[Federal Register Volume 79, Number 169 (Tuesday, September 2, 2014)]
[Notices]
[Pages 52059-52072]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-20671]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0193]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of 
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission 
(NRC) is publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 7, 2014 to August 20, 2014. The last 
biweekly notice was published on August 19, 2014.

DATES: Comments must be filed by October 2, 2014. A request for a 
hearing must be filed by November 3, 2014.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0193. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: 3WFN-06-A44M, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Angela Baxter, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-2976, email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0193 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0193.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2014-0193 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 http://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.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of Title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, or (2)

[[Page 52060]]

create the possibility of a new or different kind of accident from any 
accident previously evaluated; or (3) involve a significant reduction 
in a margin of safety. The basis for this proposed determination for 
each amendment request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

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

    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR Part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC's regulations are accessible electronically from the NRC Library on 
the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed by the above date, the Commission or a presiding officer 
designated by the Commission or by the Chief Administrative Judge of 
the Atomic Safety and Licensing Board Panel, will rule on the request 
and/or petition; and the Secretary or the Chief Administrative Judge of 
the Atomic Safety and Licensing Board will issue a notice of a hearing 
or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten 10 days prior to the filing deadline, the participant should 
contact the Office of the Secretary by email at [email protected], 
or by telephone at 301-415-1677, to request (1) a digital 
identification (ID) certificate, which allows the participant (or its 
counsel or representative) to digitally sign 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.

[[Page 52061]]

    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at http://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to 
[email protected], or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the 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 in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Duke Energy Progress Inc., Docket No. 50-400, Shearon Harris Nuclear 
Power Plant, Unit 1, New Hill, North Carolina
    Date of amendment request: June 19, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14174A118.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.2, ``Engineered Safety Features 
Actuation System Instrumentation,'' Table 3.3-4, ``Engineered Safety 
Features Actuation System Instrumentation Trip Setpoints.'' 
Specifically, the instrument trip setpoint and associated allowable 
value are being revised to ensure that the trip of the safety-related 
alternating current bus will occur at a voltage at or above the minimum 
voltage necessary to operate the applicable safety-related loads.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the TS Table 3.3-4 Functional Unit 
9.a, Loss-of-Offsite Power 6.9 kV Emergency Bus Undervoltage--
Primary, instrumentation trip setpoint and

[[Page 52062]]

allowable value. The Loss-of-Offsite Power, 6.9 kV Emergency Bus 
Undervoltage--Primary instrumentation is not an initiator to any 
accident previously evaluated. As such, the probability of an 
accident previously evaluated is not increased. The Loss-of-Offsite 
Power, 6.9 kV Emergency Bus Undervoltage--Primary instrumentation 
revised values continue to provide reasonable assurance that the 
Functional Unit 9.a will continue to perform its intended safety 
functions. As a result, the proposed change will not increase the 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the TS Table 3.3-4 Functional Unit 
9.a, Loss-of-Offsite Power 6.9 kV Emergency Bus Undervoltage--
Primary, instrumentation trip setpoint and allowable value. No new 
operational conditions beyond those currently allowed are 
introduced. This change is consistent with the safety analyses 
assumptions and current plant operating practices. This simply 
corrects the setpoint consistent with the accident analyses and 
therefore cannot create the possibility of a new or different kind 
of accident from any previously evaluated accident.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises the TS Table 3.3-4 Functional Unit 
9.a, Loss-of-Offsite Power 6.9 kV Emergency Bus Undervoltage--
Primary, instrumentation trip setpoint and allowable value. Function 
9.a protects the emergency power system against loss of voltage. 
This change is consistent with the safety analyses assumptions and 
current plant operating practices. No new operational conditions 
beyond those currently allowed are created by these changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, 
Charlotte, NC 28202.
    NRC Acting Branch Chief: Lisa M. Regner.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan
    Date of amendment request: June 11, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14162A079.
    Description of amendment request: The proposed amendment would 
modify technical specification (TS) requirements to adopt the changes 
described in TS Task Force (TSTF)-426, Revision 5, ``Revise or Add 
Actions to Preclude Entry into LCO [limiting condition for operation] 
3.0.3--RITSTF [Risk-Informed TSTF] Initiatives 6b & 6c'' (ADAMS 
Accession No. ML113260461).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change provides a short Completion Time to restore 
an inoperable system for conditions under which the existing 
Technical Specifications require a plant shutdown to begin within 1 
hour in accordance with LCO 3.0.3. Entering into Technical 
Specification Actions is not an initiator of any accident previously 
evaluated. As a result, the probability of an accident previously 
evaluated is not significantly increased. The consequences of any 
accident previously evaluated that may occur during the proposed 
Completion Times are no different from the consequences of the same 
accident during the existing 1 hour allowance. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different type of accident from any accident previously evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change increases the time the plant may operate 
without the ability to perform an assumed safety function. The 
analysis in WCAP-16125-NP-A, ``Justification for Risk-Informed 
Modifications to Selected Technical Specifications for Conditions 
Leading to Exigent Plant Shutdown,'' Revision 2, August 2010, 
demonstrated that there is an acceptably small increase in risk due 
to a limited period of continued operation in these conditions and 
that the risk is balanced by avoiding the risks associated with a 
plant shutdown. As a result, the change to the margin of safety 
provided by requiring a plant shutdown within 1 hour is not 
significant.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Branch Chief: David L. Pelton.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania
    Date of amendment request: July 11, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14192B143.
    Description of amendment request: The proposed amendment would 
incorporate several miscellaneous administrative changes to the 
Facility Operating License and the Technical Specifications. For 
example, the amendment would delete historical items that are no longer 
applicable, correct errors, and remove references that are no longer 
valid.
    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.
    No physical changes to the facility will occur as a result of 
this proposed amendment. The proposed changes will not alter the 
physical design or operational procedures associated with any plant 
structure, system, or component. The proposed changes are 
administrative in nature and have no effect on plant operation.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?

