Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 74176-74188 [2013-29168]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 78, Number 237 (Tuesday, December 10, 2013)]
[Notices]
[Pages 74176-74188]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-29168]


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NUCLEAR REGULATORY COMMISSION

[NRC-2013-0266]


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

Background

    Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from November 14, 2013 to November 27, 2013. The 
last biweekly notice was published on November 26, 2013 (78 FR 70589).

ADDRESSES: You may submit comment by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0266. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section 
of this document.
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: 3WFN, 06-44M, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2013-0266 when contacting the NRC 
about the availability of information regarding this document. You may 
access publicly-available information related to this action by the 
following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0266.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly-available documents online in the NRC 
Library at https://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to pdr.resource@nrc.gov. The ADAMS accession number 
for each document referenced in this notice (if that document is 
available in ADAMS) is provided the first time that a document is 
referenced.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2013-0266 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission.

[[Page 74177]]

The NRC posts all comment submissions at https://www.regulations.gov as 
well as entering the comment submissions into ADAMS. The NRC does not 
routinely edit comment submissions to remove identifying or contact 
information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Section 50.92 of Title 10 of the Code 
of Federal Regulations (10 CFR), this means that operation of the 
facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination; any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC regulations are accessible electronically from the NRC Library on 
the NRC's Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed by the above date, the Commission or a presiding officer 
designated by the Commission or by the Chief Administrative Judge of 
the Atomic Safety and Licensing Board Panel, will rule on the request 
and/or petition; and the Secretary or the Chief Administrative Judge of 
the Atomic Safety and Licensing Board will issue a notice of a hearing 
or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in

[[Page 74178]]

accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007). 
The E-Filing process requires participants to submit and serve all 
adjudicatory documents over the internet, or in some cases to mail 
copies on electronic storage media. Participants may not submit paper 
copies of their filings unless they seek an exemption in accordance 
with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with the NRC's guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through ADAMS in the NRC's Library at https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter

[[Page 74179]]

problems in accessing the documents located in ADAMS should contact the 
NRC PDR's Reference staff at 1-800-397-4209, 301-415-4737, or by email 
to pdr.resource@nrc.gov.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: September 12, 2013.
    Description of amendment request: The proposed amendments revise 
technical specification 3.3.2, Emergency Safety Feature Actuation 
System (ESFAS) Instrumentation, to support planned plant modifications 
associated with NRC Order EA-12-049, Order Modifying Licenses with 
Regard to Requirements for Mitigation Strategies for Beyond-Design-
Basis External Events. Specifically, the amendment modifies the 
Allowable Value and Nominal Trip Setpoints listed in Table 3.3.2-1, 
Function 6.f, Auxiliary Feedwater pump suction transfer on low suction 
pressure.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1: Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed TS changes are in support of a plant modification 
involving the installation of an AC-independent AFW Suction Transfer 
scheme and hardware to ensure a continuous AFW suction source during 
an Extended Loss of AC Power (ELAP) event. The purpose of Table 
3.3.2-1 Function 6.f is to preserve the AFW pumps by ensuring a 
continuous suction supply to the pumps. The proposed change will 
cause the AFW pumps to align to the safety-related suction source 
sooner than under the current setpoint values for design basis 
events. The result of the proposed TS setpoint changes will be an 
increase in margin for AFW pump suction. The new TS setpoints were 
selected with sufficient margin for instrument uncertainty to ensure 
that the safety-related AFW suction transfer function actuates 
before the new AC independent AFW suction transfer function and to 
prevent any adverse interaction of the two schemes. In other words, 
the proposed change will ensure the safety-related suction transfer 
is initiated before the non-safety AC independent AFW suction 
transfer initiates. The specific TS changes are associated with 1) 
the specific Nominal Trip Setpoint and Allowable Values for the AFW 
Pump Suction Transfer on Suction Pressure--Low feature, 2) the 
addition of specific requirements to be taken if the as-found 
channel setpoint is outside its predefined as-found tolerance, and 
3) the addition of specific requirements regarding resetting of an 
channel setpoint within an as-left tolerance.
    The AFW Pump Suction Transfer on Suction Pressure--Low feature 
does not affect the probability of any accident being initiated. In 
addition, none of the abovementioned proposed TS changes affect the 
probability of any accident being initiated.
    Actuation of the AFW Pump Suction Transfer on Suction Pressure--
Low feature will continue to ensure that adequate AFW pump suction 
is maintained during design bases events. Transfer to the safety-
related suction source will actually occur earlier due to the 
proposed change. The proposed changes to Nominal Trip Setpoints and 
Allowable Values are based on accepted industry standards and will 
preserve assumptions in the applicable accident analyses. None of 
the proposed changes alter any assumption previously made in the 
radiological consequences evaluations, nor do they affect mitigation 
of the radiological consequences of an accident previously 
evaluated.
    In summary, the proposed changes will not involve any increase 
in the probability or consequences of an accident previously 
evaluated.

Criterion 2: Does the proposed amendment reate the possibility of a new 
or different kind of accident from any accident previously evaluated?

    Response: No.
    No new accident scenarios, failure mechanisms, or single 
failures are introduced as a result of any of the proposed changes. 
The AFW Pump Suction Transfer feature is not an accident initiator. 
No changes to the overall manner in which the plant is operated are 
being proposed. Therefore, none of the proposed changes will create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.

Criterion 3: Does the proposed amendment involve a significant 
reduction in the margin of safety?

    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their intended functions. 
These barriers include the fuel cladding, the reactor coolant system 
pressure boundary, and the containment barriers. The proposed TS 
setpoints serve to ensure proper AFW system suction transfer for 
design bases events, whereby the proposed TS changes will not have 
any effect on the margin of safety of fission product barriers. In 
addition, the proposed TS changes will not have any impact on these 
barriers. No accident mitigating equipment will be adversely 
impacted as a result of the modification. Therefore, existing safety 
margins will be preserved. None of the proposed changes will involve 
a significant reduction in a margin of safety.
    Based on the above, it is concluded that the proposed amendment 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Robert J. Pascarelli.

Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit 2, Darlington County, South Carolina

    Date of amendment request: September 10, 2013.
    Description of amendment request: The proposed change would revise 
Technical Specification Limiting Condition for Operation 3.8.1, 
Required Action (RA) B.3.2.2, ``One DG [Diesel Generator] Inoperable--
Perform SR [Surveillance Requirement] 3.8.1.2 for OPERABLE DG within 96 
hours,'' by a NOTE clarifying RA B.3.2.2 that states, ``Not required to 
be performed when the cause of the inoperable DG is pre-planned 
maintenance and testing.''
    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 eliminates a conditional surveillance of the 
Operable EDG [emergency diesel generator] whenever the alternate 
division EDG is out of service for pre-planned maintenance and 
testing. The EDG are [is] not an initiator of any accident 
previously evaluated. As a result, the probability of any accident 
previously evaluated is not significantly increased.
    The consequences of any accident previously evaluated are not 
increased, as the EDG will continue to meet its safety function to 
supply backup AC [alternating current] power as specified in the 
accident analysis, in a highly reliable manner, as a common cause 
problem between the two EDGs will have been precluded, the alternate 
division EDG will no longer be taken out of service for testing, and 
its normally scheduled surveillances will be met.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 74180]]

    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. The changes do not alter assumptions made in the safety 
analysis for EDG performance.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change eliminates a conditional surveillance of the 
Operable EDG whenever the alternate division EDG is out of service 
for pre-planned maintenance and testing. The EDG will continue to 
meet its specified safety function in the safety analysis to provide 
backup AC power, in a highly reliable manner, as a common cause 
problem between the two EDGs will have been precluded, the alternate 
division EDG will no longer be taken out of service for testing, and 
its normally scheduled surveillances will be met.
    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: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tryon Street, Charlotte, NC 28202.
    NRC Branch Chief: Jessie F. Quichocho.

Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit 2, Darlington County, South Carolina

    Date of amendment request: September 30, 2013.
    Description of amendment request: The proposed amendment implements 
the Nuclear Regulatory Commission (NRC)-approved Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler, TSTF-491, ``Removal of Main Steam and Main Feedwater Valve 
Isolation Times from Technical Specifications,'' via the Consolidated 
Line Item Improvement Process (CLIIP). This request will modify the 
current Unit 2 Technical Specifications (TSs) 3.7.2, Main Steam 
Isolation Valves and 3.7.3, Main Feedwater Isolation Valves, Main 
Feedwater Regulation Valves and Bypass Valves by relocating the 
specific isolation time for the isolation valves from the associated 
Surveillance Requirements (SRs). The isolation time in the TS SRs is 
replaced with the requirement to verify the valve isolation time is 
``within limits.'' The specific isolation times will be maintained in 
the Unit 2 Technical Requirements Manual.
    The NRC staff published a notice of opportunity for comment in the 
Federal Register on October 5, 2006 (71 FR 58884), on possible 
amendments adopting TSTF-491, Revision 2, including a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination, using the CLIIP. The NRC staff subsequently issued a 
notice of availability of the models for referencing in license 
amendment applications in the Federal Register on December 29, 2006 (71 
FR 78472). The licensee affirmed the applicability of the following 
NSHC determination in its application dated September 30, 2013.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The proposed change allows relocating main steam and main 
feedwater valve isolation times to the Licensee Controlled Document 
that is referenced in the Bases. The proposed change is described in 
Technical Specification Task Force (TSTF) Standard TS Change 
Traveler TSTF-491 related to relocating the main steam and main 
feedwater valves isolation times to the Licensee Controlled Document 
that is referenced in the Bases and replacing the isolation time 
with the phase, ``within limits.''
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
The proposed changes relocate the main steam and main feedwater 
isolation valve times to the Licensee Controlled Document that is 
referenced in the Bases. The requirements to perform the testing of 
these isolation valves are retained in the TS. Future changes to the 
Bases or licensee-controlled document will be evaluated pursuant to 
the requirements of 10 CFR 50.59, ``Changes, test and experiments,'' 
to ensure that such changes do not result in more than minimal 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological consequences of any accident previously 
evaluated. Further, the proposed changes do not increase the types 
and the amounts of radioactive effluent that may be released, nor 
significantly increase individual or cumulative occupation/public 
radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.

Criterion 2: The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed changes relocate the main steam and main feedwater 
valve isolation times to the Licensee Controlled Document that is 
referenced in the Bases. In addition, the valve isolation times are 
replaced in the TS with the phase ``within limits.'' The changes do 
not involve a physical altering of the plant (i.e., no new or 
different type of equipment will be installed) or a change in 
methods governing normal pant operation. The requirements in the TS 
continue to require testing of the main steam and main feedwater 
isolation valves to ensure the proper functioning of these isolation 
valves.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.

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

    The proposed changes relocate the main steam and main feedwater 
valve isolation times to the Licensee Controlled Document that is 
referenced in the Bases. In addition, the valve isolation times are 
replaced in the TS with the phase ``within limits.'' Instituting the 
proposed changes will continue to ensure the testing of main steam 
and main feedwater isolation valves. Changes to the Bases or license 
controlled document are performed in accordance with 10 CFR 50.59. 
This approach provides an effective level of regulatory control and 
ensures that main steam and feedwater isolation valve testing is 
conducted such that there is no significant reduction in the margin 
of safety.
    The margin of safety provided by the isolation valves is 
unaffected by the proposed changes since there continue to be TS 
requirements to ensure the testing of main steam and main feedwater 
isolation valves. The proposed changes maintain sufficient controls 
to preserve the current margins of safety.

    The NRC staff proposes to determine that the amendment request 
involves NSHC.
    Attorney for licensee: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tryon Street, Charlotte, NC 28202.

