable MSLB leakage limit. 
For the Operational Assessment, the difference in the leakage 
between the allowable leakage and the accident induced leakage from 
sources other than the tubesheet expansion region will be divided by 
1.75 and compared to the observed operational leakage. As noted 
above, an administrative limit of 75 gpd has been established at IP2 
to assure that the allowable accident induced leakage is not 
exceeded.
    Based on the above, the performance criteria of NEI 97-06 and 
Regulatory Guide (RG) 1.121 continue to be met and 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 excludes the lower portion of steam 
generator tubes from inspection by implementing the alternate repair 
criteria (H*). The proposed change does not introduce any new 
equipment, create new failure modes for existing

[[Page 15148]]

equipment, or create any new limiting single failures resulting from 
tube degradation. The proposed change does not affect the design of 
the SGs or their method of operation. In addition, the proposed 
change does not impact any other plant system or component. Plant 
operation will not be altered, and all safety functions will 
continue to perform as previously assumed in accident analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change defines the safety significant portion of 
the SG tubing that must be inspected and repaired. WCAP-17828-P 
identifies the inspection depth below which any type of degradation 
is shown to have no impact on the steam generator tube integrity 
performance criteria in NEI 97-06. The proposed change does not 
affect tube design or operating environment. The proposed change 
will continue to require monitoring of the physical condition of the 
SG tubes but will limit inspection within the tubesheet to the 
portion of the tube from the top of the tubesheet to a distance H* 
below the top of the tubesheet.
    The proposed change maintains the required structural margins of 
the SG tubes for both normal and accident conditions. For axially 
oriented cracking located within the tubesheet, tube burst is 
precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, the supporting Westinghouse 
analyses define a length of degradation-free expanded tubing that 
provides the necessary resistance to tube pullout due to the 
pressure induced forces, with applicable safety factors applied. 
Application of the limited hot and cold leg tubesheet inspection 
criteria will preclude unacceptable primary to secondary leakage 
during all plant conditions. The MSLB leak rate factor for IP2 is 
1.75. Multiplying the IP2 administrative leak rate limit of 75 gpd/
SG by this factor shows that the primary-to-secondary leak rate 
during a postulated SLB is not exceeded.
    Therefore, the proposed change does not involve a significant 
reduction in any margin of safety.
    Based on the above, Entergy concludes that the proposed 
amendment to the Indian Point 2 Technical Specifications presents no 
significant hazards consideration under the standards set forth in 
10 CFR 50.92(c), and accordingly, a finding of `no significant 
hazards consideration' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Benjamin G. Beasley.
    Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades 
Nuclear Plant, Van Buren County, Michigan
    Date of amendment request: June 25, 2013, supplemented by letter 
dated August 7, 2013.
    Description of amendment request: The proposed amendment would 
revise Palisades Nuclear Plant Site Emergency Plan (SEP) to increase 
the staff augmentation response times for certain Emergency Response 
Organization positions from 30 to 60 minutes. Entergy Nuclear 
Organization has reviewed the proposed changes against the standards in 
Sec.  50.47(b) and the requirements in 10 CFR Part 50, Appendix E.
    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 extension of staff augmentation times has no effect 
on normal plant operation or on any accident initiator. The change 
affects the response to radiological emergencies under the Palisades 
Nuclear Plant SEP. The ability of the emergency response 
organization to respond adequately to radiological emergencies has 
been evaluated. Changes in the on-shift organization, such as the 
addition of staff and reassignment of key on-shift emergency 
response functions, provide assurance of emergency response without 
competing or conflicting duties. An analysis was also performed on 
the effect of the proposed change on the timeliness of performing 
major tasks for the major functional areas of the SEP. The analysis 
concluded that extension of staff augmentation times would not 
significantly affect the ability to perform the required tasks.
    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 affects the required response times for 
supplementing onsite personnel in response to a radiological 
emergency. It has been evaluated and determined not to significantly 
affect the ability to perform that function. It has no effect on the 
plant design or on the normal operation of the plant and does not 
affect how the plant is physically operated under emergency 
conditions. The extension of staff augmentation times in the SEP 
does not affect the plant operating procedures which are performed 
by plant staff during all plant conditions.
    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 does not affect plant design or method of 
operation. Section 50.47(b) and 10 CFR Part 50, Appendix E establish 
emergency planning standards and requirements that require adequate 
staffing, satisfactory performance of key functional areas and 
critical tasks, and timely augmentation of the response capability. 
Since the SEP was originally developed, there have been improvements 
in the technology used to support the SEP functions and in the 
capabilities of onsite personnel. A functional analysis was 
performed on the effect of the proposed change on the timeliness of 
performing major tasks for the functional areas of SEP. The analysis 
concluded that an increase in staff augmentation times would not 
significantly affect the ability to perform the required SEP tasks. 
Thus, the proposed change has been determined not to adversely 
affect the ability to meet the emergency planning standards as 
described in 10 CFR 50.47(b) and requirements in 10 CFR Part 50, 
Appendix E.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William Dennis, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 
10601.
    NRC Branch Chief: Robert D. Carlson.
    Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades 
Nuclear Plant, Van Buren County, Michigan
    Date of amendment request: December 11, 2013.
    Description of amendment request: The proposed amendment would 
modify Palisades Nuclear Plant technical specifications (TS) 
requirements for unavailable barriers by adding limiting condition for 
operation (LCO) 3.0.9. The changes are consistent with the NRC's 
approved industry/Technical Specification Task Force (TSTF) Standard 
Technical Specification (STS) change TSTF-427, ``Allowance for Non-
Technical Specification Barrier Degradation on Supported System 
OPERABILITY,'' Revision 2.
    Basis for proposed no significant hazards consideration 
determination: The licensee has affirmed the