[[Page 52063]]

    Response: No.
    The proposed changes are administrative in nature. The proposed 
changes do not alter the physical design, safety limits, or safety 
analysis assumptions associated with the operation of the plant. 
Accordingly, the changes do not introduce any new accident 
initiators, nor do they reduce or adversely affect the capabilities 
of any plant structure, system, or component to perform their safety 
function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes conform to NRC regulatory guidance 
regarding the content of plant Technical Specifications. The 
proposed changes are administrative in nature. The proposed changes 
do not alter the physical design, safety limits, or safety analysis 
assumptions associated with the operation of the plant.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for Licensee: J. Bradley Fewell, Vice President and Deputy 
General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, 
Kennett Square, PA 19348.
    Acting NRC Branch Chief: Robert G. Schaaf.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania
    Date of application for amendments: July 10, 2014. A publicly-
available version is in ADAMS under Accession No. ML14191B190.
    Description of amendment request: The proposed amendment would 
revise and add Technical Specification (TS) surveillance requirements 
to address the concerns discussed in NRC Generic Letter 2008-01, 
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat 
Removal, and Containment Spray Systems,'' dated January 11, 2008 (ADAMS 
Accession No. ML072910759). The proposed TS changes are based on NRC-
approved TS Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic 
Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013 
(ADAMS Accession No. ML13053A075). The NRC staff issued a Notice of 
Availability for TSTF-523, Revision 2, for plant-specific adoption 
using the Consolidated Line Item Improvement Process, in the Federal 
Register on January 15, 2014 (79 FR 2700).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises or adds Surveillance Requirements 
(SRs) that require verification that the Emergency Core Cooling 
Systems, the Suppression Pool Cooling System, the Suppression Pool 
Spray System, the Drywell Spray System, the Shutdown Cooling System, 
and the Reactor Core Isolation Cooling System are not rendered 
inoperable due to accumulated gas and to provide allowances which 
permit performance of the revised verification. Gas accumulation in 
the subject systems is not an initiator of any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. The proposed SRs ensure 
that the subject systems continue to be capable of performing their 
assumed safety function and are not rendered inoperable due to gas 
accumulation. Thus, the consequences of any accident previously 
evaluated are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the Emergency Core Cooling Systems, the 
Suppression Pool Cooling System, the Suppression Pool Spray System, 
the Drywell Spray System, the Shutdown Cooling System, and the 
Reactor Core Isolation Cooling System are not rendered inoperable 
due to accumulated gas and to provide allowances which permit 
performance of the revised verification. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. In addition, the proposed 
change does not impose any new or different requirements that could 
initiate an accident. The proposed change does not alter assumptions 
made in the safety analysis and is consistent with the safety 
analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the Emergency Core Cooling Systems, the 
Suppression Pool Cooling System, the Suppression Pool Spray System, 
the Drywell Spray System, the Shutdown Cooling System, and the 
Reactor Core Isolation Cooling System are not rendered inoperable 
due to accumulated gas and to provide allowances which permit 
performance of the revised verification. The proposed change adds 
new requirements to manage gas accumulation in order to ensure the 
subject systems are capable of performing their assumed safety 
functions. The proposed SRs are more comprehensive than the current 
SRs and will ensure that the assumptions of the safety analysis are 
protected. The proposed change does not adversely affect any current 
plant safety margins or the reliability of the equipment assumed in 
the safety analysis. Therefore, there are no changes being made to 
any safety analysis assumptions, safety limits or limiting safety 
system settings that would adversely affect plant safety as a result 
of the proposed change.
    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: J. Bradley Fewell, Esquire, Vice President 
and Deputy General Counsel, Exelon Generation Company, LLC, 200 Exelon 
Way, Kennett Square, PA 19348.
    Acting NRC Branch Chief: Robert G. Schaaf.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
    Date of amendment request: July 10, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14191A059.
    Description of amendment request: The proposed amendment would 
revise and add Technical Specification (TS) Surveillance Requirements 
to address the concerns discussed in NRC Generic Letter 2008-01, 
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat 
Removal, and Containment Spray Systems,'' dated January 11, 2008 (ADAMS 
Accession No. ML072910759). The proposed TS changes are based on NRC-
approved TS Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic 
Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013 
(ADAMS Accession

[[Page 52064]]

No. ML13053A075). The NRC staff issued a Notice of Availability for 
TSTF-523, Revision 2, for plant-specific adoption using the 
Consolidated Line Item Improvement Process, in the Federal Register on 
January 15, 2014 (79 FR 2700).
    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 adds Surveillance Requirements (SRs) that 
require verification that the Emergency Core Cooling System (ECCS), 
the Decay Heat Removal (DHR) System, and the Reactor Building Spray 
(RB Spray) System are not rendered inoperable due to accumulated gas 
and to provide allowances which permit performance of the revised 
verification. Gas accumulation in the subject systems is not an 
initiator of any accident previously evaluated. As a result, the 
probability of any accident previously evaluated is not 
significantly increased. The proposed SRs ensure that the subject 
systems continue to be capable of performing their assumed safety 
function and are not rendered inoperable due to gas accumulation. 
Thus, the consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. 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 adds SRs that require verification that the 
ECCS, the DHR, and the RB Spray System are not rendered inoperable 
due to accumulated gas and to provide allowances which permit 
performance of the revised verification. The proposed change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. In addition, the proposed 
change does not impose any new or different requirements that could 
initiate an accident. The proposed change does not alter assumptions 
made in the safety analysis and is consistent with the safety 
analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change adds SRs that require verification that the 
ECCS, the DHR, and the RB Spray System are not rendered inoperable 
due to accumulated gas and to provide allowances which permit 
performance of the revised verification. The proposed change adds 
new requirements to manage gas accumulation in order to ensure that 
the subject systems are capable of performing their assumed safety 
functions. The proposed SRs are more comprehensive than the current 
SRs and will ensure that the assumptions of the safety analysis are 
protected. The proposed change does not adversely affect any current 
plant safety margins or the reliability of the equipment assumed in 
the safety analysis. Therefore, there are no changes being made to 
any safety analysis assumptions, safety limits, or limiting safety 
system settings that would adversely affect plant safety as a result 
of the proposed change.
    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: J. Bradley Fewell, Vice President and Deputy 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    Acting NRC Branch Chief: Robert G. Schaaf.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
    Date of amendment request: July 10, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14191B180.
    Description of amendment request: The proposed amendment would 
revise and add Technical Specification (TS) surveillance requirements 
to address the concerns discussed in NRC Generic Letter 2008-01, 
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat 
Removal, and Containment Spray Systems,'' dated January 11, 2008 (ADAMS 
Accession No. ML072910759). The proposed TS changes are based on NRC-
approved TS Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic 
Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013 
(ADAMS Accession No. ML13053A075). The NRC staff issued a Notice of 
Availability for TSTF-523, Revision 2, for plant-specific adoption 
using the Consolidated Line Item Improvement Process, in the Federal 
Register on January 15, 2014 (79 FR 2700).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises or adds Surveillance Requirements 
(SRs) that require verification that the Emergency Core Cooling 
System (ECCS), the Residual Heat Removal (RHR) System, the Shutdown 
Cooling (SDC) System, the Containment Spray (CS) System, and the 
Reactor Core Isolation Cooling (RCIC) System are not rendered 
inoperable due to accumulated gas and to provide allowances which 
permit performance of the revised verification. Gas accumulation in 
the subject systems is not an initiator of any accident previously 
evaluated. As a result, the probability of any accident previously 
evaluated is not significantly increased. The proposed SRs ensure 
that the subject systems continue to be capable of performing their 
assumed safety function and are not rendered inoperable due to gas 
accumulation. Thus, the consequences of any accident previously 
evaluated are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, the RHR, the SDC, the CS, and the RCIC 
Systems are not rendered inoperable due to accumulated gas and to 
provide allowances which permit performance of the revised 
verification. The proposed change does not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. In addition, the proposed change does not impose any new 
or different requirements that could initiate an accident. The 
proposed change does not alter assumptions made in the safety 
analysis and is consistent with the safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, the RHR, the SDC, the CS, and the RCIC 
Systems are not rendered inoperable due to accumulated gas and to 
provide allowances which permit performance of the revised 
verification. The proposed change revises or adds new requirements 
to manage gas accumulation in order to ensure the subject

