[Federal Register Volume 79, Number 150 (Tuesday, August 5, 2014)]
[Notices]
[Pages 45470-45484]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2014-18395]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0180]


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 July 10, 2014 to July 23, 2014.

DATES: Comments must be filed by September 4, 2014. A request for a 
hearing must be filed by October 6, 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-2014-0180. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: 3WFN-06-A44M, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-5411, email: Shirley.Rohrer@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0180 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-2014-0180.
     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 pdr.resource@nrc.gov. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2014-0180 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information in comment submissions that you do not want to be publicly 
disclosed in your comment submission. The NRC will post all comment 
submissions at http://www.regulations.gov as well as enter the comment 
submissions into ADAMS, and the NRC does not routinely edit comment 
submissions to remove identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of Title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this

[[Page 45471]]

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 hearing.docket@nrc.gov, 
or by telephone at 301-415-1677, to request (1) a digital 
identification (ID) certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System

[[Page 45472]]

requirements for accessing the E-Submittal server are detailed in the 
NRC's ``Guidance for Electronic Submission,'' which is available on the 
agency's public Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may attempt to use other software not 
listed on the Web site, but should note that the NRC's E-Filing system 
does not support unlisted software, and the NRC Meta System Help Desk 
will not be able to offer assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Dominion Energy Kewaunee (DEK), Docket No. 50-305, Kewaunee Power 
Station (KPS), Kewaunee County, Wisconsin
    Date of amendment request: January 16, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14029A076.
    Description of amendment request: The proposed amendment would 
modify the KPS renewed facility operating license by revising the 
emergency plan and the associated emergency action level (EAL) scheme 
consistent with the KPS permanent shutdown and defueled status. On 
February 25, 2013, DEK submitted a certification of permanent cessation 
of power operations pursuant to 10 CFR, Part 50, Section 
50.82(a)(1)(i), stating that DEK had decided to permanently cease power 
operation of KPS on May 7, 2013. With the docketing of subsequent 
certification for permanent removal of fuel from the reactor vessel 
pursuant to 10 CFR 50.82(a)(1)(ii) on May 14, 2013, the 10 CFR Part 50 
license for KPS no longer authorizes operation of the reactor or 
emplacement or retention of fuel into the reactor vessel, as specified 
in 10 CFR 50.82(a)(2). The proposed changes to the emergency plan and 
EAL scheme are being submitted to the U.S. Nuclear Regulatory 
Commission (NRC) for approval prior to implementation, as required 
under 10 CFR 50.54(q)(4) and 10 CFR Part 50, Appendix E, Section 
IV.B.2.
    DEK states that the proposed emergency plan changes do not meet all 
the standards of 10 CFR 50.47(b) and requirements of 10 CFR Part 50, 
Appendix E. By letter dated July 31,

[[Page 45473]]

2013 (ADAMS Accession No. ML13221A182), DEK submitted requests to the 
NRC for exemptions from portions of 10 CFR 50.47(b), 10 CFR 
50.47(c)(2), and 10 CFR Part 50, Appendix E, Section IV, that the 
proposed emergency plan does not meet. The proposed emergency plan 
revision is predicated on the approval of the requested exemptions.
    Basis for proposed no significant hazards consideration 
determination: Pursuant to 10 CFR 50.92, the NRC staff has provided its 
analysis of the issue of no significant hazards consideration which is 
presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    KPS has permanently ceased operation and is permanently 
defueled. Because the 10 CFR Part 50 license for KPS no longer 
authorizes operation of the reactor or emplacement or retention of 
fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2), 
the occurrence of postulated accidents associated with reactor 
operation is no longer credible. Analyses of the remaining credible 
accidents, as documented in the KPS Updated Safety Analysis Report 
(USAR), show that any releases beyond the site boundary would be 
below the Environmental Protection Agency (EPA) Protective Action 
Guides (PAGs) exposure levels, as detailed in the EPA's ``Protective 
Action Guide and Planning Guidance for Radiological Incidents,'' 
Draft for Interim Use and Public Comment dated March 2013.
    The proposed amendment would revise the emergency plan and EAL 
scheme to reflect the permanently defueled status of the plant. The 
proposed changes discontinue offsite emergency planning requirements 
and reduce the scope of onsite emergency planning requirements by 
removing positions that are no longer credited or needed for the 
remaining credible design basis accidents. The revised emergency 
plan and EAL scheme focus on responding to the emergencies that may 
arise from off-normal events and conditions which could indicate a 
degradation of the level of safety or indicate a security threat 
bounded by the type and significance of the remaining credible 
design basis accidents in a permanently shutdown and defueled 
condition.
    The proposed changes to the emergency plan do not impact the 
function of plant structures, systems, or components (SSCs). The 
proposed changes do not affect accident initiators or precursors, 
nor do they alter design assumptions. Therefore, the proposed 
changes to the emergency plan do not involve an increase in the 
probability of an accident previously evaluated.
    The proposed changes to the emergency plan remove positions from 
the emergency plan that are no longer credited or needed for the 
remaining credible design basis accidents. The proposed changes do 
not prevent the ability of the emergency response organization to 
perform its intended functions to mitigate the onsite consequences 
of an event for the remaining credible design basis accidents. The 
proposed changes do not increase the types or amounts of effluent 
releases beyond the site boundary from the remaining credible design 
basis accidents.
    Therefore, the proposed changes to the emergency plan do not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    The proposed changes to the EAL scheme limit the emergency 
classification levels to an Unusual Event and Alert. Because no 
remaining credible accidents can result in releases beyond the site 
boundary that exceed EPA PAG exposure levels, the need for emergency 
classifications of Site Area Emergency or General Emergency would 
not be required at a permanently shutdown and defueled facility. The 
changes to the EAL scheme do not involve any physical plant changes. 
The EALs and installed EAL equipment are not accident initiators and 
therefore the proposed changes to the EAL scheme do not involve an 
increase in the probability of an accident previously evaluated.
    The proposed EAL scheme changes do not affect the capability of 
SSCs to mitigate a design basis accident. Thus, the proposed changes 
do not involve a significant increase in the consequences of an 
accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment would revise the emergency plan and EAL 
scheme to reflect the permanently defueled status of the plant. The 
proposed changes do not involve installation of new equipment or 
modification of existing equipment, so that no new equipment failure 
modes are introduced. Also, the proposed changes do not result in a 
change to the way that the equipment or facility is operated so that 
no new accident initiators are created.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment would revise the emergency plan and EAL 
scheme to reflect the permanently defueled status of the plant. The 
proposed changes to the emergency plan and EAL scheme do not involve 
a change in the plant's design, configuration, or operation. The 
proposed changes do not affect the way the plant structures, 
systems, and components perform their safety functions or their 
design margins as they apply to the remaining credible accidents. 
The proposed changes do not involve a change to the technical 
specifications. Because there is no change to the physical design or 
operation of the plant, no change to the accident analyses, and no 
change to the safety analysis acceptance criteria as a result of 
this amendment, there is no change to any of these margins.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    Based on the NRC staff's analysis, 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: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 
120 Tredegar Street, Richmond, VA 23219.
    NRC Branch Chief: Douglas A. Broaddus.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station (ONS), Units 1, 2, and 3, Oconee County, South 
Carolina
    Date of amendment request: May 20, 2014. A publicly available 
version is in ADAMS under Accession No. ML14141A415.
    Description of amendment request: The proposed amendment requests 
removal of Technical Specification requirements for ONS units that did 
not have the Reactor Protection System (RPS)/Engineered Safeguards 
Protective System (ESPS) digital upgrades or Low Pressure Service Water 
(LPSW) Reactor Building (RB) Waterhammer Prevention System (WPS) 
modifications. The Licensee stated that these Technical Specification 
requirements no longer pertain to ONS since the RPS/ESPS digital 
upgrade and the LPSW RB WPS modification have been implemented for all 
three ONS units. The proposed amendment also deletes a Note statement 
for the Emergency Condenser Circulating Water (ECCW) System Technical 
Specification that states the Technical Specification is not applicable 
until after completion of the Service Water upgrade modifications on 
each respective ONS unit. The licensee stated that the Service Water 
upgrade modifications have been implemented for each ONS unit.
    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 changes to Technical Specifications 3.3.1, 3.3.3, 
3.3.5, 3.3.7, 3.3.27,