[[Page 15149]]

applicability of the model proposed non-significant hazards 
consideration published on October 2, 2006 (71 FR 58444), as part of 
the Consolidated Line Item Improvement Process, ``Notice of 
Availability of the Model Safety Evaluation.'' The licensee has 
concluded that the findings presented in that evaluation are applicable 
to PNP and is hereby referenced below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an unavailable barrier if risk is assessed and managed. 
The postulated initiating events which may require a functional 
barrier are limited to those with low frequencies of occurrence, and 
the overall TS system safety function would still be available for 
the majority of anticipated challenges. Therefore, the probability 
of an accident previously evaluated is not significantly increased, 
if at all. The consequences of an accident while relying on the 
allowance provided by proposed LCO 3.0.9 are no different than the 
consequences of an accident while relying on the TS required actions 
in effect without the allowance provided by proposed LCO 3.0.9. 
Therefore, the consequences of an accident previously evaluated are 
not significantly affected by this change. The addition of a 
requirement to assess and manage the risk introduced by this change 
will further minimize possible concerns.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident from any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to an unavailable barrier, if risk is 
assessed and managed, will not introduce new failure modes or 
effects and will not, in the absence of other unrelated failures, 
lead to an accident whose consequences exceed the consequences of 
accidents previously evaluated. The addition of a requirement to 
assess and manage the risk introduced by this change will further 
minimize possible concerns.
    Thus, this change does not create the possibility of a new or 
different kind of accident from an accident previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety.