[[Page 52065]]

systems are capable of performing their assumed safety functions. 
The proposed SRs are more comprehensive than the current SRs and 
will ensure that the assumptions of the safety analysis are 
protected. The proposed change does not adversely affect any current 
plant safety margins or the reliability of the equipment assumed in 
the safety analysis. Therefore, there are no changes being made to 
any safety analysis assumptions, safety limits or limiting safety 
system settings that would adversely affect plant safety as a result 
of the proposed change.
    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: J. Bradley Fewell, Vice President and Deputy 
General Counsel, Exelon Generation Company, LLC, 200 Exelon Way, 
Kennett Square, PA 19348.
    Acting NRC Branch Chief: Robert G. Schaaf.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and 2, 
Ogle County, Illinois
    Date of amendment request: April 24, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14120A039.
    Description of amendment request: The proposed amendment would add 
new ``low degraded voltage relays'' and timers, with appropriate 
settings, on each engineered safety feature electrical bus. The 
technical specifications and surveillance requirements would be changed 
to add appropriate operational and testing requirements for the new 
relays and timers.
    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, with NRC staff revisions provided in [brackets], which 
is presented below:

    EGC [Exelon Generation Company, LLC] has evaluated the proposed 
change for Braidwood Station and Byron Station, using the criteria 
in 10 CFR 50.92, and has determined that the proposed change does 
not involve a significant hazards consideration. The following 
information is provided to support a finding of no significant 
hazards consideration.
    Criteria
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to add new ``low degraded voltage relays'' 
(LDVRs) and associated CHANNEL CALIBRATION surveillance test 
provides a third level of undervoltage protection for the Engineered 
Safeguards Features (ESF) electrical buses. These new relays will 
further ensure that the normally operating safety-related motors/
equipment, which are powered from the ESF buses, are appropriately 
isolated from the normal off-site power source and will not be 
damaged in the event of sustained degraded bus voltage. The addition 
of the LDVRs will continue to allow the existing undervoltage 
protection circuitry to function as originally designed; i.e., the 
first-level ``loss of voltage'' protection and the second-level 
``degraded voltage'' protection will remain in place and be 
unaffected by this change. The proposed change does not affect the 
probability of any accident resulting in a loss of voltage or 
degraded voltage condition on the ESF electrical buses; and will 
positively impact the consequences of accidents previously evaluated 
as this change further ensures continued operation of safety-related 
equipment throughout the accident scenarios.
    Specific analysis was performed and determined that the proposed 
LDVRs, with the specified allowable values and time delay, will 
ensure that the 4.16 kV ESF buses will be isolated from the normal 
off-site power source, at the appropriate voltage level, under 
nonaccident sustained degraded voltage conditions. The normally 
operating safety related motors will be subsequently sequenced back 
on to the 4.16 kV ESF buses powered by the EDGs [Emergency Diesel 
Generators]; and therefore, will not be damaged in the event of 
sustained degraded bus voltage during the time delay period prior to 
initiation of the first level loss of voltage trip function.
    Therefore, these safety-related loads will be available to 
perform their design basis function should a loss-of-coolant 
accident (LOCA) occur concurrent with a loss-of-offsite power (LOOP) 
following the degraded voltage condition. The loading sequence 
(i.e., timing) of safety-related equipment back onto the ESF bus, 
powered by the EDG, is not affected by the addition of the new 
LDVRs.
    The addition of new LDVRs will have no impact on accident 
initiators or precursors; does not alter the accident analysis 
assumptions or the manner in which the plant is operated or 
maintained; and does not affect the probability of operator error.
    Based on the above, 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 involves the addition of new ``low degraded 
voltage relays'' (LDVRs); i.e., a third level of undervoltage 
protection for the ESF electrical buses, and adds an associated 
CHANNEL CALIBRATION surveillance test. This change helps ensure that 
the assumptions in the previously evaluated accidents, which may 
involve a degraded voltage condition, continue to be valid.
    The proposed changes do not result in the creation of any new 
accident precursors; do not result in changes to any existing 
accident scenarios; and do not introduce any operational changes or 
mechanisms that would create the possibility of a new or different 
kind of accident. A specific failure mode and effects review was 
completed for the new LDVRs, considering their potential failure, 
and concluded that the addition of these relays would not affect the 
existing ``loss of voltage'' and ``degraded voltage'' protection 
schemes; would not affect the number of occurrences of degraded 
voltage conditions that would cause the actuation of the existing 
Loss of Voltage Relays (LVRs), Degraded Voltage Relays (DVRs) or new 
LVDRs; would not affect the failure rate of the existing protection 
relays; and would not impact the assumptions in any existing 
accident scenario.
    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 current ``loss of voltage'' and ``degraded voltage'' 
protection circuitry is designed to appropriately isolate the 
normally operating safety-related motors/equipment, which are 
powered from the ESF buses, from the normal off-site power source 
such that the subject equipment will not be damaged in the event of 
sustained degraded bus voltage. The loss of voltage relays (LVRs) 
isolate the ESF buses at a TS [technical specifications] voltage 
value of approximately 66% of the nominal bus value after a short 
time delay (i.e., 1.9 seconds); while the degraded voltage relays 
(DVRs) isolate the ESF buses at a TS voltage value of 94.5% for 
Braidwood (91.2% for Byron Station) of the nominal bus voltage after 
a longer time delay of up to 5 minutes and 40 seconds (if no safety 
injection signal is present). After the ESF buses are isolated from 
the offsite power supply, the normally operating safety related 
motors will be sequenced back on to the 4.16 kV EFS bus powered by 
the EDG; and continue to perform their design basis function to 
mitigate the consequences of an accident, with a specified margin of 
safety.
    A concern exists that ESF motors/equipment may be damaged when 
operating and/or starting safety-related equipment when bus voltage 
drops to just above the loss of voltage relay setpoint for the 
duration of the 5 minutes and 40 second time delay. The new LDVRs 
are being added to resolve this concern. Analysis has been performed 
that shows the ESF equipment will not be damaged at 75% of bus 
voltage; therefore, the LDVR setpoint will be set at 75% of nominal 
ESF bus voltage. With the addition of this new third level of 
undervoltage protection,