[[Page 45474]]

3.6.5, 3.7.7, and 3.7.8 do not modify the Reactor Protective System 
(RPS), Engineered Safeguards Protective System (ESPS), Low Pressure 
Service Water (LPSW) System, the LPSW Reactor Building (RB) 
Waterhammer Protection System (WPS) or the Emergency Condenser 
Circulating Water (ECCW) System, nor make any physical changes to 
the facility design, material, or construction standards. The 
proposed changes remove obsolete information from the Technical 
Specifications that no longer apply to ONS; delete Surveillance 
Requirements (SRs) for the RPS RB High Pressure trip function and 
the ESPS RB Pressure--High High actuation parameter that are not 
applicable; and correct a wording error in a Condition statement for 
TS 3.7.7 which results in a more stringent Condition. Since the 
removed information no longer applies to ONS, and the deleted SRs 
are for equipment features that do not exist for the RPS RB High 
Pressure trip function and the ESPS RB Pressure--High High actuation 
parameter, removal of the information and deletion of the SRs do not 
result in operation that will increase the probability of initiating 
an analyzed event. Likewise, the more restrictive requirement in the 
corrected Condition statement continues to ensure process variables, 
structures, systems, and components are maintained consistent with 
the safety analyses and licensing basis. The proposed Technical 
Specification changes do not alter assumptions relative to 
mitigation of an accident or transient event. The removal of the 
obsolete Technical Specification information, deletion of SRs for 
features that do not exist, and correction of the Technical 
Specification Condition statement have no effect on the process 
variables, structures, systems, and components that must be 
maintained consistent with the safety analyses and licensing basis. 
Therefore, the proposed Technical Specification changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the Proposed Change Create the Possibility of a New or 
Different Kind of Accident From Any Accident Previously Evaluated?
    Response: No.
    The proposed changes to Technical Specifications 3.3.1, 3.3.3, 
3.3.5, 3.3.7, 3.3.27, 3.6.5, 3.7.7, and 3.7.8 only remove obsolete 
information from the Technical Specifications pertaining to the RPS/
ESPS digital upgrade, the LPSW RB WPS modification installation, and 
the ECCW System Service Water upgrade modification completion. The 
proposed changes also delete SRs that verify features that do not 
exist for the RPS RB High Pressure trip function and the ESPS RB 
Pressure--High High actuation parameter. Lastly, the proposed 
changes correct a wording error in a Condition statement for TS 
3.7.7 which results in a more stringent Condition. The changes do 
not alter the plant configuration (no new or different type of 
equipment will be installed) or make changes in the methods 
governing normal plant operation. The RPS, ESPS, LPSW System, LPSW 
RB WPS, and ECCW System are not associated with any design accident 
initiation; they only mitigate accidents. However, these proposed 
Technical Specification changes are consistent with the assumptions 
in the safety analyses and licensing basis. Therefore, the proposed 
Technical Specification changes do not create the possibility of a 
new or different kind of accident from any kind of accident 
previously evaluated.
    3. Does the Proposed Change Involve a Significant Reduction in a 
Margin of Safety?
    Response: No.
    The proposed changes to Technical Specifications 3.3.1, 3.3.3, 
3.3.5, 3.3.7, 3.3.27, 3.6.5, 3.7.7, and 3.7.8 remove information 
from the Technical Specifications pertaining to the RPS/ESPS digital 
upgrade, the LPSW RB WPS modification installation, and the ECCW 
System Service Water upgrade modification completion. The proposed 
changes also delete SRs that verify features that do not exist for 
the RPS RB High Pressure trip function and the ESPS RB Pressure--
High High actuation parameter. Lastly, the proposed changes correct 
a wording error in a Condition statement for TS 3.7.7 which results 
in a more stringent Condition. The removed Technical Specification 
information no longer applies to ONS operation and is considered 
obsolete; the deleted SRs cannot be performed since the affected 
plant equipment will not support SR testing by design; and the 
corrected TS 3.7.7 Condition statement results in a more 
conservative Technical Specification. Removal of the Technical 
Specification obsolete information has no impact on the margin of 
safety since the equipment that the Technical Specification 
information applied to no longer exists at ONS. Deletion of SRs on 
the subject RPS/ESPS equipment has no impact on the margin of safety 
since the RPS/ESPS equipment, by design, will not support SR 
testing. Correction of the TS 3.7.7 Condition statement has no 
impact on the margin of safety since the correction results in a 
more conservative Technical Specification. The changes maintain 
requirements within the safety analyses and licensing basis. As 
such, no question of safety is involved. Therefore, the proposed 
Technical Specification 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: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana
    Date of amendment request: November 11, 2013. A publicly-available 
version is in ADAMS under Accession No. ML13316C052.
    Description of amendment request: Entergy Operations, Inc. (the 
licensee), has proposed to change the Waterford Steam Electric Station, 
Unit 3 Updated Final Safety Analysis Report (UFSAR). This change will 
clarify in the UFSAR how the pressurizer heaters function is met for 
natural circulation at the onset of a loss-of-offsite power concurrent 
with the specific single point vulnerability.
    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 would describe the specific common circuit 
breaker associated with the control power closing circuitry to the 
Switchgears 32A and 32B Supply Circuit Breakers in UFSAR 1.9.26 and 
5.4.10 as contained in Attachment 2 [of the licensee's letter dated 
November 11, 2013] and that local manual operation outside of the 
Control Room would be necessary to reenergize Pressurizer Heaters 
during a loss of offsite power concurrent with the specific common 
circuit breaker being open. Plant Operators are trained and have 
procedural guidance including manual operator action to address 
Natural Circulation Cooldown with a Loss of Offsite Power. The 
Pressurizer Heaters are not themselves a credible initiator of any 
accident, and the requested amendment makes no change to the 
Pressurizer Heaters themselves, so the probability of an accident 
will not be increased. The proposed change would not change the 
source term nor adversely impact any mitigating systems, so the 
consequences of an accident will not be increased.
    Therefore, the probability or consequences of any accident 
previously evaluated will not be increased by the proposed change.
    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 would describe the specific common circuit 
breaker associated with the control power closing circuitry to the 
Switchgears 32A and 32B Supply Circuit Breakers in UFSAR 1.9.26 and 
5.4.10 as contained in Attachment 2 [of the licensee's letter dated 
November 11, 2013] and that local manual operation outside of the 
Control Room would be necessary to reenergize Pressurizer Heaters 
during a loss of offsite power concurrent with the specific common 
circuit breaker being open.
    The proposed changes do not involve a change in the design, 
configuration, or method of operation of the plant that could create 
the possibility of a new or different