    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an unavailable 
barrier, if risk is assessed and managed. The postulated initiating 
events which may require a functional barrier are limited to those 
with low frequencies of occurrence, and the overall TS system safety 
function would still be available for the majority of anticipated 
challenges. The risk impact of the proposed TS changes was assessed 
following the three-tiered approach recommended in RG 1.177. A 
bounding risk assessment was performed to justify the proposed TS 
changes. This application of LCO 3.0.9 is predicated upon the 
licensee's performance of a risk assessment and the management of 
plant risk. The net change to the margin of safety is insignificant 
as indicated by the anticipated low levels of associated risk (ICCDP 
and ICLERP) as shown in Table 1 of Section 3.1.1 in the Safety 
Evaluation.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William Dennis, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White Plains, NY 
10601.
    NRC Branch Chief: Robert D. Carlson.
    Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit 1, Washington County, Nebraska
    Date of amendment request: August 16, 2013.
    Description of amendment request: The proposed amendment would 
revise the design basis method in the Fort Calhoun Station Updated 
Safety Analysis Report for controlling the raw water intake cell level 
during periods of elevated river levels.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed modification engineering change (EC) 55394, Raw 
Water [RW] Pump Operation and Safety Classification of Components 
during a Flood, installed intake cell flood water inlet valves at 
Fort Calhoun Station (FCS). The modification would employ the trash 
rack blowdown portion of the circulating water system to allow river 
water to flow into four of those pipes and then through four newly 
installed safety class valves for control of cell level (RW pump 
suction level) using river level as the driving force. This 
modification EC 55394 enhances the flood protection provided to the 
RW pumps for an external flooding event thus assuring the 
availability of the ultimate heat sink and core cooling. As such, 
the proposed change does not increase the consequences of an 
accident previously evaluated.
    In addition, implementing this strategy eliminates the need for 
the exterior sluice gates to be safety class and allows for 
continuous control of the intake cell level during a design basis 
flood event. The proposed Updated Safety Analysis Report (USAR) 
changes for implementing modification EC 55394 allow for maintaining 
RW pump operation during a flooding event at FCS.
    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 modification EC 55394 to provide control of the 
intake cell level by operation of the manual valves and the 
associated USAR changes do not alter the safety limits or safety 
analysis assumptions associated with the operation of the plant. 
Hence, the proposed changes do not introduce any new accident 
initiators, nor do they reduce or adversely affect the capabilities 
of any plant structure or system in the performance of their safety 
function. The proposed amendment revises the USAR to include the 
necessary information to support the implementation of the 
modification allowing for maintaining RW pump operation during an 
abnormal operating procedure AOP-01 flooding event at FCS.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed modification, which provides control of the intake 
cell level by operation of the manual valves, and the associated 
USAR changes do not alter the safety limits or safety analysis 
assumptions associated with the operation of the plant. The proposed 
modification and associated USAR revisions ensure there is adequate 
protection to the RW pumps from an external flood hazard thus 
assuring adequate protection during a flood. Providing RW pump 
intake cell level control during flooding conditions allows for 
adjustment of flow and control of the intake cell level throughout 
the duration of the flood since the new valves are located inside 
the intake structure; thereby ensuring the RW pumps remain operable 
during a flood condition and will not adversely impact any margin of 
safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are

[[Page 15150]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 
K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Michael T. Markley.
    Southern Nuclear Operating Company, Inc. Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: November 21, 2013.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91 and NPF-92 for the Vogtle Electric 
Generating Plant (VEGP) Units 3 and 4 by departing from the approved 
AP1000 Design Control Document (DCD) Tier 2 information as incorporated 
into the Updated Final Safety Analysis Report (UFSAR) to allow use of a 
new methodology to determine the effective thermal conductivity 
resulting from oxidation of the inorganic zinc (IOZ) used in the 
containment vessel coating system.
    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.
    Implementation of a methodology which specifies an effective 
thermal conductivity and oxidation progression for the inorganic 
zinc coating of the containment vessel is used to eliminate non-
mechanistic modeling of inorganic zinc thermal conductivity in the 
containment integrity analyses to show that the value for inorganic 
zinc thermal conductivity used in the containment integrity analyses 
is conservative, but is not used to change any of the parameters 
used in those analyses. There is no change to any accident initiator 
or condition of the containment that would affect the probability of 
any accident. The containment peak pressure analysis as reported in 
the UFSAR is not affected; therefore, the previously reported 
consequences are not affected.
    Therefore, the proposed amendment does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment to implement a methodology which 
specifies an effective thermal conductivity and oxidation 
progression and effects for the inorganic zinc coating of the 
containment vessel is used to eliminate non-mechanistic modeling of 
inorganic zinc thermal conductivity in the containment integrity 
analyses to show that the value for inorganic zinc thermal 
conductivity used in the containment integrity analyses is 
conservative, but is not used to change any of the parameters used 
in the containment peak pressure analysis. The change in methodology 
does not change the condition of containment; therefore, no new 
accident initiator is created. The containment peak pressure 
analysis as currently evaluated is not affected, and the 
consequences previously reported are not changed. The new 
methodology does not change the containment; therefore, no new fault 
or sequence of events that could lead to containment failure or 
release of radioactive material is created.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed implementation of a methodology which specifies an 
effective thermal conductivity and oxidation progression and effects 
for the inorganic zinc coating of the containment vessel is used to 
eliminate non-mechanistic modeling of inorganic zinc thermal 
conductivity in the containment integrity analyses to show that the 
value for inorganic zinc thermal conductivity used in the 
containment integrity analyses is conservative, but is not used to 
change any of the parameters used in the containment peak pressure 
analysis. The change in methodology does not change the condition of 
the containment and the integrity of the containment vessel is not 
affected. The containment peak pressure analysis as currently 
evaluated is not affected, and the consequences previously reported 
are not changed. No safety analysis or design basis acceptance 
limit/criterion is changed by the proposed change, thus no margin of 
safety is reduced.