[[Page 52066]]

the capability of the ESF equipment will be assured; and thus the 
equipment will continue to perform its design basis function to 
mitigate the consequences of the previously analyzed accidents; and 
maintain the existing margin to safety currently assumed in the 
accident analyses.
    An EDG start due to a safety injection signal (i.e., Loss of 
Coolant Accident) and the subsequent sequencing of ESF loads back on 
to the ESF buses, powered by the EDG, is not adversely affected by 
this change. If an actual loss of voltage condition occurs on the 
ESF buses, the loss of voltage time delays will continue to isolate 
the 4.16 kV ESF distribution system from the offsite power source 
prior to the EDG assuming the ESF loads.
    The ESF loads will sequence back on to the bus in a specified 
order and time interval; again ensuring that the existing accident 
analysis assumptions remain valid and the existing margin to safety 
is unaffected.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, EGC concludes that the proposed amendments 
do not involve a 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Travis L. Tate.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
1, Rockingham County, New Hampshire
    Date of amendment request: June 24, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14177A503.
    Description of amendment request: The proposed amendment would 
revise and add Technical Specification (TS) Surveillance Requirements 
(SRs) to address the concerns discussed in NRC Generic Letter 2008-01, 
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat 
Removal, and Containment Spray Systems,'' dated January 11, 2008 (ADAMS 
Accession No. ML072910759). The proposed TS changes are based on NRC-
approved TS Task Force (TSTF) Traveler TSTF-523, Revision 2, ``Generic 
Letter 2008-01, Managing Gas Accumulation,'' dated February 21, 2013 
(ADAMS Accession No. ML13053A075). The NRC staff issued a Notice of 
Availability for TSTF-523, Revision 2, for plant-specific adoption 
using the Consolidated Line Item Improvement Process, in the Federal 
Register on January 15, 2014 (79 FR 2700).
    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, with NRC staff revisions provided in [brackets], which 
is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the Emergency Core Cooling Systems (ECCS), 
Residual Heat Removal (RHR) System, and Containment Spray (CS) 
System are not rendered inoperable due to accumulated gas and to 
provide allowances which permit performance of the revised 
verification. Gas accumulation in the subject systems is not an 
initiator of any accident previously evaluated. As a result, the 
probability of any accident previously evaluated is not 
significantly increased. The proposed SRs ensure that the subject 
systems continue to be capable to perform their assumed safety 
function and are not rendered inoperable due to gas accumulation. 
Thus, the consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, RHR System, and CS System are not 
rendered inoperable due to accumulated gas and to provide allowances 
which permit performance of the revised verification. The proposed 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or a change in 
the methods governing normal plant operation. In addition, the 
proposed change does not impose any new or different requirements 
that could initiate an accident. The proposed change does not alter 
assumptions made in the safety analysis and is consistent with the 
safety analysis assumptions[.]
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, RHR System, and CS System are not 
rendered inoperable due to accumulated gas and to provide allowances 
which permit performance of the revised verification. The proposed 
change adds new requirements to manage gas accumulation in order to 
ensure that the subject systems are capable of performing their 
assumed safety functions. The proposed SRs are more comprehensive 
than the current SRs and will ensure that the assumptions of the 
safety analysis are protected. The proposed change does not 
adversely affect any current plant safety margins or the reliability 
of the equipment assumed in the safety analysis. Therefore, there 
are no changes being made to any safety analysis assumptions, safety 
limits, or limiting safety system settings that would adversely 
affect plant safety as a result of the proposed change.
    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: James Petro, Managing Attorney, Florida 
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    Acting NRC Branch Chief: Robert G. Schaaf.
South Carolina Electric and Gas Docket Nos.: 52-027 and 52-028, Virgil 
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County, 
South Carolina
    Date of amendment request: May 20, 2014, as supplemented by letter 
dated June 3, 2014. Publicly-available versions are in ADAMS under 
Accession Nos. ML14140A637 and ML14155A257, respectively.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-93 and NPF-94 for VCSNS Units 2 and 3 by 
departing from the plant-specific Design Control Document (DCD) Tier 
1(and corresponding Combined License Appendix C information) material 
by making various nontechnical changes to correct editorial and 
consistency errors in Tier 1. This is being done to promote consistency 
within the Updated Final Safety Analysis Report (UFSAR).
    Because, this proposed change requires a departure from Tier 1 
information in the Westinghouse Advanced Passive 1000 DCD, the licensee 
also requested an exemption from the requirements of the Generic DCD 
Tier 1 in accordance with 10 CFR 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 52067]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, with NRC staff revisions provided in [brackets], 
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 editorial and consistency plant-specific Tier 1 and 
corresponding COL [combined operating license] Appendix C update 
does not involve a technical change, e.g., there is no design 
parameter or requirement, calculation, analysis, function or 
qualification change. No structure, system, or component (SSC) 
design or function would be affected. No design or safety analysis 
would be affected. The proposed changes do not affect any accident 
initiating event or component failure, thus the probabilities of the 
accidents previously evaluated are not affected. No function used to 
mitigate a radioactive material release and no radioactive material 