[[Page 45475]]

accident. Equipment will be operated in a manner for which it is 
currently designed. This license amendment request does not impact 
any plant systems that are accident initiators or adversely impact 
any accident mitigating systems. The Pressurizer Heaters are not 
themselves a credible initiator of any accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change would describe the specific common circuit 
breaker associated with the control power closing circuitry to the 
Switchgears 32A and 32B Supply Circuit Breakers in UFSAR 1.9.26 and 
5.4.10 as contained in Attachment 2 [of the licensee's letter dated 
November 11, 2013] and that local manual operation outside of the 
Control Room would be necessary to reenergize Pressurizer Heaters 
during a loss of offsite power concurrent with the specific common 
circuit breaker being open. Plant Operators are trained and have 
procedural guidance including manual operator action to address 
Natural Circulation Cooldown with a Loss of Offsite Power.
    This amendment does not change the manner in which safety limits 
or limiting safety settings are determined. Because the Pressurizer 
Heaters will continue to be monitored and controlled as per 
Technical Specification 3.4.3.1 and Technical Requirements Manual 
3.4.3.1, this proposed change to the UFSAR will not present an 
adverse impact to plant operation or result in a significant 
reduction in a margin of safety.
    Therefore, the proposed change will 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: Joseph A. Aluise, Associate General 
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Douglas A. Broaddus.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana
    Date of amendment request: December 9, 2013. A publicly-available 
version is in ADAMS under Accession No. ML13345A686.
    Description of amendment request: Entergy Operations, Inc. (the 
licensee), has proposed to change the Waterford Steam Electric Station, 
Unit 3 Technical Specifications (TS). Specifically, the amendment would 
revise:
     TS 3.3.1, Reactor Protective Instrumentation;
     TS 3.1.3.4, Shutdown CEA [Control Element Assembly];
     TS 3.3.2, Engineered Safety Features Actuation System 
Instrumentation;
     TS 3.3.3.1, Radiation Monitoring Instrumentation;
     TS 3.3.3.6, Accident Monitoring Instrumentation;
     TS 3.3.3.11, Explosive Gas Monitoring Instrumentation;
     TS 4.8.2.1, D.C. [Direct Current] Sources;
     TS 6.1, Responsibility;
     TS 6.2.1, Offsite and Onsite Organizations;
     TS 6.2.2, Unit Staff; and
     TS 6.12, High Radiation Area.
    These changes would improve clarity, correct administrative and 
typographical errors, or establish consistency with NUREG-1432, 
Standard Technical Specifications Combustion Engineering Plants, 
Revision 4.0 (NUREG-1432).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the Technical Specifications to 
improve clarity, correct administrative and typographical errors, 
and establish consistency with NUREG-1432. This includes two 
technical changes.
    A provision to an existing surveillance test has been added that 
limits the total battery inter-cell resistance to maintain battery 
terminal voltage above the required operating voltage. A change to 
limit the total battery inter-cell resistance has no effect on the 
probability of an accident previously evaluated. The proposed change 
to limit the total battery inter-cell resistance does not involve a 
significant increase in the consequences of an accident previously 
evaluated. This is because the addition of this limit will ensure 
that the battery is demonstrated as capable to meet its safety 
function.
    The other technical change extends the Completion Time from 1 
hour to 4 hours for verifying that the departure from nucleate 
boiling ratio (DNBR) limit is met and disabling the Reactor Power 
Cutback when one or both CEACs [Control Element Assembly 
Calculators] are inoperable. A change to the Completion Time for 
Actions in response to inoperable equipment has no effect on the 
probability of an accident previously evaluated. The proposed change 
to the Completion Time for Actions in response to inoperable 
equipment does not involve a significant increase in the 
consequences of an accident previously evaluated. This is because 
the safety function of a CEAC is to identify and compensate for a 
misaligned CEA [control element assembly], and there is a low 
probability of occurrence during the four hour Completion Time that 
one or more misaligned CEAs could significantly adversely affect: 
Core power distribution, shutdown margin, ejected CEA worth, or 
initial reactivity insertion rate during a reactor trip.
    Consequently, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes revise the Technical Specifications to 
improve clarity, correct administrative and typographical errors, 
and establish consistency with NUREG-1432. This includes two 
technical changes.
    A provision to an existing surveillance test has been added that 
limits the total battery inter-cell resistance to maintain battery 
terminal voltage above the required operating voltage. A change to 
limit the total battery inter-cell resistance does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. This is because the addition of this limit 
will ensure that the battery is demonstrated as capable to meet its 
existing safety function and does not change the safety function in 
any manner.
    The other technical change extends the Completion Time from 1 
hour to 4 hours for verifying that the departure from nucleate 
boiling ratio (DNBR) limit is met and disabling the Reactor Power 
Cutback when one or both CEACs are inoperable. A change to the 
Completion Time for Actions in response to inoperable equipment does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Consequently, the proposed changes do not create the possibility 
of a new or different kind of accident.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes revise the Technical Specifications to 
improve clarity, correct administrative and typographical errors, 
and establish consistency with NUREG-1432. This includes two 
technical changes.
    A provision to an existing surveillance test has been added that 
limits the total battery inter-cell resistance to maintain battery 
terminal voltage above the required operating voltage. A change to 
limit the total battery inter-cell resistance does not involve a 
significant reduction in a margin of safety. This is because the 
addition of this limit will ensure that the battery is demonstrated 
as having margin to meet its safety function.
    The other technical change extends the Completion Time from 1 
hour to 4 hours for verifying that the departure from nucleate 
boiling ratio (DNBR) limit is met and disabling the Reactor Power 
Cutback when

[[Page 45476]]

