[Federal Register Volume 77, Number 132 (Tuesday, July 10, 2012)]
[Notices]
[Pages 40647-40657]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2012-16656]


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NUCLEAR REGULATORY COMMISSION

[NRC-2012-0161]


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 (Commission 
or the NRC) is publishing this regular biweekly notice. The Act 
requires the Commission 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 June 14 to June 27, 2012. The last biweekly 
notice was published on June 26, 2012 (77 FR 38094-38099).

ADDRESSES: You may access information and comment submissions related 
to this document, which the NRC possesses and are publically available, 
by searching on http://www.regulations.gov under Docket ID NRC-2012-
0161. You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0161. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected].
     Mail comments to: Cindy Bladey, Chief, Rules, 
Announcements, and Directives Branch (RADB), Office of Administration, 
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
     Fax comments to: RADB at 301-492-3446.

[[Page 40648]]

    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-2012-0161 when contacting the NRC 
about the availability of information regarding this document. You may 
access information related to this document by any of the following 
methods:
     Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0161.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in 
ADAMS by performing a search on the document date and docket number.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2012-0161 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information in comment submissions that you do not want to be publicly 
disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into 
ADAMS, and the NRC does not edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information in their comment submissions 
that they do not want to be publicly disclosed. Your request should 
state that the NRC will not edit comment submissions to remove such 
information before making the comment submissions available to the 
public or entering the comment submissions into ADAMS.
Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses and Combined Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing
    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Title 10 of the Code of Federal 
Regulations (10 CFR) 50.92, this means that operation of the facility 
in accordance with the proposed amendment would not (1) Involve a 
significant increase in the probability or consequences of an accident 
previously evaluated; or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ''Rules of 
Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested person(s) should consult a current copy of 10 CFR 2.309, 
which is available at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. The NRC regulations are accessible electronically from the NRC 
Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or a presiding 
officer designated by the Commission or by the Chief Administrative 
Judge of the Atomic Safety and Licensing Board Panel, will rule on the 
request and/or petition; and the Secretary or the Chief Administrative 
Judge of the Atomic Safety and Licensing Board will issue a notice of a 
hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of

[[Page 40649]]

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

[[Page 40650]]

the exemption from use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 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. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209, 
301-415-4737, or by email to [email protected].