release source term is involved, thus the radiological releases in 
the accident analyses are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed editorial and consistency plant-specific Tier 1 and 
corresponding COL Appendix C update would not affect the design or 
function of any SSC, but will instead provide consistency between 
the SSC designs and functions currently presented in the UFSAR and 
the Tier 1 information. The proposed changes would not introduce a 
new failure mode, fault or sequence of events that could result in a 
radioactive material release.
    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 editorial and consistency plant-specific Tier 1 and 
corresponding COL Appendix C update is considered non-technical for 
reasons discussed above, thus would not affect any design parameter, 
function or analysis. There would be no change to an existing design 
basis, design function, regulatory criterion, or analysis. No safety 
analysis or design basis acceptance limit/criterion is involved.
    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: Kathryn M. Sutton, Morgan, Lewis & Bockius 
LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2 (VEGP), Burke 
County, Georgia
    Date of amendment request: August 31, 2012, as supplemented 
September 13, 2013, May 2, July 22, and August 11, 2014. Publicly-
available versions are in ADAMS under Accession Nos. ML12248A035, 
ML13256A306, ML14122A364, ML14203A252 and, ML14223A616, respectively.
    Description of amendment request: The proposed amendments would 
revise the licensing basis for the VEGP by adding license conditions 
that would allow for the voluntary implementation of 10 CFR 50.69, 
``Risk-informed categorization and treatment of structures, systems, 
and components for nuclear power reactors.'' As indicated in Sec.  
50.69, a licensee may voluntarily comply with Sec.  50.69 as an 
alternative to compliance with the following requirements for certain 
SSCs: (i) 10 CFR part 21, (ii) a portion of Sec.  50.46, (iii) Sec.  
50.49, (v) certain requirements of Sec.  50.55a, (vi) Sec.  50.65, 
(vii) Sec.  50. 72, (viii) Sec.  50.73,[middot](ix) Appendix B to Part 
50, (x) certain containment leakage testing requirements, and (xi) 
certain requirements of Appendix A to part 100.
    Basis for proposed no significant hazards consideration 
determination: The licensee responded in its letter dated August 11, 
2014, to the NRC staff's request for additional information regarding 
the licensee's no significant hazards consideration determination, 
which is required by 10 CFR 50.91(a). Portions of the licensee's 
response regarding each of the no significant hazards consideration 
standards, with NRC staff revisions provided in [brackets], are 
presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Operation of the Vogtle Electric Generating Plant (VEGP) in 
accordance with the proposed amendment does not result in a 
significant increase in the probability or consequences of accidents 
previously evaluated. The Updated Final Safety Analysis Report 
(UFSAR) documents the analysis of design basis accidents at VEGP. 
The proposed amendment does not affect accident initiators, nor does 
it alter design assumptions, conditions, or configurations of the 
facility that would increase the probability of accidents previously 
evaluated, nor does it adversely alter design assumptions, 
conditions, or configurations of the facility, and it does not 
adversely impact the ability of structures, systems, or components 
(SSCs) to perform their intended function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits, nor do they affect assumed failure modes for accidents 
described and evaluated in the UFSAR. The proposed changes do not 
affect the way in which required systems perform their functions as 
required by the accident analysis. Structures, systems, and 
components required to safely shut down the reactor and maintain it 
in a safe shutdown condition will remain capable of performing their 
design functions.
    Furthermore, the source term and radiological release 
assumptions of previously evaluated events are not affected by the 
alternative treatments permitted under 10 CFR 50.69; containment 
isolation devices assumed to function under accident conditions will 
not have their reliability adversely affected by the proposed 
amendment. Consequently, operating under the proposed amendment will 
not result in a significant increase in the radiological dose 
consequences assumed for previously analyzed events.
    Section 50.69 defines the terminology ``safety significant 
function'' as functions whose loss or degradation could have a 
significant adverse effect on defense-in-depth, safety margins, or 
risk. For SSCs determined to be safety significant, 50.69 maintains 
the current regulatory requirements. These current requirements are 
adequate for addressing design basis performance of these SSCs.
    The purpose of this amendment is to permit VEGP to adopt a new 
risk-informed licensing basis for categorization and treatment of 
structures, systems and components. The proposed VEGP Units 1 and 2 
OL [operating license] LCs [license conditions] will allow for the 
voluntary implementation of 10 CFR 50.69. The SNC [Southern Nuclear 
Operating Company] risk-informed categorization process has been 
documented per the requirements of 10 CFR 50.69(b)(2) and meets the 
requirements of 10 CFR 50.69(c). A probabilistic approach to 
regulation enhances and extends the traditional deterministic 
approach by allowing consideration of a broader set of potential 
challenges to safety and providing a logical means for prioritizing 
these challenges based on safety significance. The SNC risk-informed 
categorization process will be used to modify the scope of SSCs 
subject to special treatment requirements. Alternative treatments 
permitted per 10 CFR 50.69(b)(1) and 10 CFR 50.69(d)(2) can then be 
applied consistent with the categorization of the SSCs. The process 
provides reasonable confidence that, for SSCs categorized as RISC-3, 
sufficient safety margins are maintained and that any potential 
increases in CDF [core damage frequency] and LERF [large early 
release frequency] resulting from changes in treatment are small per 
10 CFR 50.69(c)(1)(iv). The proposed OL LCs do not result in or 
require any physical or operational changes to VEGP SSCs, including 
SSCs intended for the prevention or