one or both CEACs are inoperable. A change to the Completion Time 
for Actions in response to inoperable equipment does not affect 
protection criterion for plant equipment and does not reduce the 
margin of safety. This change provides Operators time to assess and 
perform the required activities in a controlled manner consistent 
with the risk associated with an inoperable CEAC function. Actions 
associated with this Condition involve disabling the Control Element 
Drive Mechanism Control System (CEDMCS), and signaling all OPERABLE 
CPC [core protection calculator] channels that both CEACs are 
failed. This applies a large penalty factor associated with two CEAC 
failures within CPC calculations. The penalty factor for two failed 
CEACs is sufficiently large that power must be maintained 
significantly <100% Reactor Thermal Power. The Completion Time of 4 
hours is adequate to accomplish these actions while minimizing 
risks. Meeting the DNBR margin requirements ensures that power level 
and ASI [axial shape index] are within a conservative region of 
operation based on actual core conditions. In addition to the above 
actions, the Reactor Power Cutback System is disabled. This ensures 
that CEA position will not be affected by Reactor Power Cutback 
operation.
    Consequently, there is no significant reduction in a margin of 
safety due to the proposed 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: Joseph A. Aluise, Associate General 
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, LA 70113.
    NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit No. 1, (TMI-1) Dauphin County, Pennsylvania
    Date of amendment request: May 7, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14127A424.
    Description of amendment request: The amendment would change the 
TMI-1 technical specifications. Specifically, the proposed amendment 
would replace an existing Surveillance Requirement to operate 
ventilation systems with charcoal filters for a 10-hour period every 31 
days with a requirement to operate the systems for greater than or 
equal to 15 continuous minutes every 31 days in accordance with 
Technical Specification Task Force (TSTF) Traveler TSTF-522, Revision 
0, ``Revise Ventilation System Surveillance Requirements to Operate for 
10 hours per Month'' (ADAMS Accession No. ML100890316).
    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, along with NRC edits in square 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change replaces an existing [Surveillance 
Requirement] SR to operate the Emergency Control Room Air Treatment 
System and the Fuel Handling Building [Engineered Safety Feature] 
ESF Air Treatment System for a 10-hour period at a frequency 
controlled in accordance with the [Surveillance Frequency Control 
Program] SFCP with a requirement to operate the systems for greater 
than or equal to 15 continuous minutes at a frequency controlled in 
accordance with the SFCP.
    These systems are not accident initiators and therefore, these 
changes do not involve a significant increase in the probability of 
an accident. The proposed system and filter testing changes are 
consistent with current regulatory guidance for these systems and 
will continue to assure that these systems perform their design 
function, which may include mitigating accidents. Thus, the change 
does not involve a significant increase in the consequences of an 
accident.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change replaces an existing SR to operate the 
Emergency Control Room Air Treatment System and the Fuel Handling 
Building ESF Air Treatment System for a 10-hour period at a 
frequency controlled in accordance with the SFCP with a requirement 
to operate the systems for greater than or equal to 15 continuous 
minutes at a frequency controlled in accordance with the SFCP.
    The change proposed for these ventilation systems does not 
change any system operations or maintenance activities. Testing 
requirements will be revised and will continue to demonstrate that 
the Limiting Conditions for Operation are met and the system 
components are capable of performing their intended safety 
functions. The change does not create new failure modes or 
mechanisms and no new accident precursors are generated.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change replaces an existing SR to operate the 
Emergency Control Room Air Treatment System and the Fuel Handling 
Building ESF Air Treatment System for a 10-hour period at a 
frequency controlled in accordance with the SFCP with a requirement 
to operate the systems for greater than or equal to 15 continuous 
minutes at a frequency controlled in accordance with the SFCP. The 
proposed change is consistent with regulatory guidance.
    Therefore, it is concluded that this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Esquire, Exelon 
Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Acting Branch Chief: Robert G. Schaaf.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, 
Beaver Valley Power Station, (BVPS) Unit No. 1, Beaver County, 
Pennsylvania
    Date of amendment request: July 30, 2013. A publicly-available 
version is in ADAMS under Accession No. ML13212A027.
    Description of amendment request: The amendment would change the 
BVPS Facility Operating License. Specifically, the amendment requests 
authorization to implement 10 CFR 50.61a, ``Alternate fracture 
toughness requirements for protection against pressurized thermal shock 
events,'' in lieu of 10 CFR 50.61, ``Fracture toughness requirements 
for protection against pressurized thermal shock events.'' The 10 CFR 
50.61 screening criteria define a limiting level of reactor pressure 
vessel embrittlement beyond which plant operation cannot continue 
without further evaluation. As described in NUREG-1806, ``Technical 
Basis for Revision of the Pressurized Thermal Shock (PTS) Screening 
Limit in the PTS Rule (10 CFR 50.61),'' the screening criteria in the 
PTS rule is overly conservative and the risk of through wall cracking 
due to a PTS event is much lower than previously estimated. A publicly-
available version of NUREG-1806 is in ADAMS under Accession No. 
ML072830074.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 45477]]

issue of no significant hazards consideration, which is presented 
below, with NRC edits in square brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This amendment request would allow implementation of the 
alternate PTS [pressurized thermal shock] rule in lieu of 10 CFR 
50.61 and would not involve a significant increase in the 
probability or consequences of an accident. Application of the 
alternate PTS rule in lieu of 10 CFR 50.61 would not result in 
physical alteration of a plant structure, system or component, or 
installation of new or different types of equipment. Further, 
application of the alternate PTS rule would not significantly affect 
the probability of accidents previously evaluated in the Updated 
Final Safety Analysis Report (UFSAR) or cause a change to any of the 
dose analyses associated with the UFSAR accidents because accident 
mitigation functions would remain unchanged. Use of the alternate 
PTS rule would change how fracture toughness of the reactor vessel 
is determined and does not affect reactor vessel neutron radiation 
fluence. As such, implementation of the alternate PTS rule in lieu 
of 10 CFR 50.61 would not increase the likelihood of a malfunction.
    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 amendment request would allow implementation of the 
alternate PTS rule in lieu of 10 CFR 50.61. No new accident 
scenarios, failure mechanisms, or limiting single failures are 
introduced as a result of the proposed change. No physical plant 
alterations are made as a result of the proposed change. The 
proposed change does not challenge the performance or integrity of 
any safety-related system.
    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 amendment request would authorize implementation of the 
alternate PTS rule in lieu of 10 CFR 50.61. The alternate PTS rule 
would maintain the same functional requirements for the facility as 
10 CFR 50.61. The alternate PTS rule establishes screening criteria 
that limit levels of embrittlement beyond which operation cannot 
continue without further plant-specific evaluation or modifications. 
Sufficient safety margins are maintained to ensure that any 
potential increases in core damage frequency and large early release 
frequency resulting from implementation of the alternate PTS rule 
are negligible. As such, there would be no significant reduction in 
the margin of safety as a result of use of the alternate PTS rule. 
The margin of safety associated with the acceptance criteria of 
accidents previously evaluated in the UFSAR is unchanged. The 
proposed change would have no affect on the availability, 
operability, or performance of the safety-related systems and 
components.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Acting Branch Chief: Robert G. Schaaf.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, (BVPS-1 and 
BVPS-2) Beaver County, Pennsylvania
    Date of amendment request: April 16, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14111A291.
    Description of amendment request: The amendment would change BVPS-1 
and BVPS-2 technical specifications (TSs). Specifically, the proposed 
license amendment would revise TS 5.5.12, ``Containment Leakage Rate 
Testing Program,'' Item a, by deleting reference to the BVPS-1 
exemption letter dated December 5, 1984 (ADAMS Accession No. 
ML003766713), and requiring compliance with 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,'' 
(ADAMS Accession No. ML12221A202) instead of Regulatory Guide 1.163, 
``Performance-Based Containment Leak Test Program,'' (ADAMS Accession 
No. ML003740058) including listed exceptions. In summary, the amendment 
would allow extension of the Type A Reactor Containment Integrated Leak 
test, required by 10 CFR Part 50, Appendix J, interval to one test in 
15 years and an extension of the Type C test interval to 75 months, 
based on acceptable performance history of the containment test as 
defined in NEI 94-01, Revision 3-A.
    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, along with NRC edits in square 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment adopts the NRC-accepted guidelines of NEI 
94-01, Revision 3-A, ``Industry Guideline for Implementing 
Performance-Based Option of 10 CFR Part 50, Appendix J,'' for 
development of the Beaver Valley Power Station, Unit No. 1 (BVPS-1) 
and Unit No.2 (BVPS-2) performance-based containment testing 
program. NEI 94-01 allows, based on risk and performance, an 
extension of Type A and Type C containment leak test intervals. 
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 findings of the Beaver Valley Power Station risk assessment 
confirm the general findings of previous studies that the risk 
impact with extending the containment leak rate is small. Per the 
guidance provided in Regulatory Guide 1.174, [An Approach for using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis (ADAMS Accession No. 
ML100910006)] [* * * ] an extension of the leak test interval in 
accordance with NEI 94-01 [Revision 3-A] results in an estimated 
change within the very small change region.
    Since the change is implementing a performance-based containment 
testing program, the proposed amendment does not involve either a 
physical change to the plant or a change in the manner in which the 
plant is operated or controlled. The requirement for leakage rate 
acceptance will not be changed by this amendment. Therefore, the 
containment will continue to perform its design function as a 
barrier to fission product releases.
    Therefore, the proposed amendment does not significantly 
increase 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 to implement a performance-based containment 
testing program, associated with integrated leakage rate test 
frequency, does not change the design or operation of structures, 
systems, or components of the plant. In addition, the proposed 
changes would not impact any other plant system or component.
    The proposed changes would continue to ensure containment 
integrity and would ensure operation within the bounds of existing 
accident analyses. There are no accident initiators created or 
affected by these changes. Therefore, the proposed changes will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. [* * * ]