[[Page 74181]]

    NRC Branch Chief: Jessie F. Quichocho.

Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: July 29, 2013.
    Description of amendment request: The amendment would add a 
permanent exception to the River Bend Station (RBS) Technical 
Requirements Manual (TRM) Section 3.9.14, ``Crane Travel--Spent and New 
Fuel Storage, Transfer, and Upper Containment Fuel Pools,'' to allow 
for movement of fuel pool gates over fuel assemblies for maintenance. 
This exception will also be described by revision to the RBS Updated 
Safety Analysis Report (USAR) Section 9.1.2.2.2, ``Fuel Building Fuel 
Storage,'' and Section 9.1.2.3.3, ``Protection Features of Spent Fuel 
Storage Facilities.''
    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. Involved a significant increase in the probability or 
consequences of an accident previously evaluated.
    Response: No.
    The RBS fuel building fuel storage facilities consist of three 
interconnected stainless steel-lined concrete pools. The spent fuel 
storage pool is the largest of these pools. Adjacent to the fuel 
storage pool are the cask pool and the lower IFTS [inclined fuel 
transfer system] pool. Each of these two pools is separated from the 
fuel storage pool by a full-height wall encompassing a watertight 
gate. The watertight gates are normally open, but are closed to seal 
their respective pools during cask handling and equipment 
maintenance operations. It is necessary to lift the gates from the 
pools for maintenance or seal replacement. The total weight of the 
gate including the rigging equipment is 2000 pounds. This lift is 
considered as a heavy load lift since it is higher than the current 
analyzed light load limit of 1200 pounds for movement of loads over 
fuel assemblies. TRM 3.9.14 prohibits any load in excess of 1200 
pounds from travel over fuel assemblies in the storage pool.
    Each of the gates is designed with a pneumatic seal that, when 
pressurized, seals the respective pool from the spent fuel pool, 
forming a watertight barrier. No provisions for moving the gates 
over fuel assemblies were included in the current licensing basis 
for RBS heavy loads. However, the service life qualification of the 
gate seals necessitates that they be replaced several times over the 
life of the plant. Therefore, approval of an exception to the 
current prohibition is required to allow for replacement of the gate 
seals.
    To perform the movement of the gate from its installed position 
to a position where the seal can be replaced, an engineering plan 
that meets the intent of the applicable regulatory guidance has been 
developed. RBS' program for control of heavy load movements complies 
with that guidance, and this will prevent the gate from dropping 
onto the spent fuel assemblies during the movement activity. The 
program features include the design of the lifting devices, design 
of the cask and fuel bridge cranes, crane operator training, and the 
use of written procedures. The regulatory guidance will be met in 
all respects, except that, in lieu of a single failure-proof crane, 
the method will employ redundant and diverse means to meet the 
intent of single-failure proof movements.
    Entergy proposes to lift the spent fuel pool gate using a 
rigging method that complies with the intent of the guidance of 
References 10.c through 10.f [of the licensee's letter dated July 
29, 2013]. The proposed method will be accomplished through the use 
of fuel building bridge crane and the cask crane at the same time to 
provide the redundancy required to make the lift single-failure 
proof and satisfy single-failure proof criteria.
    In the proposed method, the fuel building bridge crane and the 
cask crane will be used to perform the gate lifting and movement. 
The intent of the applicable regulatory guidance is that in lieu of 
providing a single-failure-proof crane system, the control of heavy 
loads guidelines can be satisfied by establishing that the potential 
for a heavy load drop is extremely small. The gate lifting using the 
bridge crane and cask crane will conform to applicable regulatory 
guidelines, in that the probability of the gate drop over the spent 
fuel assemblies is extremely small. Both cranes have a rated 
capacity of 15 tons. The maximum weight of the gate and rigging is 
2000 pounds. Therefore, there is ample safety factor margin for 
lifting and movements of the subject spent fuel pool gate. Special 
lifting devices, which have redundancy or ultimate strength of at 
least ten times the lifted load, will also be utilized during the 
rigging process. Even though neither the fuel building bridge crane 
or the cask crane is a single-failure proof crane, rigging the spent 
fuel pool gate using both cranes will provide the required 
redundancy that meets the intent of single-failure proof criteria.
    The proposed load lift of the fuel pool gate for replacement of 
the seal conforms to all of the applicable regulatory guidelines. 
The design of the lifting lugs and associated rigging (e.g., chains, 
slings, shackles, hoists, etc.) conforms to the guidelines of NUREG-
0612, [``Control of Heavy Loads at Nuclear Power Plants,''] Section 
5.1.6, and ``Single-Failure Proof Handling System,'' and References 
10.d through 10.f [of the licensee's letter dated July 29, 2013]. 
The auxiliary hook of the cask crane has a rated capacity of 15 
tons. The cask crane is not a single-failure-proof crane. However, 
it meets NUREG-0612 criteria of Section 5.1.1(6) and is designed for 
seismic loading. As discussed above, the cask crane, alone, will 
handle the gate only after the gate is located inside the cask pool 
where drop of the gate above the spent fuel rack is no longer a 
concern. The cask pool area has been evaluated for an accidental 
drop of the spent fuel cask. There is no safety-related equipment 
inside the cask pool. The analyzed maximum weight of the gate and 
rigging is 2500 pounds. Therefore, there is ample safety factor 
margin for lifting the gate with the cask crane.
    The probability and consequences of a seismic event are not 
affected by the proposed gate lift. The consequences of a seismic 
event during the gate lifting are insignificant since both cranes, 
the fuel building bridge crane and the cask crane, are seismically 
qualified for the lifted load. In addition, the design of all 
rigging conforms to NUREG-0612 guidelines, with a safety factor of 
10 for the weight of the load.
    Consistent with the defense-in-depth approach outlined in the 
guidance, the movement will be conducted according to load handling 
instructions. Operator training will be conducted on the activity 
prior to the movement, and the equipment will be inspected before 
the movement will be performed. NUREG-0612 gives guidance that when 
a particular heavy load must be brought over spent fuel, alternative 
measures may be used. The combination of preventative measures, as 
proposed, minimizes the risks inherent in hauling large loads over 
spent fuel to permissible levels. Considering these provisions and 
the applicable regulatory guidance, the increase in probability of a 
load drop is negligible.
    It is therefore concluded that the proposed gate lifting and 
movement does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Response: No.
    The lifting of the fuel pool gate in the spent fuel pool as 
described above minimizes the possibility of a heavy load drop onto 
spent fuel assemblies as not credible in accordance with single-
failure-proof criteria. In addition, movement of the gate in the 
cask pool using the cask crane does not create the possibility of a 
new or different kind of accident. The cask drop accident scenario 
in the current RBS licensing basis (since the cask crane is not a 
single-failure-proof crane) envelops the accidental drop of the gate 
in the cask pool during handling by the cask crane. The analyzed 
weight of a cask is 125 tons, as compared to the 1 ton combined 
weight of the gate and the rigging.
    It is therefore concluded that the proposed gate lifting does 
not create the possibility of a new or different kind of accident 
from any previously analyzed.
    3. Invoke a significant reduction in a margin of safety.
    Response: No.
    By following the guidance of References 10.c through 10.f [of 
the licensee's letter dated July 29, 2013], the movement of the 
spent fuel pool gates will have no impact on the analyses of 
postulated design basis events for RBS. The NRC guidance provides an 
acceptable means of ensuring the appropriate level of safety and 
protection against load drop accidents. Therefore, there is no 
reduction in the margin of safety associated