[[Page 52068]]

mitigation of accidents. Implementation of 10 CFR 50.69 in 
compliance with 10 CFR 50.69 requirements ensures that RISC-1 and 
RISC-3 SSCs remain capable of performing their design basis 
functions, including safety-related functions, under design basis 
conditions. In addition, the process ensures that RISC-2 SSCs are 
capable of performing their safety significant functions.
    Based on the above, implementation of this amendment to 
implement 10 CFR 50.69 risk informed categorization and treatment of 
structures, systems, and components does not involve a significant 
increase in the probability of any accident previously evaluated. In 
addition, all equipment required to mitigate an accident remains 
capable of performing the assumed function.
    Therefore, consequences of any accident previously evaluated are 
not significantly increased with the implementation of this License 
Amendment.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Operation of VEGP in accordance with the proposed amendment does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated. The proposed amendment does 
not impact any scenario or previously analyzed accident with offsite 
dose consequences included in the evaluation of design basis 
accidents (DBA) documented in the FSAR [final safety analysis 
report]. The proposed change does not alter the requirements or 
functions for systems required during accident conditions, nor does 
it alter the required mitigation systems as assumed in the licensing 
basis analyses and/or DBA radiological consequences evaluations. 
Implementation of the 50.69 categorization will not result in new or 
different accidents.
    The proposed amendment does not adversely affect accident 
initiators nor alter design assumptions, or conditions of the 
facility. The proposed amendment does not introduce new or different 
accident initiators; neither does it introduce new modes of 
operation. The proposed amendment does not adversely affect the 
ability of SSCs to perform their design function. SSCs required to 
safely shutdown the reactor and maintain it in a safe shutdown 
condition remain capable of performing their design function.
    Section 50.69 represents an alternative set of requirements 
whereby a licensee may voluntarily undertake categorization of its 
SSCs consistent with the requirements in 50.69(c), remove the 
special treatment requirements listed in 50.69(b) for SSCs that are 
determined to be of low safety significance, and implement 
alternative treatment requirements in 50.69(d). The regulatory 
requirements not removed continue to apply. These requirements are 
adequate for addressing design basis performance of these SSCs. This 
license amendment continues to maintain the principles that the net 
increase in plant risk is small, defense-in-depth is maintained, and 
safety margins are maintained.
    The proposed VEGP Units 1 and 2 OL LCs will allow for the 
voluntary implementation of 10 CFR 50.69. The SNC risk-informed 
categorization process has been documented per the requirements of 
10 CFR 50.69(b)(2) and meets the requirements of 10 CFR 50.69(c). 
The SNC risk-informed categorization process will be used to modify 
the scope of SSCs subject to special treatment requirements. 
Alternative treatments permitted per 10 CFR 50.69(b)(1) and 10 CFR 
50.69(d)(2) can then be applied consistent with the categorization 
of the SSCs. The process provides reasonable confidence that, for 
SSCs categorized as RISC-3, sufficient safety margins are maintained 
and that any potential increases in CDF and LERF resulting from 
changes in treatment are small per 10 CFR 50.69(c)(1)(iv). The 
proposed OL LCs do not result in or require any physical or 
operational changes to VEGP SSCs, including SSCs intended for the 
prevention or mitigation of accidents. Implementation of 10 CFR 
50.69 in compliance with 10 CFR 50.69 requirements ensures that 
RISC-1 and RISC-3 SSCs remain capable of performing their design 
basis functions, including safety-related functions, under design 
basis conditions. In addition, the process ensures that RISC-2 SSCs 
are capable of performing their safety significant functions. 
Therefore, even though there was not an individual evaluation done 
of every UFSAR accident with potential off-site dose consequences, 
it can be concluded that the SSCs, assumed to mitigate the 
consequences of any and all previously evaluated events, will not be 
adversely affected by the alternative treatments allowed under 10 
CFR 50.69. Consequently, the dose consequences of previously 
analyzed events will not significantly increase as a result of the 
alternative treatment of SSCs. Additionally, implementation of 10 
CFR 50.69 will not create new failure mechanisms that initiate new 
accidents because the process does not result in or require any 
physical or operational changes for VEGP SSCs nor does it alter the 
functions or functional requirements of those SSCs.
    Based on this, implementation of the proposed amendment would 
not create the possibility of a new or different kind of accident 
from any kind of accident previously evaluated. No new accident 
scenarios, transient precursors, failure mechanisms, or limiting 
single failures will be introduced as a result of this amendment. 
There will be no adverse effect or challenges imposed on required 
systems as a result of this amendment.
    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.
    Operation of VEGP in accordance with the proposed amendment does 
not involve a significant reduction in the margin of safety. 
Implementation of a new risk informed categorization and treatment 
of structures, systems, and components licensing basis that complies 
with the requirements of 10 CFR 50.69 does not alter the manner in 
which safety limits, limiting safety system settings, or limiting 
conditions for operation are determined. The safety analysis 
acceptance criteria are not affected by this change. The proposed 
amendment does not adversely affect existing plant safety margins or 
the reliability of equipment assumed in the UFSAR to mitigate 
accidents. The proposed change does not adversely affect the ability 
of SSCs to perform their design function. The 10 CFR 50.69 process 
provides reasonable confidence that SSCs categorized as RISC-1, 
RISC-2, and RISC-3 maintain sufficient safety margins. The proposed 
amendment does not adversely impact systems required to safely 
shutdown the plant and maintain it in a safe condition.
    The proposed VEGP Units 1 and 2 OL LCs will allow for the 
voluntary implementation of 10 CFR 50.69. The SNC risk-informed 
categorization process has been documented per the requirements of 
10 CFR 50.69(b)(2) and meets the requirements of 10 CFR 50.69(c). 
The SNC risk-informed categorization process will be used to modify 
the scope of SSCs subject to special treatment requirements. 
Alternative treatments permitted per 10 CFR 50.69(b)(1) and 10 CFR 
50.69(d)(2) can then be applied consistent with the categorization 
of the SSCs. Although there were no calculations or evaluations 
performed for the express purpose of demonstrating that the 
implementation of 10 CFR 50.69 will not result in a significant 
reduction in the margin of safety, the process provides reasonable 
confidence that, for SSCs categorized as RISC-3, sufficient safety 
margins are maintained and that any potential increases in CDF and 
LERF resulting from changes in treatment are small per 10 CFR 
50.69(c)(1)(iv). The only requirements that are relaxed for SSCs, 
consistent with their categorization, are those related to 
treatment. The safety margins associated with SSCs design basis 
functions and design technical requirements remain unchanged. 
Additionally, it is required that there be reasonable confidence 
that any potential increases in CDF and LERF be small from assumed 
changes in reliability resulting from the treatment changes 
permitted by 10 CFR 50.69. As a result individual SSCs continue to 
be capable of performing their design basis functions. It is 
concluded that sufficient safety margins are preserved.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Leigh D. Perry, SVP & General Counsel, 
Southern Nuclear Operating Company, 40 Inverness Center Parkway, 
Birmingham, AL 35242.
    NRC Branch Chief: Robert Pascarelli.

[[Page 52069]]

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2 (HNP), Appling County, Georgia
    Date of amendment request: August 15, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14227A921.
    Description of amendment request: The proposed amendments would 
modify Technical Specification (TS) 3.8.7 to add two new safety-related 
instrument buses to the HNP electrical distribution system. Certain 
instruments will be re-located from existing safety-related electrical 
instrument buses to these new ``critical instrumentation buses.'' The 
existing instrument bus is listed in TS 3.8.7 of the HNP, Units 1 and 
2, TSs and, since some of the instruments powered from this bus will be 
moved to the critical instrumentation bus, the new bus will be added to 
the list of the existing electrical buses in TS 3.8.7.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided an analysis of the issue of no significant hazards 
consideration, with NRC staff revisions provided in [brackets], as 
presented below:

    Southern Nuclear Operating Company has evaluated whether or not 
a significant hazards consideration is involved with the proposed 
amendment by focusing on the three standards set forth in 10 CFR 
50.92, ``Issuance of Amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously 
identified?
    Response: No.
    These new critical instrumentation buses and their inverters are 
not intended for the prevention of any previously analyzed transient 
or accident. They are intended to provide power to instruments which 
may be necessary to aid the operator in the mitigation of a beyond 
design basis external event. The new critical instrumentation buses 
perform the same function as existing instrumentation buses except 
they will have the added capability of obtaining primary power from 
DC [direct current] through their inverters connected to the station 
service DC power supplies.
    The new equipment (inverters and critical instrumentation bus) 
will be installed as safety related, seismically and environmentally 
qualified equipment, with the primary power coming from the safety 
related DC station service buses, and alternate power available from 
the safety related AC [alternating current] essential cabinets. 
Therefore, the instruments being moved to the critical 
instrumentation bus will have a highly reliable source of power. 
Consequently, should the operator require the use of one of these 
instruments to aid in mitigating the consequences of a previously 
analyzed design basis event, it is highly likely that they will be 
available to him/her. It is therefore unlikely that the consequences 
of a previously evaluated accident would increase due to an 
inability to monitor a key containment parameter.
    The TSs are being revised to add these instrument buses to the 
LCO [limiting condition for operation] requirements for the 
electrical distribution buses. No other TS LCOs are changing, no 
Surveillance Requirements are changing, and no instrument setpoints 
are changing. In fact, this TS change does not reduce any 
requirements. All of the components required to be Operable by the 
TSs before this revision request, will be required to be Operable 
following this change, as well as the new critical instrumentation 
bus. The TS requirements will therefore remain the same for the 
instruments being powered from the new critical instrumentation bus 
as well as for the instruments remaining on the AC instrument buses. 
In other words, the power supplies for these instruments will still 
be included in the TS as LCO requirements, as they were before the 
design change to add the critical instrumentation buses. The TS 
requirements will therefore continue to ensure that these indicators 
remain Operable during design basis events.
    For the above reasons, revising the TS to include the new 
critical instrument buses in the electrical bus distribution 
Limiting Condition for Operation does not increase the probability, 
or consequences, of a previously analyzed event.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    TS LCO 3.8.7 is being changed to add the new critical 
instrumentation bus. No new modes of operation or new failure modes 
result from the actual TS change to any system intended for the 
prevention of accidents.
    The design function of the instruments being moved from the 
existing instrument buses to the critical instrumentation buses will 
not change. Also, the operation of these instruments during any type 
of event is not changing. Only their power supply is being changed 
and thus no new modes of operation are created for these 
instruments. It is true that new components are being introduced, 
i.e., the inverters and instrumentation buses, thus introducing a 
potential failure that would not be present before the modification. 
However, their failure cannot cause a new or different type of 
accident. Furthermore the addition of these instruments will not 
affect any other system intended for the prevention of accidents.
    The design change does not impact the existing essential 
cabinets or instrument buses, except to remove some loads from the 
instrument bus. Consequently, the design function, operation, 
maintenance, and testing of these existing power supplies will not 
change.
    Finally, the new inverters and the critical instrumentation 
buses are not potential accident initiators; they are not intended 
to prevent an accident in that they do not serve as a barrier to the 
release of radiation either from the direct fission product 
boundary, or from the containment. Rather, they are intended to 
power instruments which serve the operators in their attempt to 
mitigate the consequences of accidents. Therefore, failure of these 
power supplies, or failure of any instrument being powered from 
them, cannot create an accident.
    For the above reasons, the proposed amendment will not create 
the possibility of a new or different type of accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The new critical instrumentation buses being referenced in the 
TS will power several instruments currently being powered by the 
safety related instrument bus. The new inverters and critical 
instrumentation buses will also be safety related, as will their 
primary power source, the DC station service buses. Additionally, 
the inverters are alternately powered from the safety related 
essential cabinets. Therefore, because of the reliability and 
diversity of power supplies, the margin of safety of a loss of power 
event to the relocated instruments is not significantly reduced.
    Loading calculations confirm that adequate design margin still 
exists for the DC station service buses with respect to their 
loading for design basis events, even with the additional loads of 
the added instruments.
    Additionally, area heat load calculations were performed for the 
130 foot elevation of the Units 1 and 2 Control Buildings which 
account for the new inverters, instrumentation bus and supporting 
components. These calculations concluded that there are no adverse 
effects on the [Final Safety Analysis Report] FSAR design functions.
    Adding the critical instrumentation buses to the TS ensures that 
the new power supplies to the safety related instruments have the 
same TS requirements as their previous power supply. Therefore, no 
TS requirements have been eliminated or reduced.

    For the above reasons, the margin of safety is not significantly 
reduced.
    On the basis of the evaluation above provided by the licensee, the 
NRC staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Leigh D. Perry, SVP & General Counsel of 
Operations and Nuclear, Southern Nuclear Operating Company, Inc., 40 
Inverness Center Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Robert Pascarelli.

[[Page 52070]]

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, (NAPS) Louisa County, Virginia
    Date of amendment request: June 30, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14183B318.
    Description of amendment request: The proposed license amendment 
requests the changes to the Technical Specification (TS) TS 5.5.15, 
``Containment Leakage Rate Testing Program,'' by replacing the 
reference to Regulatory Guide (RG) 1.163, ``Performance-Based 
Containment Leak-Test Program,'' with a reference to Nuclear Energy 
Institute (NEI) topical report NEI 94-01, Revision 3-A, ``Industry 
Guideline for Implementing Performance-Based Option of 10 CFR Part 50, 
Appendix J,'' as the implementation document used to develop the North 
Anna performance-based leakage testing program in accordance with 
Option B of 10 CFR Part 50, Appendix J.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1--Does the proposed change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response: No.
    The proposed amendment involves changes to the NAPS Containment 
Leakage Rate Testing Program. The proposed amendment does not 
involve a physical change to the plant or a change in the manner in 
which the plant is operated or controlled. The primary containment 
function is to provide an essentially leak tight barrier against the 
uncontrolled release of radioactivity to the environment for 
postulated accidents. As such, the containment and the testing 
requirements to periodically demonstrate the integrity of the 
containment exist to ensure the plant's ability to mitigate the 
consequences of an accident, and do not involve any accident 
precursors or initiators.
    Therefore, the probability of occurrence of an accident 
previously evaluated is not significantly increased by the proposed 
amendment.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for development of the NAPS performance-based 
testing program. Implementation of these guidelines continues to 
provide adequate assurance that during design basis accidents, the 
primary containment and its components will limit leakage rates to 
less than the values assumed in the plant safety analyses. The 
potential consequences of extending the ILRT [integrated leak rate 
test] interval to 15 years have been evaluated by analyzing the 
resulting changes in risk. The increase in risk in terms of person-
rem per year within 50 miles resulting from design basis accidents 
was estimated to be acceptably small and determined to be within the 
guidelines published in RG 1.174 [``An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific changes to the Licensing Basis'']. Additionally, the 
proposed change maintains defense-in-depth by preserving a 
reasonable balance among prevention of core damage, prevention of 
containment failure, and consequence mitigation. NAPS has determined 
that the increase in Conditional Containment Failure Probability due 
to the proposed change is very small.
    Therefore, it is concluded that the proposed amendment does not 
significantly increase the consequences of an accident previously 
evaluated.
    Based on the above discussion, it is concluded that the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Criterion 2--Does the change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for the development of the NAPS performance-
based leakage testing program, and establishes a 15-year interval 
for the performance of the containment ILRT. The containment and the 
testing requirements to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident, do not involve any accident precursors 
or initiators. The proposed change does not involve a physical 
change to the plant (i.e., no new or different type of equipment 
will be installed) or a change to the manner in which the plant is 
operated or controlled.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    Criterion 3--Does this change involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, for the development of the NAPS performance-
based leakage testing program, and establishes a 15-year interval 
for the performance of the containment ILRT. This amendment does not 
alter the manner in which safety limits, limiting safety system 
setpoints, or limiting conditions for operation are determined. The 
specific requirements and conditions of the Containment Leakage Rate 
Testing Program, as defined in the TS, ensure that the degree of 
primary containment structural integrity and leak-tightness that is 
considered in the plant's safety analysis is maintained. The overall 
containment leakage rate limit specified by the TS is maintained, 
and the Type A, Type B, and Type C containment leakage tests will be 
performed at the frequencies established in accordance with the NRC-
accepted guidelines of NEI 94-01, Revision 3-A.
    Containment inspections performed in accordance with other plant 
programs serve to provide a high degree of assurance that the 
containment will not degrade in a manner that is not detectable by 
an ILRT. A risk assessment using the current NAPS PRA [probabilistic 
risk assessment] model concluded that extending the ILRT test 
interval from 10 years to 15 years results in a small change to the 
NAPS risk profile.