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: February 22, 2012.
    Description of amendment request: The proposed amendments would 
allow the use of the nuclear service water system (NSWS) pump discharge 
crossover valves and associated piping to cross tie McGuire Nuclear 
Station, Units 1 and 2 (McGuire 1 and 2) NSWS trains to mitigate a Loss 
of Service Water (LOSW) event at McGuire 1 or 2.
    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.
    MNS' [McGuire Nuclear Station's] Final Safety Analysis Report 
(FSAR) conforms to the standard format and content of Revision 1 to 
Regulatory Guide (RG) 1.70 with exceptions described in the 
applicable sections of the FSAR. With regard to Chapter 15 
``Accident Analysis,'' MNS committed to analyzing the anticipated 
operational occurrences and postulated design basis accidents listed 
in Chapter 15 on pages 15T-1, 15T-2, and 15T-3 of RG 1.70 Revision 
1. MNS' FSAR Chapter 15 described an exception to a Loss of Service 
Water event (RG 1.70, Rev. 1, page 15T-3, item 30) and stated, in 
part, ``Loss of the Nuclear Service Water System is not considered a 
credible accident because of the redundancy provided in the 
system.'' The FSAR was later updated (UFSAR) to conform to Chapter 
15 accidents listed on pages 15-10, 15-11, and 15-12 of RG 1.70 
Revision 3. The initial FSAR Chapter 15 exception to RG 1.70 Rev. 1 
LOSW event was no longer required since LOSW events were no longer 
included in Chapter 15 of subsequent RG 1.70 revisions (revision 2 
or 3). Based on the licensing history, the LOSW event is not an 
anticipated operational occurrence or postulated design basis 
accident and was not previously analyzed in Chapter 15 of the UFSAR. 
A failure of the NSWS does not initiate any of the accidents 
previously evaluated in Chapter 15 of the UFSAR; therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Criterion 2:
    Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    McGuire [Nuclear Station] is a multi-unit site comprised of two 
nuclear stations, Unit 1 and Unit 2. Each unit has two NSWS trains 
and each train is designed to remove core decay heat following a 
design basis LOCA. Each train has a service water pump discharge 
crossover valve installed which allows the trains to be cross-
connected in any combination. The NSWS pump discharge crossover 
valves are described in the UFSAR as providing operational 
flexibility. Although designed to cross-connect unit NSWS trains, 
MNS has never licensed their use. The proposed change, consistent 
with the UFSAR description and [Generic Letter] GL 91-13, will 
provide the operational flexibility to allow one unit's NSWS to be 
aligned to another unit that has lost all service water.
    During normal operation, only one pump, per unit, is in 
operation to supply NSWS flow to the essential and non-essential 
headers for each unit. Cross-connecting NSWS between units will 
require a unit's standby NSWS pump to be placed in service 
(operating), opening its respective discharge crossover valve, and 
opening a LOSW unit's NSWS pump discharge crossover valve to 
establish service water flow to a LOSW unit's NSWS train. With 
exception to the flow path, the shared train is operated as 
designed. If the proposed [license amendment request] LAR is 
approved, the necessary site procedures will be revised to govern 
system operation and use of the crossover design feature to mitigate 
a LOSW event.
    The use of the NSWS pump discharge crossover valves within their 
design limitations and maintaining compliance to [technical 
specification] TS 3.7.7 [limiting condition for operation] LCO does 
not create any credible new failure mechanisms, malfunctions, or 
accident initiators that will prevent the ability of the NSWS to 
perform its design function. Operating the NSWS within the 
allowances of TS 3.7.7, which allow a train to be removed from 
service for up to 72 hours, does not impact the redundant 
capabilities afforded by the other train or the ``low probability of 
a design basis accident (DBA) occurring during this time period'' as 
stated in TS 3.7.7 Bases. Therefore, the proposed change does not 
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 a 
margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident situation. These barriers include 
the fuel cladding, the reactor coolant system, and the containment 
system. The performance of these barriers will not be impacted by 
the proposed change. The use of a NSWS pump discharge cross-over to 
cross-tie units is not a credited flow path in design basis and is 
not needed to perform the specified safety function. Cross-
connecting the units is an additional strategy made available if a 
total LOSW should occur.
    The proposed change will allow a unit to share a portion of an 
available service water train's capacity with a unit that has lost 
all service water. The shared alignment requires the use of service 
water pump discharge crossover valves which are not designated as 
shared components. Their use will improve the availability of 
service water and decreases the probability of core damage. 
Therefore the change will improve the margin of safety for each unit 
with respect to mitigating LOSW events.
    Placing a NSWS train in a shared alignment prevents the train 
from automatically performing its safety function and the train does 
not comply with GDC-5 [10 CFR Part 50, Appendix A, ``General Design 
Criteria for Nuclear Power Plants,'' Criterion 5, ``Sharing of 
structures, systems, and components''] and is declared inoperable. 
Limiting the time a train is inoperable to 72 hours manages the

[[Page 40651]]

vulnerability to single failure consistent with current TS required 
actions and completion times. In accordance with TS LCO 3.0.2 
allowances, TS 3.7.7 allows one train to be removed from service for 
up to 72 hours to perform surveillance testing, preventive 
maintenance, corrective maintenance, modifications, or investigation 
of operational problems. Although a NSWS train is declared 
inoperable for these activities, several can be accomplished while 
maintaining the train available while others, such as corrective 
maintenance, may also render the NSWS train unavailable. The 72 hour 
[completion time] CT is bounded by the worst case allowed by TS LCO 
3.0.2 which assumes a train is both inoperable and unavailable.
    Sharing a unit's redundant [nuclear service water] NSW pump 
requires the shared unit's service water pump to be taken out of 
standby and placed in service (operating). Therefore, the shared 
train remains available to the shared unit in event it must be 
restored. The shared train will be supplying the service water 
necessary to support operation of the shared unit's diesel generator 
(emergency power) and to assure long term operation of the shared 
pump. Although redundancy is lost in terms of performing its 
specified safety function on the designated unit, availability and 
functionality is maintained by the proposed amendment.
    The reason a redundant NSWS pump is inoperable and/or 
unavailable does not change the probability its redundant train will 
fail during the 72 hour CT or change the probability of a [loss-of-
coolant-accident] LOCA occurring during that time. In the event a 
train fails while its redundant train is shared, immediate action 
can be taken to restore the shared train from the shared alignment 
or the unit can be shutdown.
    Since a unit's redundant service water train is placed in a 
shared configuration to mitigate a LOSW event, margin of safety is 
considered on each unit. Technical Specifications allows a nuclear 
service water train to be removed from service for up to 72 hours. 
The shared unit's margin of safety is maintained by limiting the 
shared alignment to <72 hour completion time consistent with current 
TS allowances. Implementation of this amendment will improve the 
margin of safety on a unit experiencing a LOSW event consistent with 
the intent of NRC Generic Letter 91-13.
    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, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Nancy L. Salgado.