[[Page 45478]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to implement a performance-based containment 
testing program, associated with integrated leakage rate test 
frequency, does not affect plant operations, design functions, or 
any analysis that verifies the capability of a structure, system, or 
component of the plant to perform a design function. In addition, 
this change does not affect safety limits, limiting safety system 
setpoints, or limiting conditions for operation.
    The specific requirements and conditions of the Technical 
Specification Containment Leak Rate Testing Program exist to ensure 
that the degree of containment structural integrity and leak-
tightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by 
Technical Specifications is maintained. This ensures that the margin 
of safety in the plant safety analysis is maintained. The design, 
operation, testing methods and acceptance criteria for Type A, B, 
and C containment leakage tests specified in applicable codes and 
standards would continue to be met, with the acceptance of this 
proposed change, since these are not affected by implementation of a 
performance-based containment testing program.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, 
Akron, OH 44308.
    NRC Acting Branch Chief: Robert G. Schaaf.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska
    Date of amendment request: June 2, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14157A006.
    Description of amendment request: The proposed amendment would 
revise the Cooper Nuclear Station Technical Specifications (TS) to 
update Figure 4.1-1, ``Site and Exclusion Area Boundaries and Low 
Population Zone,'' to reflect the current site layout.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change updates a figure with the current site 
layout. An administrative change such as this is not an initiator of 
any accident previously evaluated. As a result, the probability of 
an accident previously evaluated is not affected. The consequences 
of an accident with the incorporation of this administrative change 
are not different than the consequences of the same accident without 
this change. As a result, the consequences of an accident previously 
evaluated are not affected by this change.
    Based on the above, it is concluded that the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not modify the plant design, nor does 
the proposed change alter the operation of the plant or equipment 
involved in either routine plant operation or in the mitigation of 
design basis accidents. The proposed change is administrative only.
    Based on the above, it is concluded that the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change consists of an administrative change to 
update a figure of the site layout. The change 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 change will not result in plant operation in a 
configuration outside of the design basis. 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: John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Michael T. Markley.
Northern States Power Company--Minnesota (NSPM), Docket No. 50-263, 
Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota
    Date of amendment request: November 14, 2013. A publicly-available 
version is in the Agencywide Documents Access and Management System 
under Accession No. ML13322A446.
    Description of amendment request: NSPM proposes to revise the MNGP 
technical specification (TS) 5.5.11, ``Primary Containment Leakage Rate 
Testing Program,'' airlock testing conditions. Specifically, NSPM 
proposes to remove the reduced pressure testing option for drywell 
airlock door leakage testing in accordance with the requirements of 
Part 50 to Title 10 of the Code of Federal Regulations (10 CFR 50), 
Appendix J, Option B, since this capability is not required and does 
not reflect the current testing practice at MNGP.
    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 provided below.

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No
    The proposed change removes the TS allowance to test the leakage 
rate of the drywell personnel airlock doors at a reduced pressure. 
However, overall airlock leakage rate testing will continue to be 
performed in accordance with Option B of 10 CFR 50, Appendix J. 
Removal of this capability does not affect, nor is it a precursor 
for, an accident or transient analyzed in the MNGP Updated Safety 
Analysis Report. The proposed change does not change the total 
allowable primary containment leakage rate, nor does it involve a 
change to the physical design and operation of the plant.
    Therefore, operation of the facility in accordance with 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 removes the TS allowance to test the leakage 
rate of the drywell personnel airlock doors at a reduced pressure. 
However, overall airlock leakage rate testing will continue to be 
performed in accordance with Option B to 10 CFR 50, Appendix J. The 
change being proposed will not change the physical plant or modes of 
operation defined in the facility license. The proposed change does 
not increase the total allowable primary containment leakage rate. 
The change does not involve the addition or modification of 
equipment, nor does it alter the design or operation of plant 
systems.
    Therefore, operation of the facility in accordance with the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

[[Page 45479]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No
    The proposed change removes the TS allowance to test the leakage 
rate of the drywell personnel airlock doors at a reduced pressure. 
However, overall airlock leakage rate testing will continue to be 
performed in accordance with Option B to 10 CFR 50, Appendix J. The 
proposed change does not affect plant safety analyses or change the 
physical design or operation of the plant. The proposed change does 
not increase the total allowable primary containment leakage rate.
    Therefore, operation of the facility in accordance with the 
proposed change does not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: David L. Pelton.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota
    Date of amendment request: June 9, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14160A593.
    Description of amendment request: The proposed amendments would 
revise the Prairie Island Nuclear Generating Plant, Units 1 and 2, 
Surveillance Requirements 3.8.1.2, 3.8.1.6, and 3.8.1.9 associated with 
steady state voltage and frequency limits in Technical 
Specification3.8.1, ``AC [Alternating Current] Sources--Operating.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No
    This license amendment request proposes to revise specific 
emergency diesel generator steady states voltage and frequency 
limits in the Technical Specification Surveillance Requirements 
which are more restrictive than the current limits.
    The emergency diesel generators and the equipment on the 
safeguards buses supplied by the emergency diesel generators are not 
accident initiators, and therefore the proposed voltage and 
frequency limits changes do not involve an increase in the 
probability of an accident.
    The proposed emergency diesel generator surveillance test 
voltage and frequency limits assure the emergency diesel generators 
are capable of providing electrical power at voltages and 
frequencies that are adequate to operate the required equipment on 
the safeguards buses and thus maintain the current licensing basis 
for accident mitigation. Thus the proposed voltage and frequency 
limit changes do not involve a significant increase in the 
consequences of an accident.
    Therefore, the proposed Technical Specification changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No
    This license amendment request proposes to revise specific 
emergency diesel generator steady state voltage and frequency limits 
in the Technical Specification Surveillance Requirements which are 
more restrictive than the current limits.
    The proposed Technical Specification changes which revise the 
emergency diesel generator voltage and frequency limits do not 
change any system operations or maintenance activities. The changes 
do not involve physical alteration of the plant; that is, no new or 
different type of equipment will be installed. The changes do not 
alter assumptions made in the safety analyses but ensure that the 
diesel generators are capable of operating equipment as assumed in 
the accident analyses. These changes do not create new failure modes 
or mechanisms which are not identifiable during testing and no new 
accident precursors are generated.
    Therefore, the proposed Technical Specification changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No
    This license amendment request proposes to revise specific 
emergency diesel generator steady state voltage and frequency limits 
in the Technical Specification Surveillance Requirements which are 
more restrictive than the current limits.
    Since this license amendment proposes Technical Specification 
changes which further restrict the acceptable voltage and frequency 
limits, both upper and lower, margins of safety are increased, and 
no margin of safety is reduced as part of this change.
    Therefore, the proposed Technical Specification 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 requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401
    NRC Branch Chief: David L. Pelton.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028, 
Virgil C. Summer Nuclear Station (VCS) Units 2 and 3, Fairfield County, 
South Carolina
    Date of amendment request: March 19, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14079A599.
    Description of amendment request: The requested amendment 
reclassifies portions of the five Tier 2* Human Factors (HF) 
Verification & Validation (V&V) planning documents listed in the 
Updated Final Safety Analysis Report (UFSAR) Table 1.6-1 and Chapter 
18, Subsection 18.11.2. These five documents outline the overall plan 
for the HF V&V, including the Human Factors Engineering (HFE) design 
verification, task support verification, integrated system validation, 
discrepancy resolution process, and verification at plant startup. The 
licensee stated that the requested amendment identifies the portions of 
the five HF V&V planning documents that would more appropriately be 
classified as Tier 2, due to those portions having no impact on safety, 
and proposes the necessary departures to reclassify this information. 
This differentiation between Tier 2 and Tier 2* information in the HF 
V&V planning documents will allow for revisions of these documents 
using the Tier 2 change process provided in 10 CFR Part 52 Appendix D, 
Section VIII.B.5. Because this proposed change requires a departure 
from Tier 2* information in the Westinghouse Advanced Passive 1000 
design control document (DCD), the licensee also requested an exemption 
from the requirements of the Generic DCD Tier 2* in accordance with 10 
CFR Part 52 Appendix D Section VIII B.6.c.(15).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No