    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Robert Pascarelli.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.

[[Page 52071]]

    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.
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: December 16, 2011, as 
supplemented by letters dated January 20, March 1, March 16, April 18, 
July 11, July 20, August 31, and November 2, 2012; April 5, June 28, 
August 7, and December 18, 2013; and February 14, April 3, April 11, 
and July 24, 2014.
    Brief description of amendments: The amendments revised the 
Technical Specifications and the Updated Final Safety Analysis Report 
to add the new Protected Service Water (PSW) System to the plant's 
licensing basis as an additional method of achieving and maintaining 
safe shutdown of the reactors in the event of a high-energy line break 
or a fire in the turbine building, which is shared by all three units.
    Date of Issuance: August 13, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 386, 388, and 387. A publicly-available version is 
in ADAMS under Accession No. ML14206A790. Documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    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: July 10, 2012, (77 FR 
40652). The supplemental letters dated January 20, March 1, March 16, 
April 18, July 11, July 20, August 31, and November 2, 2012; April 5, 
June 28, August 7, and December 18, 2013; and February 14, April 3, 
April 11, and July 24, 2014, provided additional information that 
clarified the application, did not expand the scope of the application 
as noticed, and did not change the staff's 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 August 13, 2014.
    No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit 1, Washington County, Nebraska
    Date of amendment request: August 5, 2013, as supplemented by 
letter dated January 28, 2014.
    Brief description of amendment: The amendment revised the 
structural design basis related to the leak-before-break analysis for 
the reactor coolant system piping described in Section 4.3.6 of the 
Updated Safety Analysis Report.
    Date of issuance: August 7, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 276. A publicly-available version is in ADAMS under 
Accession No. ML14209A027; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-40: The amendment 
revised the design basis as described in the Updated Safety Analysis 
Report.
    Date of initial notice in Federal Register: April 8, 2014 (79 FR 
19400).
    The Commission's related evaluation of the amendment is contained 
in a safety evaluation dated August 7, 2014.
    No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama
    Date of application for amendment: August 20, 2012, as supplemented 
by letters dated October 25, 2012, November 8, 2012, July 2, 2013, and 
June 16, 2014.
    Brief description of amendments: The amendments revise the 
condensate storage tank level requirement specified in Technical 
Specification surveillance requirement 3.7.6.1.
    Date of issuance: August 15, 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--195, Unit 2--191. A publicly-available 
version is in ADAMS under Accession No. ML14155A302; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-2 and NPF-8: The amendments 
revised the Renewed Facility Operating Licenses and Technical 
Specifications.
    Date of initial notice in Federal Register: January 15, 2013 (78 FR 
3037). The supplemental letters dated October 25, 2012, November 8, 
2012, July 2, 2013, and June 16, 2014, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 15, 2014.
    No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units 1 and 2, Louisa County, Virginia and 
Docket No. 50-280 and 50-281, Surry Power Station, Units 1 and 2, Surry 
County, Virginia
    Date of application for amendment: June 26, 2013, as supplemented 
by letter dated January 23, 2014.
    Brief description of amendment: The license amendments approve the 
generic application of Appendix D, ``Qualification of the ABB-NV and 
WLOP Critical Heat Flux (CHF) Correlations in the Dominion VIPRE-D 
Computer Code,'' to Fleet Report DOM-NAF-2-A, ``Reactor Core Thermal-
Hydraulics Using the VIPRE-D Computer Code,'' the plant-specific 
applications of Appendix D to Fleet Report DOM-NAF-2-A to North Anna 
and Surry Power Stations, an added Surry reactor core safety limit, an 
increase in the Surry Minimum Temperature for Criticality (MTC), and 
modified references to MTC.
    Date of issuance: August 12, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 271, 253, 283, and 283. A publicly-available 
version is in ADAMS under Accession No. ML14169A359. Documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendment.
    Renewed Facility Operating License Nos. NPF-4 and NPF-7, DPR-32 and 
DPR-37: Amendments changed the licenses.
    Date of initial notice in Federal Register: September 3, 2013 (78 
FR 54292). The supplemental dated January 23, 2014, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed,

[[Page 52072]]

and did not change the staffs original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 12, 2014.
    No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas
    Date of amendment request: September 23, 2013.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.6.5, ``CORE OPERATING LIMITS REPORT (COLR),'' to 
replace the methodology of Westinghouse Electric Company LLC topical 
report WCAP-11596-P-A, ``Qualification of the Phoenix-P/ANC Nuclear 
Design System for Pressurized Water Reactor Cores,'' with WCAP-16045-P-
A, ``Qualification of the Two-Dimensional Transport Code PARAGON,'' and 
WCAP-16045-P-A, Addendum 1-A, ``Qualification of the NEXUS Nuclear Data 
Methodology,'' to determine core operating limits.
    Date of issuance: August 7, 2014.
    Effective date: As of its date of issuance and shall be implemented 
prior to core reload during Refueling Outage 20, currently expected to 
begin in January 2015.
    Amendment No.: 209. A publicly-available version is in ADAMS under 
Accession No. ML14156A246; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2013 (78 
FR 74186).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 7, 2014.
    No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas
    Date of amendment request: January 23, 2014. A redacted version was 
provided by letter dated March 31, 2014.
    Brief description of amendment: The amendment revised the Cyber 
Security Plan Implementation Milestone No. 8 completion date and the 
physical protection license condition.
    Date of issuance: August 14, 2014.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 210. A publicly-available version is in ADAMS under 
Accession No. ML14209A023; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Operating License.
    Date of initial notice in Federal Register: June 6, 2014 (79 FR 
32765).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 14, 2014.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 22nd day of August 2014.

    For the Nuclear Regulatory Commission.
A. Louise Lund,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2014-20671 Filed 8-29-14; 8:45 am]
BILLING CODE 7590-01-P