    Therefore, the proposed amendment does not reduce the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence J. Burkhart.
    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri
    Date of amendment request: December 6, 2013.
    Description of amendment request: The amendment would add a new 
pipe crack exclusion allowance to Final Safety Analysis Report (FSAR) 
Standard Plant Section 3.6.2.1.2.4, ``ASME [American Society of 
Mechanical Engineers] Section III and Non-Nuclear Piping-Moderate-
Energy,'' and FSAR Standard Plant Table 3.6-2, ``Design Comparison to 
Regulatory Positions of Regulatory Guide 1.46, Revision 0, dated May 
1973, titled `Protection Against Pipe Whip Inside Containment,' '' in 
particular regard to the high-density polyethylene (HDPE) piping 
installed in ASME Class 3 line segments of the essential service water 
(ESW) system. New Reference 25 would be added to FSAR Standard Plant 
Section 3.6.3 to cite the NRC-approved version of the HDPE requirements 
covered by Relief Request I3R-10.
    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.
    There are no new design changes associated with the proposed 
amendment. All design, material, and construction standards that 
were applicable prior to this amendment request, including those 
standards in place following the NRC approval of using the HDPE 
piping, will continue to be applicable.
    The proposed change will not increase the likelihood of accident 
initiators or precursors or adversely alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained with respect to such initiators 
or precursors.
    The proposed changes do not affect the way in which safety-
related systems perform their functions.
    All accident analysis acceptance criteria will continue to be 
met with the proposed changes. The proposed changes will not affect 
the source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated. The proposed changes will not alter 
any assumptions or change any mitigation actions in the radiological 
consequence evaluations in the FSAR.
    The applicable radiological dose acceptance criteria will 
continue to be met.
    Since the proposed change is based on a calculation that 
demonstrates that a moderate energy crack in the ESW HDPE piping is 
unlikely, there are no impacts on the plant's existing hazard 
analyses.
    The proposed change does not physically alter safety-related 
systems or affect the way in which safety-related systems perform 
their functions per the intended plant design.

[[Page 15151]]

    As such, the proposed change will not alter or prevent the 
capability of structures, systems, and components (SSCs) to perform 
their intended functions for mitigating the consequences of an 
accident and meeting applicable acceptance limits.
    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.
    With respect to any new or different kind of accident, there are 
no new design changes being proposed nor are there any changes in 
the method by which any safety-related plant SSC performs its 
specified safety function. The proposed change will not affect the 
normal method of plant operation. No new transient precursors will 
be introduced as a result of this amendment.
    The HDPE piping design change was previously approved by the NRC 
under Relief Request I3R-10. The proposed change in this amendment 
request does not create the possibility of a new type of accident, 
rather the proposed change seeks to eliminate the need to postulate 
an existing type of hazard event (moderate energy piping leakage 
crack) for the subject HDPE piping which has been shown to 
experience such low stresses that such a crack, and the potential 
flooding for that hazard event, need not be postulated.
    The change does not have a detrimental impact on the manner in 
which plant equipment operates or responds to an actuation signal.
    The proposed change does not, therefore, create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    There will be no effect on those plant systems necessary to 
assure the accomplishment of protection functions associated with 
reactor operation or the reactor coolant system. The design factor 
(DF) of 0.50 discussed in ULNRC-05553 dated October 9, 2008 has not 
changed. This DF was approved by the NRC in Relief Request 13R-10 
(Reference 6.2 to this Evaluation). There will be no impact on the 
overpower limit, departure from nucleate boiling ratio (DNBR) 
limits, heat flux hot channel factor (FQ), nuclear enthalpy rise hot 
channel factor (F[Delta]H), loss of coolant accident peak cladding 
temperature (LOCA PCT), peak local power density, or any other limit 
and associated margin of safety. Required shutdown margins in the 
COLR [core operating limits report] will not be changed. The 
proposed change does not eliminate any surveillances or alter the 
frequency of surveillances required by the Technical Specifications.
    As such, the proposed change does not involve a significant 
reduction in a margin of safety as defined in any regulatory 
requirement or guidance document.