[[Page 45480]]

    The proposed changes reclassify portions of the five Tier 2* 
Human Factors (HF) Verification & Validation (V&V) planning 
documents listed in the Updated Final Safety Analysis Report 
(UFSAR). These changes do not modify the design, construction, or 
operation of any plant structures, systems, or components (SSC), nor 
do they change any procedures or method of control for any SSCs. 
Because the proposed changes do not change the design, construction, 
or operation of any SSCs, they do not adversely affect any design 
function as described in the UFSAR. Therefore, the proposed 
amendment does not affect the probability of an accident previously 
evaluated. Similarly, because the proposed changes do not alter the 
design or operation of the nuclear plant or any plant SSCs, the 
proposed changes do not represent a change to the radiological 
effects of an accident, and therefore, they do not involve an 
increase in the consequences of an accident previously evaluated.
    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 changes are not a modification, addition to, or 
removal of any plant SSCs. Furthermore, the proposed changes are not 
a change to procedures or method of control of the nuclear plant or 
any plant SSCs. The only impact of this activity is the 
reclassification of portions of the five HF V&V planning documents 
as Tier 2 information. Because the proposed amendment does not 
change the design, construction, or operation of the nuclear plant 
or any plant operations, it does not affect the possibility of an 
accident.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No
    The proposed changes reclassify portions of the five Tier 2* HF 
V&V planning documents listed in the UFSAR from Tier 2* to Tier 2. 
The proposed amendment only affects the classification of planning 
documents and does not change the design, construction, or operation 
of the nuclear plant or any plant operations; therefore, the changes 
do not affect any margin of safety.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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 Docket Nos.: 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: July 3, 2014. A publicly available 
version is available in the Agencywide Documents Access and Management 
System under Accession No. ML14187A533.
    Description of amendment request: The purpose of the proposed 
license amendment request is to address proposed changes related to the 
design details of the containment internal structural wall modules 
(CA01, CA02, and CA05). The proposed changes to Tier 2 information in 
the Updated Final Safety Analysis Report (UFSAR), and the involved 
plant-specific Tier 1 and corresponding combined license Appendix C 
information would allow the use of thicker than normal faceplates to 
accommodate local demand or connection loads in certain areas without 
the use of overlay plates or additional backup structures. Additional 
proposed changes to Tier 2 information and involved Tier 2* information 
would allow:
    (1) A means of connecting the structural wall modules to the base 
concrete via use of structural shapes, reinforcement bars, and shear 
studs extending horizontally from the structural module faceplates and 
embedded during concrete placement as an alternative to the use of 
embedment plates and vertically oriented reinforcement bars,
    (2) A variance in structural module wall thicknesses from the 
thicknesses identified in UFSAR Figure 3.8.3-8, ``Structural Modules--
Typical Design Details,'' for some walls that separate equipment spaces 
from personnel access areas, and
    (3) The use of steel plates, structural shapes, reinforcement bars, 
or tie bars between the module faceplates, as needed to support 
localized loads and ensure compliance with applicable codes.
    Because this proposed change requires a departure from Tier 1 
information in the Westinghouse Advanced Passive 1000 design control 
document (DCD), the licensee also requested an exemption from the 
requirements of the Generic DCD Tier 1 in accordance with 52.63(b)(1).
    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 requested amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No
    The design function of the internal containment structures is to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in those structures. 
These structures are structurally designed to meet seismic Category 
I requirements as defined in Regulatory Guide 1.29.
    The changes to the design details for the structural modules do 
not have an adverse impact on the response of the nuclear island 
structures to safe shutdown earthquake ground motions or loads due 
to anticipated transients or postulated accident conditions, nor do 
they change the seismic Category I classification.
    Evaluations have been performed which determined that the 
proposed changes do not have a significant impact on the calculated 
loads for the affected structural modules, or critical locations, 
and no significant impact on the global seismic model. The changes 
to the design details for the structural modules do not impact the 
support, design, or operation of mechanical and fluid systems. There 
is no change to plant systems or the response of systems to 
postulated accident conditions. There is no change to the predicted 
radioactive releases due to postulated accident conditions. The 
plant response to previously evaluated accidents or external events 
is not adversely affected, nor does the change described create any 
new accident precursors.
    Therefore, the requested amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the requested amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes are to revise design details for the 
internal containment structural modules. The changes do not change 
the design requirements of the nuclear island structures, nor do 
they change the seismic Category I classification. The changes to 
the design details for the internal containment structural modules 
do not change the design function, support, design, or operation of 
mechanical and fluid systems. The changes to the design details for 
the internal containment structural modules do not result in a new 
failure mechanism for the nuclear island structures or introduce any 
new accident precursors. As a result, the design function of the 
nuclear island structures is not adversely affected by the proposed 
change.
    Therefore, the requested amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?