[[Page 74182]]

with postulated design basis events at RBS in allowing the proposed 
change to the RBS licensing basis. RBS will continue to meet its 
commitment to comply with the applicable guidance.

    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 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Douglas A. Broaddus.

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

    Date of amendment request: September 5, 2013.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 5.5.13, ``Primary Containment Leakage 
Rate Testing Program,'' to increase the peak calculated primary 
containment internal pressure, Pa, from 39.9 psig to 42.6 
psig. The proposed increase in Pa reflects a lower initial 
drywell temperature and a number of other modeling changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided on 
September 5, 2013, 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 to Pa does not alter the assumed 
initiators to any analyzed event. The probability of an accident 
previously evaluated will not be increased by this proposed change 
since this change does not modify the plant or how it is operated.
    The change in Pa will not affect radiological dose 
consequence analyses. LSCS radiological dose consequence analyses 
are based on the maximum allowable containment leakage rate. Even 
though the test pressure at which leak rate testing is performed is 
Pa, the maximum allowable containment leakage rate is 
defined in terms of a percentage of weight of the original content 
of containment air, which is independent of the peak calculated 
primary containment internal pressure. The Appendix J containment 
leak rate testing program will continue to ensure that containment 
leakage remains within the leakage assumed in the offsite dose 
consequence analyses. The consequences of an accident previously 
evaluated will not be increased by this proposed change.
    Therefore, operation of the facility in accordance with the 
proposed change to Pa will 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 provides a higher Pa than 
currently described in the TS. This change is the result of a LOCA-
Drywell Temperature sensitivity analysis performed by General 
Electric Hitachi. The peak calculated primary containment internal 
pressure remains below the containment design pressure of 45 psig. 
This change does not involve any alteration in the plant 
configuration (no new or different type of equipment will be 
installed) or make changes in the methods governing normal plant 
operation. The change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, operation of the facility in accordance with the 
proposed change to TS 5.5.13 would 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 peak calculated primary containment internal pressure 
remains below the containment design pressure of 45 psig. LSCS 
radiological consequence analyses are based on the maximum allowable 
containment leakage rate. The change in the peak calculated primary 
containment internal pressure does not represent a significant 
change in the margin of safety. Operation of the facility in 
accordance with the proposed change to TS 5.5.13 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Tamra Domeyer, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Travis L. Tate.

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

    Date of amendment request: September 20, 2013.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3.3.8.1-1, ``Loss of Power 
Instrumentation,'' Table 1, to change the allowable values to address 
non-conservative assumptions. The proposed change involves revising the 
surveillance requirements to modify the allowable values for the 4.16 
kV emergency buses during loss of voltage testing and calibration to 
ensure that existing design requirements remain satisfied.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided on 
September 20, 2013, 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 to the 4.16 kV [engineered safety functions] 
ESF bus loss of voltage allowable values allow the protection scheme 
to function as originally designed. (This change will involve 
alteration of nominal trip setpoints in the field and will also be 
reflected in revisions to the calibration procedures.) The proposed 
change does not affect the probability or consequences of any 
accident. Analysis was conducted and demonstrates that the proposed 
allowable values will allow the normally operating safety-related 
motors to continue to operate without sustaining damage or tripping 
during the worst-case, non-accident degraded voltage condition for 
the maximum possible time-delay of 5.7 minutes. Thus, these safety-
related loads will be available to perform their safety function if 
a loss-of-coolant accident (LOCA) concurrent with a loss-of-offsite 
power (LOOP) occurs following the degraded voltage condition.
    The proposed changes do not adversely affect accident initiators 
or precursors, and do not alter the design assumptions, conditions, 
or configuration or the plant or the manner in which the plant is 
operated or maintained. The proposed allowable values ensure that 
the 4.16 kV distribution system remains connected to the offsite 
power system when adequate offsite voltage is available and motor 
starting transients are considered. The diesel start due to a LOCA 
signal is not adversely affected by this change. During an actual 
loss of voltage condition, the loss of voltage time delay will 
continue to isolate the 4.16 kV distribution system from offsite 
power before the diesel is ready to assume the emergency loads, 
which is the limiting time basis for mitigating system responses to 
the accident. For this reason, the existing loss of power/LOCA 
analysis continues to be valid.
    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.