    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.
    Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas
    Date of amendment request: November 21, 2013.
    Description of amendment request: The amendment would revise the 
approved Fire Protection Program as described in the Updated Safety 
Analysis Report, based on the reactor coolant system thermal hydraulic 
response evaluation of a postulated control room fire, performed for 
changes to the alternative shutdown methodology.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design function of structures, systems and components (SSCs) 
are not impacted by the proposed deviations from [10 CFR Part 50] 
Appendix R, Sections III.L.1 and III.L.2, and Calculation XX-E-013. 
The proposed changes to the approved fire protection program are 
based on the RCS [reactor coolant system] thermal-hydraulic response 
(Evaluation SA-08-006) for a postulated control room fire performed 
for changes to the alternative shutdown methodology outlined in 
letter SLNRC 84-0109, ``Fire Protection Review.'' Drawing E-1F9915, 
``Design Basis Document for OFN RP-017, Control Room Evacuation,'' 
Revision 5, Evaluation SA-08-006, ``RETRAN-3D Post-Fire Safe 
Shutdown (PFSSD) Consequence Evaluation for a Postulated Control 
Room Fire,'' Revision 3, and Calculation WCNOC-CP-003, ``VIPRE-01 
MDNBR Analyses of Control Room Fire Scenarios,'' Revision 0 
demonstrate the adequacy of the revised alternative shutdown 
procedure, OFN RF-017. The proposed changes do not alter or prevent 
the ability of SSCs from performing their intended function to 
mitigate the consequences of an initiating event within the assumed 
acceptance limits.
    Therefore, the probability of any accident previously evaluated 
is not increased. Equipment required to mitigate an accident remains 
capable of performing the assumed function.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes will not alter the requirement or function 
for systems required during accident conditions. The design function 
of structures, systems and components are not impacted by the 
proposed change. Evaluation SA-08-006 and Calculation WCNOC-CP-003 
determined natural circulation is maintained and adequate core 
cooling is maintained. The fission product boundary integrity is not 
affected and safe shutdown capability is maintained.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. The revised alternative shutdown 
methodology provides the ability to achieve and maintain safe 
shutdown in the event of a fire. Evaluation SA-08-006 and 
Calculation WCNOC-CP-003 determined natural circulation is 
maintained and adequate core cooling is maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.
    Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas
    Date of amendment request: December 17, 2013.
    Description of amendment request: The amendment would revise 
Technical Specification Surveillance Requirement (SR) 3.7.10.1 and SR 
3.7.13.1 to reduce the required run time for periodic operation of the 
control room pressurization system filter trains and emergency exhaust 
system filter trains, with heaters on, from 10 hours to 15 minutes. The 
proposed amendment is consistent with plant-specific options provided 
in the NRC's model safety evaluation of Technical Specifications

[[Page 15152]]

Task Force (TSTF) Traveler TSTF-522-A, Revision 0, ``Revise Ventilation 
System Surveillance Requirements to Operate for 10 hours per Month.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change replaces existing Surveillance Requirements 
to operate the Control Room Emergency Ventilation System (CREVS) and 
the Emergency Exhaust System (EES) for a continuous 10 hour period 
with applicable heaters operating every 31 days, with requirements 
to operate these systems for 15 continuous minutes with applicable 
heaters operating every 31 days.
    These systems are not accident initiators (i.e., their 
malfunction cannot initiate an accident or transient) and therefore, 
these changes do not involve a significant increase in the 
probability of an accident. The proposed system and filter testing 
changes are consistent with current regulatory guidance for these 
systems and will continue to assure that these systems perform their 
design function which may include mitigating accidents. Therefore, 
the change does not involve a significant increase in the 
consequences of an accident.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The change proposed for these ventilation systems does not 
change any system operations or maintenance activities. Testing 
requirements will be revised and will continue to demonstrate that 
the Limiting Conditions for Operation are met and the system 
components are capable of performing their intended safety 
functions. The change does not create new failure modes or 
mechanisms and no new accident precursors are generated.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The design basis for the ventilation system heaters in the EES 
and in the pressurization trains of the CREVS includes the 
capability to heat the incoming air, reducing the relative humidity 
(and thereby increasing adsorber efficiency). The heater testing 
change proposed will continue to demonstrate that the heaters are 
capable of heating the air and will thus perform their design 
function. The proposed change is consistent with regulatory 
guidance.
    Therefore, it is concluded that this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the NRC's Public Document Room (PDR), located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible 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.
    Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone 
Power Station, Unit 3, New London County, Connecticut
    Date of amendment request: October 4, 2012, as supplemented by 
letters dated January 4, April 17, and October 30, 2013.
    Description of amendment request: The proposed amendment would 
modify Technical Specifications by relocating specific surveillance 
frequencies to a licensee controlled program with the adoption of 
Technical Specification Task Force (TSTF)-425, Revision 3, ``Relocate 
Surveillance Frequencies to Licensee Control--[Risk-Informed Technical 
Specification Task Force (RITSTF)] Initiative 5b.'' Additionally, the 
change would add a new program, the Surveillance Frequency Control 
Program (SFCP), to Technical Specification Section 6, Administrative 
Controls.
    Date of issuance: February 25, 2014.
    Effective date: As of the date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 258.
    Renewed Facility Operating License No. NPF-49: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: December 11, 2012 (77 
FR 73687).
    The supplemental letters dated January 4, 2013, April 17, 2013, and 
October 30, 2013, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 25, 2014.
    No significant hazards consideration comments received: No.
    Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