[[Page 45481]]

    Response: No.
    The requested amendment proposes changes to the structural 
details associated with the in-containment structural modules. The 
purpose of these changes is to ensure that the requirements 
contained in the applicable construction codes are met. As discussed 
in UFSAR, Section 3.8.3.5, ``Design Procedures and Acceptance 
Criteria,'' the in-containment structural modules are designed in 
accordance with ACI 349 and AISC N690. Thus, the identification of 
additional structural module connection details, the increase in 
structural module faceplate and wall thicknesses, and the addition 
of additional reinforcement in specific areas are proposed to ensure 
that the codes of record, and the associated margins contained 
therein, continue to be met as specified in the design basis. 
Structural and seismic analysis of the modified sections in 
accordance with the methodologies identified in the UFSAR has 
confirmed that the applicable requirements of ACI 349 and AISC N690 
continue to be met for affected in- containment structural modules.
    As a result, the proposed changes do not adversely affect any 
safety-related equipment or other design functions, design code 
compliance, design analysis, safety analysis input or result, or 
design/safety margin. No safety analysis or design basis acceptance 
limit/criterion is challenged or exceeded by the proposed changes. 
Therefore, the requested amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Blach & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence Burkhart.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas
    Date of amendment request: January 6, 2014, as supplemented by 
letter dated June 9, 2014. Publicly-available versions are in ADAMS 
under Accession Nos. ML14035A075 and ML14184B363.
    Description of amendment request: The proposed license amendment 
would revise Technical Specification (TS) 3.3.1, ``Reactor Trip System 
Instrumentation,'' with respect to the required actions and allowed 
outage times for inoperable reactor trip breakers. The proposed changes 
would revise the required actions to enhance plant reliability by 
reducing exposure to unnecessary shutdowns and increase operational 
flexibility by allowing more time to make required repairs for 
inoperable reactor trip breakers consistent with allowed outage times 
for associated logic trains. No modifications to setpoint actuations, 
trip setpoint, surveillance requirements or channel response that would 
affect the safety analyses are associated with the proposed changes.
    The proposed changes are consistent with requirements generically 
approved as part of NUREG-1431, Standard Technical Specifications, 
Westinghouse Plants, Revision 4 (TS 3.3.1, ''Reactor Trip System 
Instrumentation''). Justification for the proposed changes is based on 
Westinghouse Electric Company LLC's topical report WCAP-15376-P-A, 
Revision 1, ``Risk-Informed Assessment of the RTS [Reactor Trip System] 
and ESFAS [Engineered Safety Feature Actuation System] Surveillance 
Test Intervals and Reactor Trip Breaker Test and Completion Times,'' 
March 2003 (not publicly available; proprietary).
    This application was originally noticed in the Federal Register on 
April 8, 2014 (79 FR 19400), as a license amendment request containing 
sensitive unclassified non-safeguards information (SUNSI). However, by 
letter dated June 9, 2014, STP Nuclear Operating Company removed all 
proprietary markings from Attachment A of Enclosure 1, ``Topical Report 
Applicability Determination, ST-WN-NOC-13-46,'' originally included in 
the letter dated January 6, 2014. Therefore, the application is being 
renoticed in the Federal Register to remove the SUNSI designation.
    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 overall reactor trip breaker performance will remain within 
the bounds of the previously performed accident analyses since no 
hardware changes are proposed. The reactor trip breakers will 
continue to function in a manner consistent with the plant design 
basis.
    The proposed changes do not introduce any new accident 
initiators, and therefore do not increase the probability of any 
accident previously evaluated. There will be no degradation in the 
performance of or an increase in the number of challenges imposed on 
safety-related equipment assumed to function during an accident 
situation. There will be no change to normal plant operating 
parameters or accident mitigation performance. The proposed changes 
will not alter any assumptions or change any mitigation actions in 
the radiological consequence evaluations in the Updated Final Safety 
Analysis Report.
    The determination that the results of the proposed changes are 
acceptable was established in the NRC Safety Evaluation (issued by 
letter dated December 20, 2002) prepared for WCAP-15376-P-A, ``Risk-
Informed Assessment of the RTS and ESFAS Surveillance Test Intervals 
and Reactor Trip Breaker Test and Completion Times'' [ADAMS 
Accession No. ML023540534]. Implementation of the proposed changes 
will result in an insignificant risk impact. Applicability of these 
conclusions has been verified through plant-specific reviews and 
implementation of the generic analysis results in accordance with 
the respective NRC Safety Evaluation conditions.
    Therefore, the proposed changes do not increase the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not result in a change in the manner in 
which the Reactor Trip Breakers provide plant protection. The 
proposed changes do not change the response of the plant to any 
accidents. No design changes are associated with the proposed 
changes.
    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. No new 
accident scenarios, transient precursors, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria as 
stated in the Updated Final Safety Analysis Report are not impacted 
by these changes. Redundant Reactor Trip Breaker features and 
diverse trip features for each Reactor Trip Breaker are maintained. 
All signals credited as primary or secondary, and all operator 
actions credited in the accident analyses are unaffected by the 
proposed change. The proposed changes will not result in plant 
operation in a configuration outside the design basis. The proposed 
changes should enhance plant reliability by reducing exposure to 
unnecessary shutdowns and increase operational flexibility by 
allowing more time to make required repairs for inoperable reactor 
trip breakers. The calculated impact on risk is insignificant and 
meets the acceptance criteria contained in NRC Regulatory Guides 
1.174 [``An Approach

[[Page 45482]]

for Using Probabilistic Risk Assessment in Risk-Informed Decisions 
on Plant-Specific Changes to the Licensing Basis,'' Revision 2 
(ADAMS Accession No. ML100910006)] and 1.177 [``An Approach for 
Plant-Specific, Risk-Informed Decisionmaking: Technical 
Specifications,'' Revision 1 (ADAMS Accession No. ML100910008)].
    Therefore, the proposed changes do not result in 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: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Branch Chief: Michael T. Markley.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423, 
Millstone Power Station, Unit No. 3, New London County, Connecticut
    Date of amendment request: May 3, 2013, as supplemented by letters 
dated July 2 and October 2, 2013, and January 15 and May 28, 2014.
    Brief description of amendment: The amendment revised the Technical 
Specifications.
    Date of issuance: July 10, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 260. A publicly-available version is in ADAMS under 
Accession No. ML14178A599; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-49: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2013 (78 FR 
51225). The supplemental letters dated July 2 and October 2, 2013, and 
January 15 and May 28, 2014, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 10, 2014.
    No significant hazards consideration comments received: No.
Duke Energy Florida, Inc., et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit 3, Citrus County, Florida
    Date of amendment request: April 25, 2013, as supplemented by 
letters dated September 4, 2013, and February 26, 2014.
    Brief description of amendment: The amendment revised and removed 
certain requirements from the Section 5.0, ``Administrative Controls,'' 
portions of the Technical Specifications (TSs) that are no longer 
applicable to the facility in its permanently shutdown and defueled 
condition.
    Date of issuance: July 11, 2014.
    Effective date: As of the date of its issuance and shall be 
implemented within 30 days of issuance.
    Amendment No.: 244. A publicly-available version is in ADAMS under 
Accession No. ML14097A145; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-72: Amendment revised the 
Facility Operating License and TSs.
    Date of initial notice in Federal Register: July 23, 2013 (78 FR 
44174). The supplemental letter dated September 4, 2013, expanded the 
scope of the application as originally noticed; therefore, the staff 
re-noticed the application and included a revised proposed no 
significant hazards consideration determination on November 12, 2013 
(78 FR 67406). The supplemental letter dated February 26, 2014, 
provided additional information that clarified the supplement dated 
September 4, 2013, did not expand the scope of the application as 
noticed on November 12, 2013, and did not change the NRC staff's 
proposed no significant hazards consideration determination as 
published in the Federal Register on November 12, 2013.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 11, 2014.
    No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit No. 3, Westchester County, New York
    Date of amendment request: January 28, 2013, as supplemented by 
letters dated August 21, 2013, and April 22, 2014.
    Brief description of amendment(s): Nuclear Safety Advisory Letter 
11-5 identified Westinghouse methodology errors in the long-term mass 
and energy releases during a large break loss-of-coolant accident. 
These impacted the containment integrity analysis for Indian Point Unit 
No. 3 and required revisions to the limiting initial operating 
conditions (i.e., containment temperature, containment pressure, and 
refueling water storage tank temperature) and required revisions to 
Technical Specifications (TSs) 3.5.4, ``Refueling Water Storage Tank 
(RWST),'' and 3.6.4, ``Containment Pressure.'' In addition, revisions 
were made to TS 3.6.3, ``Containment Isolation Valves,'' to delete a 
redundant surveillance requirement and TS 5.5.15, ``Containment Leakage 
Rate Testing