[[Page 74183]]

    The proposed change involves the revision of 4.16 kV ESF bus 
loss of voltage allowable values to satisfy existing design 
requirements. The proposed change does not introduce any changes or 
mechanisms that create the possibility of a new or different kind of 
accident. The proposed change does not install any new or different 
type of equipment, and installed equipment is not being operated in 
a new or different manner. No new effects on existing equipment are 
created nor are any new malfunctions introduced.
    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 protection voltage allowable values are low enough 
to prevent inadvertent power supply transfer, but high enough to 
ensure that sufficient power is available to the required equipment. 
The diesel start due to a LOCA signal is not adversely affected by 
this change. During an actual loss of voltage condition, the loss of 
voltage time delays will continue to isolate the 4.16 kV 
distribution system from offsite power before the diesel is ready to 
assume the emergency loads.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Tamra Domeyer, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Travis L. Tate.

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

    Date of amendment request: October 10, 2013.
    Description of amendment request: The proposed amendment would 
revise the date for the performance of the containment leakage rate 
Type A test from ``no later than May 4, 2014,'' to ``prior to entering 
MODE 4 at the start of Cycle 18.'' Additionally, EGC is proposing to 
establish a requirement for Braidwood Station, Unit 2, to exit the 
MODEs of applicability for Containment as described in Technical 
Specification 3.6.1, ``Containment'' (i.e., MODEs 1-4), no later than 
May 4, 2014.
    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:

    EGC has evaluated the proposed change for Braidwood Station, 
Units 1 and 2 using the criteria in 10 CFR 50.92, and has determined 
that the proposed change does not involve a significant hazards 
consideration. The following information is provided to support a 
finding of no significant hazards consideration.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the Braidwood Station, Units 1 and 2 
Containment Leakage Rate Testing Program does not involve a physical 
change to the plant or a change in the manner in which the plant is 
operated or controlled. The containment function is to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the containment itself, and the testing requirements to periodically 
demonstrate the integrity of the containment, exist to ensure the 
plant's ability to mitigate the consequences of an accident do not 
involve any accident precursors or initiators. Therefore, the 
probability of occurrence of an accident previously evaluated is not 
significantly increased by the proposed amendment. Implementation of 
the proposed change will continue to provide adequate assurance that 
during design basis accidents, the containment and its components 
would limit leakage rates to less than the values assumed in the 
plant safety analyses. Therefore, the consequences of an accident 
previously evaluated will not be increased by this proposed change.
    Therefore, operation of the facility in accordance with the 
proposed administrative change to the date for the performance of 
the Unit 2, Type A containment leak rate test will 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 containment, and the testing requirements to periodically 
demonstrate the integrity of the containment, exist to ensure the 
plant's ability to mitigate the consequences of an accident, and do 
not involve any accident precursors or initiators. The proposed 
change does not involve a physical change to the plant (i.e., no new 
or different type of equipment will be installed) or a change to the 
manner in which the plant is currently operated or controlled.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    This proposed change does not alter the manner in which safety 
limits, limiting safety system setpoints, or limiting conditions for 
operation are determined. The specific requirements and conditions 
of the containment leakage rate testing program, as proposed, will 
continue to ensure that the degree of containment structural 
integrity and leak-tightness that is considered in the plant's 
safety analysis is maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above evaluation, EGC concludes that the proposed 
amendment does not involve a significant hazards consideration under 
the standards set forth in 10 CFR 50.92, paragraph (c), and 
accordingly, a finding of no significant hazards consideration is 
justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Travis L. Tate.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: June 6, 2013.
    Description of amendment request: The proposed amendment would 
change the current requirement that ``each ADS [Automatic 
Depressurization System] valve opens when manually actuated,'' to the 
requirement that ``each ADS valve actuator strokes when manually 
actuated.'' Additionally, the surveillance frequency would change from 
``24 months on a STAGGERED TEST BASIS for each valve solenoid,'' to 
``24 months.''
    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 change does not modify the method of demonstrating 
the operability of the Safety/Relief Valves (S/RVs) in both the 
safety and relief modes of operation. The proposed change does 
modify the method for demonstrating the proper mechanical 
functioning of the S/RVs. The S/RVs are required to function in the 
safety mode to prevent overpressurization of the reactor vessel and 
reactor coolant system pressure

[[Page 74184]]