[[Page 15153]]

    Date of application for amendments: April 16, 2013.
    Brief description of amendments: The amendments remove superseded 
temporary Technical Specification (TS) requirements for McGuire Nuclear 
Station (MNS), Units 1 and 2, in accordance with a licensee commitment 
described in a May 28, 2010, license amendment request.
    Date of issuance: February 28, 2014.
    Effective date: This license amendment is effective as of its date 
of issuance and shall be implemented within 60 days of issuance.
    Amendment Nos.: 272 and 252.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses and technical specifications.
    Date of initial notice in Federal Register: June 25, 2013 (78 FR 
38081).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 28, 2014.
    No significant hazards consideration comments received: No.
    Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
    Date of application for amendments: February 22, 2013, as 
supplemented on September 10, October 25, November 29, and December 16, 
2013.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.4.3, to replace its current reactor coolant system 
pressure-temperature (P-T) limits with new P-T limits applicable to 54 
effective full power years. In addition, the amendments change the 
operational requirements for unit heatup and cooldown in TS Tables 
3.4.3-1 and 3.4.3-2.
    Date of Issuance: February 27, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 384, 386, and 385.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the license and the TSs.
    Date of initial notice in Federal Register: April 16, 2013, 78 FR 
22568.
    The supplemental letters dated September 10, October 25, November 
29, and December 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 amendments is contained 
in a Safety Evaluation dated February 27, 2014.
    No significant hazards consideration comments received: No.
    Duke Energy Progress Inc., Docket Nos. 50-325 and 50-324, Brunswick 
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina.
    Date of application for amendments: June 19, 2012, as supplemented 
by letters dated January 21, May 14, and August 29, 2013, and January 
22, 2014.
    Brief description of amendments: The amendments revised the 
Technical Specification (TS) to extend the Completion Time (CT) of TS 
3.8.1 Required Action D.4 for an inoperable diesel generator. A 
commensurate change is also made to extend the maximum CT of TS 3.8.1 
Required Actions C.3 and D.4. The licensee will to add a supplemental 
AC power source (i.e., a supplemental diesel generator) with the 
capability to power any emergency bus within 1 hour from a Station 
Blackout event, and with the capacity to bring the affected unit to 
cold shutdown.
    Date of issuance: February 24, 2014.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from the 2014 Unit 1 refueling outage.
    Amendment Nos.: 264 and 292.
    Facility Operating License Nos. DPR-62 AND DPR-71: Amendments 
revised the License and TSs.
    Date of initial notice in Federal Register: October 16, 2013 (77 FR 
63346).
    The supplements dated January 21, May 14, and August 29, 2013, and 
January 22, 2014, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 24, 2014.
    No significant hazards consideration comments received: None.
    Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating, Unit 2, Westchester County, New York
    Date of application for amendment: February 6, 2013, as 
supplemented by letters dated July 9, 2013, October 3, 2013, and 
February 24, 2014.
    Brief description of amendment: The amendment changes the Technical 
Specifications by revising the reactor heatup and cooldown curves (also 
referred to as pressure-temperature (P-T) limits) and low temperature 
overpressure protection (LTOP) requirements to cover a lifetime burnup 
of 48 Effective Full Power Years (EFPY), which is an increase from the 
current value of 29.2 EFPY.
    Date of issuance: March 5, 2014.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 274.
    Facility Operating License No. DPR-26: The amendment revised the 
License and the Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2013 (78 FR 
19750).
    The supplemental letters dated July 9, 2013, October 3, 2013, and 
February 24, 2014, 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 March 5, 2014.
    No significant hazards consideration comments received: No.
    Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Nuclear Generating, Units 3 and 4, Miami-Dade County, 
Florida
    Date of application for amendment: March 22, 2013.
    Brief description of amendment: The amendments revised the 
Technical Specifications (TSs) to allow the use of Optimized ZIRLO\TM\ 
as an approved fuel rod cladding.
    Date of issuance: February 20, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 259 and 254.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the licenses and the TSs.
    Date of initial notice in Federal Register: August 20, 2013 (78 FR 
51219).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 20, 2014.
    No significant hazards consideration comments received: No.
    Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant, Units 1 and 2 (CPNPP), Somervell 
County, Texas
    Date of amendment request: August 29, 2013, as supplemented by 
letter dated February 19, 2014.