[[Page 45483]]

Program,'' to reflect a slightly higher calculated containment peak 
pressure.
    Date of issuance: July 17, 2014.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 253. A publicly-available version is in ADAMS under 
Accession No. ML14169A583; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment(s).
    Facility Operating License No. DPR-64: The amendment revised the 
Facility Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: April 2, 2013 (78 FR 
19750). The supplemental letters provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 17, 2014.
    No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point 
Nuclear Generating Unit No. 2, Westchester County, New York
    Date of amendment request: January 28, 2013, as supplemented by 
letters dated August 21, 2013, and April 22, 2014.
    Brief description of amendment(s): The amendment authorizes 
revisions to the Indian Point Unit No. 2 Updated Final Safety Analysis 
Report (UFSAR) to credit four rather than three containment fan cooler 
units in the containment integrity analysis. A re-analysis of the large 
break loss-of-coolant accident was performed to correct methodology 
errors in the long-term mass and energy releases for the containment 
integrity analysis and crediting four containment fan cooler units for 
the limiting single failure is necessary to maintain the peak 
containment pressure within the current analysis of record.
    Date of issuance: July 16, 2014.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 276. A publicly-available version is in ADAMS under 
Accession No. ML14126A809; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-26: The amendment revised the 
Facility Operating License and the UFSAR.
    Date of initial notice in Federal Register: April 2, 2013 (78 FR 
19749). The supplement letters provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 16, 2014.
    No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York
    Date of application for amendment: May 7, 2013, as supplemented by 
letter dated January 17, 2014.
    Brief description of amendment: The amendment revised License 
Condition 2.T of the JAFNPP Renewed Facility Operating License to be 
consistent with the license condition contained in NUREG-1905, ``Safety 
Evaluation Report Related to the License Renewal of James A. 
FitzPatrick Nuclear Power Plant,'' dated April 2008, and to clarify 
that the programs and activities described in the Updated Final Safety 
Analysis Report Supplement and identified in Appendix A of NUREG-1905 
are to be completed no later than the start of the period of extended 
operation (PEO). The change removes any potential inference that any of 
the activities are being implemented after the PEO begins.
    Date of issuance: July 16, 2014.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 306. A publicly-available version is in ADAMS under 
Accession No. ML14086A152, documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-59: The amendment 
revised the License.
    Date of initial notice in Federal Register: April 15, 2014 (79 FR 
21297).
    The January 17, 2014, supplement provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the NRC staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 16, 2014.
    No significant hazards consideration comments received: No.
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, Unit Nos. 1 and 
2, Ogle County, Illinois
    Date of amendment request: July 23, 2012, as supplemented by letter 
dated May 1, 2013. Publicly-available versions are in ADAMS under 
Accession Nos. ML12206A057 and ML13122A046, respectively.
    Description of amendment: The amendments delete the limiting 
condition for operation Note associated with technical specifications 
(TS) Section 3.5.3, ``ECCS--Shutdown.''
    Date of issuance: July 21, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 176/182. A publicly-available version is in ADAMS 
under Accession No. ML13311B481; documents related to these amendments 
are listed in the Safety Evaluation enclosed with the amendments.
    Facility Operating License Nos. NPF-72. NPF-77, NPF-37, and NPF-66: 
The amendments revised the Technical Specifications and License.
    Date of initial notice in Federal Register: (77 FR 67682), dated 
November 13, 2012.
    The supplement letter dated May 1, 2013, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 21, 2014.
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, Unit Nos. 1 and 
2, Ogle County, Illinois
    Date of application for amendment: December 21, 2012.
    Brief description of amendment:
    The proposed amendment would revise Technical Specification (TS) 
3.3.6, ``Containment Ventilation Isolation Instrumentation.'' 
Specifically,

[[Page 45484]]

this amendment request proposes to revise Footnote (b) of TS Table 
3.3.6-1, ``Containment Ventilation Isolation Instrumentation,'' which 
specifies the ``Containment Radiation--High'' trip setpoint for two 
containment area radiation monitors (i.e., 1(2)RE-AR011 and 1(2)RE-
AR012). The proposed changes would revise the ``Containment Radiation--
High'' trip setpoint from the current, overly conservative value (i.e., 
a submersion dose rate of less than or equal to 10 milliroentgen per 
hour (mR/hr) in the containment building), to less than or equal to 2 
times the containment building background radiation reading at rated 
thermal power, which is consistent with NUREG-1431, ``Standard 
Technical Specifications, Westinghouse Plants.'' Upon reaching the 
``Containment Radiation--High'' setpoint, these area radiation monitors 
provide an isolation signal to the containment normal purge, minipurge, 
and post-loss of coolant accident systems' containment isolation 
valves.
    Date of issuance: July 21, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 165 days.
    Amendment Nos.: 178/178; 184/184. (ADAMS Accession No. ML14106A169; 
documents related to these amendments are in the Safety Evaluation 
referenced in this notice).
    Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66: 
The amendments revised the TSs and License.
    Date of initial notice in Federal Register: (78 FR 22568), dated 
April 16, 2013.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 21, 2014.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois
    Date of amendment request: September 3, 2013, (ADAMS Accession No. 
ML13246A321).
    Brief description of amendments:
    The amendments modify technical specifications (TSs) requirements 
to operate ventilation systems with charcoal filters for 10 hours, at a 
frequency specified in the Surveillance Frequency Control Program, in 
accordance with Technical Specification Task Force (TSTF)-522, Revision 
0, ``Revise Ventilation System Surveillance Requirements to Operate for 
10 hours per Month.'' A notice of the availability of TSTF-522 and a 
model safety evaluation was published in the Federal Register on 
September 20, 2012 (77 FR 58421).
    Date of issuance: July 21, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 105 days.
    Amendment Nos.: 177/177; 183/183; 201; 241/234; 208/195; 252/247. A 
publicly-available version is in ADAMS under Accession No. ML14085A532; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66, 
NPF-62, DPR-19, DPR-25, NPF-11, NPF-18, DPR-29, and DPR-30: The 
amendments revised the TSs and Licenses.
    Date of initial notice in Federal Register: December 24, 2013 (78 
FR 77732).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 21, 2014.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 28th day of July 2014.

    For the Nuclear Regulatory Commission.
A. Louise Lund,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2014-18395 Filed 8-4-14; 8:45 am]
BILLING CODE 7590-01-P