boundary during various analyzed transients, including Main Steam 
Isolation Valve closure. S/RVs associated with the Automatic 
Depressurization System are also required to function in the relief 
mode to reduce reactor pressure to permit injection by low pressure 
Emergency Core Cooling System (ECCS) pumps during certain reactor 
coolant pipe break accidents. The current testing method 
demonstrates the proper mechanical functioning of the S/RVs in both 
modes through manual actuation of the S/RVs. The proposed testing 
method results in acceptable demonstration of the S/RV functions in 
both the safety and relief modes, and therefore provides assurance 
that the probability of S/RV failure will not increase. None of the 
accident safety analyses are affected by the requested [Technical 
Specification] TS changes and the consequences of accidents 
mitigated by the S/RVs will not increase.
    Therefore, the proposed amendment does not result in a 
significant increase in the probability or consequences of any 
previously evaluated accident.
    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 change modifies the method of testing of the S/RVs, 
but does not alter the functions or functional capabilities of the 
S/RVs. Testing under the proposed method is performed in offsite 
test facilities and in the plant during outage periods when the S/RV 
functions are not required. Existing analyses address events 
involving an S/RV inadvertently opening or failing to reclose. 
Analyses also address the failure of one or more S/RVs to open. The 
proposed change does not introduce any new failure mode.
    Therefore, it does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment provides for a complete verification of 
the functional capability of the S/RVs by performing tests, 
inspections, and maintenance activities without opening the valves 
while installed in the plant. This alternative testing and 
associated programmatic controls will provide an overall level of 
assurance that the S/RVs are capable of performing their intended 
accident mitigation safety functions. The proposed amendment does 
not affect the valve setpoints or adversely affect any other 
operational criteria assumed for accident mitigation. No changes are 
proposed that alter the setpoints at which protective actions are 
initiated, and there is no change to the operability requirements 
for equipment assumed to operate for accident mitigation. Moreover, 
it is expected that the alternative testing methodology will 
increase the margin of safety by reducing the potential for S/RV 
leakage resulting from testing. Additionally, the increased testing 
frequency of the manual actuation circuitry is beneficial since the 
valves will no longer be tested on a staggered test frequency.
    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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    Acting NRC Branch Chief: John G. Lamb.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: June 6, 2013.
    Description of amendment request: This proposed change adds a 
footnote to Function 6c in Technical Specification Table 3.3.6.1-1. 
This change allows only one Trip System to be operable in MODES 4 and 5 
for the Manual Initiation Function for Shutdown Cooling System 
isolation.
    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 manual isolation function of the RHR [Residual Heat Removal] 
Shutdown Cooling System is not credited in any FSAR [Final Safety 
Analysis Report] safety analysis. The addition of Footnote (c) to 
the manual isolation function in TS [Technical Specification] Table 
3.3.6.1-1 allows one of the two trip systems to be inoperable in 
MODES 4 and 5 and does not alter any equipment.
    Therefore, this proposed change does 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 addition of Footnote (c) to the manual isolation function in 
TS Table 3.3.6.1-1 allows one of the two trip systems to be 
inoperable in MODES 4 and 5 and is consistent with other isolation 
function required for isolation in MODES 4 and 5.
    No new equipment is being introduced, and installed equipment is 
not being operated in a new or different manner. There are no set 
points, at which protective or mitigative actions are initiated, 
affected by this change. These changes do not alter the manner in 
which equipment operation is initiated, nor will the function 
demands on credited equipment be changed. No alterations in the 
procedures that ensure the plant remains within analyzed limits are 
being proposed, and no major changes are being made to the 
procedures relied upon to respond to an off-normal event as 
described in the FSAR. As such, no new failure modes are being 
introduced. The proposed change does not alter assumptions made in 
the safety analysis and licensing basis since the manual isolation 
function of the RHR Shutdown Cooling System is not credited in any 
FSAR safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes are acceptable since no 
automatic isolation functions are being changed. Since the manual 
isolation function of the RHR Shutdown Cooling System is not 
credited in any FSAR safety analysis, this change does not affect 
the margin of safety assumed by the safety analysis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179
    Acting NRC Branch Chief: John G. Lamb.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: October 2, 2013 (TS-SQN-13-01 and 13-
02).
    Description of amendment request: The proposed amendments would 
revise Units 1 and 2 Technical Specifications (TSs) 3.7.5, ``Ultimate 
Heat Sink,'' to place additional limitations on the maximum average 
Essential Raw Cooling Water (ERCW) System supply header water 
temperature during operation with one ERCW pump per loop and operation 
with one ERCW supply strainer per

[[Page 74185]]

loop. In addition, the one-time limitations on Unit 1 ultimate heat 
sink (UHS) temperature and the associated license condition 
requirements used for the Unit 2 steam generator replacement project 
are proposed to be deleted. The proposed changes would place additional 
temperature limitations on the UHS TS Limiting Condition for Operation 
3.7.5 with associated required actions, to support maintenance on plant 
component without requiring a dual unit shutdown.
    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 determination, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The proposed change to impose additional limits on UHS 
temperature while in certain ERCW system alignments does not result 
in any physical changes to plant safety-related structures, systems, 
or components (SSCs). The UHS and associated ERCW system function is 
to remove plant system heat loads during normal and accident 
conditions. As such, the UHS and ERCW system are not accident 
initiators, but instead perform accident mitigation functions by 
serving as the heat sink for safety-related equipment to ensure the 
conditions and assumptions credited in the accident analyses are 
preserved. During operation under the proposed change with only one 
ERCW pump operable in a loop a single failure could cause a total 
loss of ERCW flow in one loop whereas with two pumps per loop 
operable only a reduction in flow would occur. In either case, one 
pump or two pumps per loop operable, the other ERCW loop will 
continue to perform the design function of the ERCW system. 
Therefore, the proposed change does not involve a significant 
increase in the probability of an accident previously evaluated.
    The purpose of this change is to modify the UHS TS to be 
consistent with the conditions and assumptions of the current design 
basis heat transfer and flow modeling analyses for the UHS and ERCW 
system. The proposed change provides assurance that the minimum 
conditions necessary for the UHS and ERCW system to perform their 
heat removal safety function is maintained. Accordingly, as 
demonstrated by TVA design heat transfer and flow modeling 
calculations, the proposed new requirements will provide the 
necessary assurance that fuel cladding, Reactor Coolant System (RCS) 
pressure boundary, and containment integrity limits are not 
challenged during worst-case post-accident conditions. Accordingly, 
the conclusions of the accident analyses will remain as previously 
evaluated such that there will be no significant increase in the 
post-accident dose consequences.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequence 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 involve any physical changes to 
plant safety related SSCs or alter the modes of plant operation in a 
manner that is outside the bounds of the current UHS and ERCW system 
design heat transfer and flow modeling analyses. The proposed 
additional limits on UHS temperature for the specified ERCW system 
alignments provide assurance that the conditions and assumptions 
credited in the accident analyses are preserved. Thus, although the 
specified ERCW system alignments result in reduced heat transfer 
flow capability, the plant's overall ability to reject heat to the 
UHS during normal operation, normal shutdown, and hypothetical 
worst-case accident conditions will not be significantly affected by 
this proposed change. Since the safety and design requirements 
continue to be met and the integrity of the RCS pressure boundary is 
not challenged, no new credible failure mechanisms, malfunctions, or 
accident initiators are created, and there will be no effect on the 
accident mitigating systems in a manner that would significantly 
degrade the plant's response to an accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change modifies the UHS TS to maintain the UHS 
temperature and associated ERCW system flows within the bounds of 
the conditions and assumptions credited in the accident analyses. As 
demonstrated by TVA design basis heat transfer and flow modeling 
calculations, the additional limits on UHS temperature for the 
specified ERCW system alignments will provide assurance that the 
design limits for fuel cladding, RCS pressure boundary, and 
containment integrity are not exceeded under both normal and post-
accident conditions. As required, these calculations include 
evaluation of the worst-case combination of meteorology and 
operational parameters, and establish adequate margins to account 
for measurement and instrument uncertainties. While operating 
margins have been reduced by the proposed change in order to support 
necessary maintenance activities, the current limiting design basis 
accidents remain applicable and the analyses conclusions remain 
bounding such that the accident safety margins are maintained. 
Accordingly, the proposed change will not significantly degrade the 
margin of safety of any SSCs that rely on the UHS and ERCW system 
for heat removal to perform their safety related functions.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Jessie F. Quichocho.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant 
(WBN), Unit 1, Rhea County, Tennessee