[[Page 15154]]

    Description of amendment request: The amendments revised Technical 
Specification (TS) 3.4.17, ``Steam Generator (SG) Tube Integrity,'' TS 
5.5.9, ``Unit 1 Model D76 and Unit 2 Model D5 Steam Generator (SG) 
Program,'' and TS 5.6.9, ``Unit 1 Model D76 and Unit 2 Model D5 Steam 
Generator Tube Inspection Report.'' The changes address implementation 
issues associated with inspection periods, and address other 
administrative changes and clarifications. The amendment is consistent 
with NRC-approved Technical Specifications Task Force (TSTF) change 
traveler TSTF-510, Revision 2, ``Revision to Steam Generator Program 
Inspection Frequencies and Tube Sample Selection,'' as part of the 
consolidated line item improvement process.
    The amendments also incorporated minor non-technical variations 
from the TS changes proposed in TSTF-510, Revision 2. The TSs for 
CPNPP, Units 1 and 2 utilize different numbering and titles than the 
Standard Technical Specifications on which TSTF-510, Revision 2, is 
based, since the steam generators for CPNPP, Units 1 and 2, are of 
different models. These differences are administrative in nature and do 
not affect the applicability of TSTF-510, Revision 2, to the TSs for 
CPNPP, Units 1 and 2.
    Date of issuance: February 27, 2014.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: Unit 1--161; Unit 2--161.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: October 1, 2013 (78 FR 
60324).
    The February 19, 2014, supplement did not expand the scope of the 
application as originally noticed, and did not change the NRC staff's 
initial proposed finding of no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 27, 2014.
    No significant hazards consideration comments received: No.
    NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, 
Unit. 1, Rockingham County, New Hampshire
    Date of amendment request: June 25, 2013.
    Description of amendment request: The amendment revised the 
Seabrook Technical Specifications (TS). Specifically, the amendment 
revised the TS to allow the use of Optimized ZIRLO\TM\ as an approved 
fuel rod cladding material.
    Date of issuance: March 5, 2014.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 139.
    Facility Operating License No. NPF-86: The amendment revised the 
License and TS.
    Date of initial notice in Federal Register: August 20, 2013 (78 FR 
51228).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 5, 2014.
    No significant hazards consideration comments received: No.
    Northern States Power Company--Minnesota (NSPM), Docket No. 50-263, 
Monticello Nuclear Generating Plant, Wright County, Minnesota
    Date of application for amendment: April 19, 2013.
    Brief description of amendment: The amendment allows NSPM to adopt 
the NRC's approved Technical Specifications Task Force (TSTF) Standard 
Technical Specifications Change Traveler TSTF-535, Revision 0, ``Revise 
Shutdown Margin Definition to Address Advanced Fuel Designs,'' dated 
August 8, 2011. The amendment modifies the Technical Specification 
definition of ``shutdown margin'' (SDM) to require calculation of the 
SDM at a reactor moderator temperature of 68[emsp14][deg]F or higher, 
representing the most reactive state throughout the operating cycle. 
This change addresses newer boiling-water reactor fuel designs which 
may be more reactive at shutdown temperatures above 68[emsp14][deg]F.
    Date of issuance: February 28, 2014.
    Effective date: This license amendment is effective as of the date 
of issuance and shall be implemented within 90 days from the date of 
issuance.
    Amendment No.: 179.
    Renewed Facility Operating License No. DPR-22: The amendment 
revises the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: September 3, 2013 (78 
FR 54285).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 28, 2014.
    No significant hazards consideration comments received: No.
    Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo 
County, California
    Date of amendment request: June 6, 2013.
    Description of amendment request: The amendments revised Technical 
Specification (TS) 3.7.10, ``Control Room Ventilation System (CRVS),'' 
and TS 5.6.5, ``Core Operating Limits Report (COLR),'' to incorporate 
editorial changes. Specifically, the proposed amendments delete 
footnote (1) from the TS 3.7.10 Condition A Completion Time, and revise 
inconsistent wording in TS 5.6.5a.4, TS 5.6.5a.5, and TS 5.6.5a.9.
    Date of issuance: February 27, 2014.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: Unit 1--217; Unit 2--219.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2013 (78 FR 
47791).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 27, 2014.
    No significant hazards consideration comments received: No.
    PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna 
Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
    Date of application for amendments: June 6, 2013, as supplemented 
by letter dated December 4, 2013.
    Brief description of amendments: The amendments change the 
Technical Specifications (TSs) for Susquehanna Steam Electric Station, 
Units 1 and 2. Specifically, these amendments change TS 3.3.6.1, 
``Primary Containment Isolation Instrumentation,'' to add a footnote to 
Function 6.c. in TS Table 3.3.6.1-1, allowing only one Trip System to 
be operable in MODES 4 and 5 for the Manual Initiation Function for 
Shutdown Cooling System isolation.
    Date of issuance: February 26, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 259 and 240.
    Renewed Facility Operating License Nos. NPF-14 and NPF-22: The 
amendments revised the license and the TS.
    Date of initial notice in Federal Register: December 10, 2013 (78 
FR 74184).
    The supplemental letter dated December 4, 2013, provided additional 
information that clarified the