    Date of amendment request: July 30, 2013.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 4.3.1.1, ``Criticality,'' to 
clarify the requirements for storage of new and spent fuel assemblies 
in the spent fuel racks. This change is necessary to update the current 
WBN Unit 1 TS to ensure consistency with the proposed TS 4.3.1.1 for 
WBN Unit 2. In addition, editorial changes are being made to TS 4.3.1. 
The proposed changes also modify the current licensing basis, as 
described in Section 4.3.2.7 of the Updated Final Safety Analysis 
Report (UFSAR).
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The proposed amendment directs the operators to directly use an 
existing control figure in the TS instead of a conflicting wording 
of slightly lower fuel storage enrichment limit in the same section 
of the TS. No change is being made to the parameters or methodology 
in evaluated accidents. As a result, there is no increase in the 
likelihood of existing event initiators.
    This figure was supported by the original analyses that 
determines the subcriticality available in the spent fuel pool and 
the associated acceptable cell loading patterns have not been 
changed. Thus the acceptance criteria as stated in the UFSAR are 
met. Implementing the change involves no facility equipment, 
procedure, or process changes that could affect the radioactive 
material actually released during an event. As a result, no 
conditions have been created that could

[[Page 74186]]

significantly increase the consequences of any of the events 
evaluated in the UFSAR.
    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 do not require any new or different 
accidents to be postulated because no changes are being made to the 
plant that would introduce any new accident causal mechanism. This 
license amendment request does not affect any plant systems that are 
potential accident initiators. The change in TS wording is 
consistent with an existing figure in the same section of the TS 
that is bounded by the original plant spent fuel pool criticality 
analysis. No change to the fuel, spent fuel racks, or spent fuel 
pool water chemistry are associated with this change.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment directs the operators to directly use an 
existing control figure in the TS instead of a conflicting wording 
of slightly lower fuel storage enrichment limit in the same section 
of the TS. The change in TS wording is consistent with an existing 
figure in the same section of the TS which is bounded the original 
plant spent fuel pool criticality analysis. The proposed changes do 
not alter the permanent plant design, including instrument set 
points.

    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Jessie F. Quichocho.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant 
(WBN), Unit 1, Rhea County, Tennessee

    Date of amendment request: August 28, 2013.
    Description of amendment request: The proposed changes would modify 
WBN, Unit 1 Technical Specifications (TS) requirements related to 
direct current (DC) electrical systems. In addition, a new ``Battery 
Monitoring and Maintenance Program'' is being proposed. The proposed TS 
changes place requirements on the battery itself rather than the 
battery cells as currently required.
    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 consequence of an accident previously evaluated?
    Response: No.
    The proposed changes restructure the Technical Specifications 
(TS) for the direct current (DC) electrical power system and are 
consistent with Technical Specifications Task Force (TSTF) change 
TSTF-360, Revision 1 and TSTF-500, Revision 2. The proposed changes 
modify TS Actions relating to battery and battery charger 
inoperability. The DC electrical power system, including associated 
battery chargers, is not an initiator of any accident sequence 
analyzed in the Updated Final Safety Analysis Report (UFSAR). 
Rather, the DC electrical power system supports equipment used to 
mitigate accidents. The proposed changes to restructure TS and 
change surveillances for batteries and chargers to incorporate the 
updates included in TSTF-360, Revision 1 as updated by TSTF-500, 
Revision 2, will maintain the same level of equipment performance 
required for mitigating accidents assumed in the UFSAR. Operation in 
accordance with the proposed TS would ensure that the DC electrical 
power system is capable of performing its specified safety function 
as described in the UFSAR. Therefore, the mitigating functions 
supported by the DC electrical power system will continue to provide 
the protection assumed by the analysis. The relocation of preventive 
maintenance surveillances, and certain operating limits and actions, 
to a licensee controlled Battery Monitoring and Maintenance Program 
will not challenge the ability of the DC electrical power system to 
perform its design function. Appropriate monitoring and maintenance 
that are consistent with industry standards will continue to be 
performed. In addition, the DC electrical power system is within the 
scope of 10 CFR 50.65, ``Requirements for monitoring the 
effectiveness of maintenance at nuclear power plants,'' which will 
ensure the control of maintenance activities associated with the DC 
electrical power system.
    The integrity of fission product barriers, plant configuration, 
and operating procedures as described in the UFSAR will not be 
affected by the proposed changes. Therefore, the consequences of 
previously analyzed accidents will not increase by implementing 
these changes.
    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 involve restructuring the TS for the DC 
electrical power system. The DC electrical power system, including 
associated battery chargers, is not an initiator to any accident 
sequence analyzed in the UFSAR. Rather, the DC electrical power 
system supports equipment used to mitigate accidents. The proposed 
changes to restructure the TS and change surveillances for batteries 
and chargers to incorporate the updates included in TSTF-360 
Revision 1 as updated by TSTF-500, Revision 2, will maintain the 
same level of equipment performance required for mitigating 
accidents assumed in the UFSAR. Administrative and mechanical 
controls are in place to ensure the design and operation of the DC 
systems continues to meet the plant design basis describe in the 
UFSAR.
    Therefore, the proposed amendment will 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 margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The equipment margins will be maintained in 
accordance with the plant-specific design bases as a result of the 
proposed changes. The proposed changes will not adversely affect 
operation of plant equipment. These changes will not result in a 
change to the setpoints at which protective actions are initiated. 
Sufficient DC capacity to support operation of mitigation equipment 
is ensured. The changes associated with the new battery Maintenance 
and Monitoring Program will ensure that the station batteries are 
maintained in a highly reliable manner. The equipment fed by the DC 
electrical sources will continue to provide adequate power to 
safety-related loads in accordance with analysis assumptions. TS 
changes made to be consistent with the changes in TSTF-360, Revision 
1, as updated by TSTF-500, Revision 2, maintain the same level of 
equipment performance stated in the UFSAR and the current TSs.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
    NRC Branch Chief: Jessie F. Quichocho.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: September 23, 2013.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 5.6.5, ``CORE OPERATING LIMITS REPORT 
(COLR),'' to replace WCAP-11596-P-A, ``Qualification of the Phoenix-P/
ANC Nuclear Design System for Pressurized Water Reactor Cores,'' with 
WCAP-16045-P-A, ``Qualification of the Two-

[[Page 74187]]

Dimensional Transport Code PARAGON,'' and WCAP-16045-P-A, Addendum 1-A, 
``Qualification of the NEXUS Nuclear Data Methodology,'' to determine 
core operating limits.
    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 analytical methodologies, which this license amendment 
proposes for determination of core operating limits, are 
improvements over the current methodologies in use at WCGS. The NRC 
staff reviewed and approved these methodologies and concluded that 
these analytical methods are acceptable as a replacement for the 
current analytical method. Thus core operating limits determined 
using the proposed analytical methods continue to assure that the 
reactor operates safely and, thus, the proposed changes do not 
involve an increase in the probability of an accident.
    Operation of the reactor with core operating limits determined 
by use of the proposed analytical methods does not increase the 
reactor power level, does not increase the core fission product 
inventory, and does not change any transport assumptions. Therefore 
the proposed methodology and TS changes do 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 provides revised analytical methods for 
determining core operating limits, and does not change any system 
functions or maintenance activities. The change does not involve 
physical alteration of the plant, that is, no new or different type 
of equipment will be installed. The change does not alter 
assumptions made in the safety analyses but ensure that the core 
will operate within safe limits. This change does not create new 
failure modes or mechanisms that are not identifiable during 
testing, 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 margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes do not physically alter safety-
related systems, nor does it affect the way in which safety related 
systems perform their functions. The setpoints at which protective 
actions are initiated are not altered by the proposed changes. 
Therefore, sufficient equipment remains available to actuate upon 
demand for the purpose of mitigating an analyzed event. The proposed 
analytical methodology is an improvement that allows more accurate 
modeling of core performance. The NRC has reviewed and approved this 
methodology for use in lieu of the current methodology; thus, the 
margin of safety is not reduced due to this change.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.

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

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

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

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529; and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of application for amendment: December 12, 2012.
    Brief description of amendment: The amendments revised the 
Technical Specifications (TSs) relating to reactor coolant system (RCS) 
activity limits by replacing the current TS limits on primary coolant 
gross specific activity with limits on primary coolant noble gas 
activity. The noble gas activity would reflect a new DOSE EQUIVALENT 
XE-133 definition that would replace the current E-bar average 
disintegration energy definition. The changes are consistent with NRC-
approved Industry/Technical Specifications Task Force (TSTF) Standard 
Technical Specification change traveler, TSTF-490, Revision 0, 
``Deletion of E-bar Definition and Revision to RCS [Reactor Coolant 
System] Specific Activity Technical Specifications,'' with deviations.
    Date of issuance: November 25, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment No.: Unit 1-192; Unit 2-192; Unit 3-192.
    Renewed Facility Operating License Nos. NPF-41, NPF-51; and NPF-74: 
The

[[Page 74188]]

amendment revised the Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: March 4, 2013 (78 FR 
14128).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 25, 2013.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power 
Station, Unit 2, New London County, Connecticut

    Date of amendment request: April 3, 2013.
    Description of amendment request: The amendment would revise 
Technical Specification 3.9.16 ``Shielded Cask,'' due to changes to the 
minimum decay time for fuel assemblies adjacent to the spent fuel pool 
cask laydown area.
    Date of issuance: November 14, 2013.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 316.
    Renewed Facility Operating License No. DPR-65: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: June 11, 2013 (78 FR 
35062).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 14, 2013.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Units 1 and 2, Salem County, New Jersey

    Date of amendment requests: November 30, 2012, as supplemented by 
letter dated May 31, 2013.
    Brief description of amendments: The amendments approve a change to 
the site Emergency Plan to remove the backup plant vent extended range 
noble gas radiation monitoring (R45) indication, recording, and alarm 
capability in the emergency response facilities. Although the R45B/C 
monitor equipment skid will be removed, the licensee will maintain a 
capability in its Emergency Plan to take post-accident samples from the 
plant vent stack, as specified by an earlier commitment to Regulatory 
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power 
Plants to Assess Plant and Environs Conditions During and Following an 
Accident.''
    Date of issuance: November 27, 2013.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 305 and 287.
    Renewed Facility Operating License Nos. DPR-70 and DPR-75: The 
amendments revised the Facility Operating License and approved 
revisions to the Emergency Plan.
    Date of initial notice in Federal Register: May 14, 2013 (78 FR 
28252). The supplemental letter dated May 31, 2013, provided 
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 November 27, 2013.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 2nd day of December 2013.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-29168 Filed 12-9-13; 8:45 am]
BILLING CODE 7590-01-P