[[Page 15155]]

application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 26, 2014.
    No significant hazards consideration comments received: No.
    South Carolina Electric and Gas Company, South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, 
Unit, Fairfield County, South Carolina
    Date of application for amendment: April 2, 2013 as supplemented by 
letter dated May 16, 2013.
    Brief description of amendment: This amendment revises the 
Technical Specifications requirements regarding steam generator tube 
inspections and reporting as described in TSTF-510, Revision 2, 
``Revision to Steam Generator Program Inspection Frequencies and Tube 
Sample Selection.''
    Date of issuance: February 28, 2014.
    Effective date: This license amendment is effective as of the date 
of its issuance.
    Amendment No.: 196.
    Renewed Facility Operating License No. NPF-12: Amendment revises 
the License.
    Date of initial notice in Federal Register: June 25, 2013 (78 FR 
38083).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 28, 2014.
    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, 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's 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 ``Agency Rules of Practice and Procedure'' in 10 CFR 
Part 2. Interested person(s) should consult a current copy of 10 CFR 
2.309, which is available at the NRC's PDR, located at One White Flint 
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852, and electronically on the Internet at the NRC's 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

[[Page 15156]]

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 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's E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on 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 the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC's guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern

[[Page 15157]]

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.
    Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
    Date of amendment request: February 17, 2014.
    Description of amendment request: The amendments revise Technical 
Specification (TS) Table 3.3.4-1, Remote Shutdown System 
Instrumentation and Controls as a result of an inoperable 
instrumentation function on Unit 2. Table 3.3.4-1 specifies 
requirements for Function 3.b., Decay Heat Removal via Steam Generators 
(SGs)--Reactor Coolant System (RCS) Cold Leg Temperature--Loop A and B 
as ``1 per loop''. Loop A of this function is presently inoperable on 
Unit 2 due to a failed resistance temperature detector (RTD). Loop B of 
this function is operable with a reliable maintenance history. The 
failed RTD on Loop A cannot be replaced in the present operating mode 
of Unit 2 (Mode 1). Therefore, Duke Energy requested the U.S. Nuclear 
Regulatory Commission (NRC) approval to allow Unit 2 to remain in Mode 
1 until such time that the failed RTD can be replaced. The replacement 
would occur in the next refueling outage or the next outage that would 
facilitate replacement, whichever occurs first.
    Date of issuance: February 27, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 272 and 268.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: 
Amendments revised the licenses and the technical specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes. The NRC staff noticed the February 17, 2014, 
application in the Rock Hill, SC local newspaper, The Herald on Friday, 
February 21, 2014, and Saturday, February 22, 2014. The notice provided 
an opportunity to submit comments on the Commission's proposed NSHC 
determination. No comments have been received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, state consultation, and final NSHC determination 
are contained in a safety evaluation dated February 27, 2014.
    Attorney for licensee: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Robert J. Pascarelli.

    Dated at Rockville, Maryland, this 10th day of March 2014.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2014-05645 Filed 3-17-14; 8:45 am]
BILLING CODE 7590-01-P