Duke Energy Carolinas, LLC, Docket No. 50-269, Oconee Nuclear Station, 
Unit 1 (ONS 1), Oconee County, South Carolina

    Date of amendment request: April 3, 2012.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to authorize a one-time, 15 
month extension to the integrated leak rate test (ILRT) of the reactor 
containment building (also known as the containment), which would align 
the test schedule with the refueling outage schedule. The ILRT is 
normally performed every 10 years. The upcoming ILRT is currently due 
by December 8, 2013.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed exemption involves a one-time extension to the 
current interval for ONS 1 Type A containment testing. The current 
test interval of 120 months (10 years) would be extended on a one-
time basis to no longer than approximately 135 months from the last 
Type A test. The proposed extension does not involve either a 
physical change to the plant or a change in the manner in which the 
plant is operated or controlled. The containment is designed to 
provide an essentially leak tight barrier against the uncontrolled 
release of radioactivity to the environment for postulated 
accidents. As such, the containment and the testing requirements 
invoked 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 the prevention or identification of 
any precursors of an accident. Therefore, this proposed extension 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    This proposed extension is for next ONS 1 Type A containment 
leak rate test only. The Type B and C containment leak rate tests 
would continue to be performed at the frequency currently required 
by the ONS 1 TS. As documented in NUREG 1493, Type B and C tests 
have identified a very large percentage of containment leakage 
paths, and the percentage of containment leakage paths that are 
detected only by Type A testing is very small. The ONS 1 Type A test 
history supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as (1) activity based and 
(2) time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with ASME [American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code] Section Xl, the Maintenance Rule, 
and TS requirements serve to provide a high degree of assurance that 
the containment would not degrade in a manner that is detectable 
only by a Type A test.
    Based on the above, the proposed extension does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Therefore, it is concluded that the proposed amendment does not 
significantly increase the consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment to the TS involves a one-time extension 
to the current interval for the ONS 1 Type A containment test. The 
containment and the testing requirements to periodically demonstrate 
the integrity of the containment exist to ensure the plant's ability 
to mitigate the consequences of an accident do not involve any 
accident precursors or initiators. The proposed change does not 
involve a physical change to the plant (i.e., no new or different 
type of equipment will be installed) or a change to the manner in 
which the plant is operated or controlled.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment to the TS involves a one-time extension 
to the current interval for the ONS 1 Type A containment test. This 
amendment does not alter the manner in which safety limits, limiting 
safety system set points, or limiting conditions for operation are 
determined. The specific requirements and conditions of the TS 
Containment Leak Rate Testing Program exist to ensure that the 
degree of containment structural integrity and leak-tightness that 
is considered in the plant safety analysis is maintained. The 
overall containment leak rate limit specified by TS is maintained.
    The proposed change involves only the extension of the interval 
between Type A containment leak rate tests for ONS 1. The proposed 
surveillance interval extension is bounded by the 15-month extension 
currently authorized within NEI 94-01, Revision 0. Type B and C 
containment leak rate tests would continue to be performed at the 
frequency currently required by TS. Industry experience supports the 
conclusion

[[Page 40652]]

that Type B and C testing detects a large percentage of containment 
leakage paths and that the percentage of containment leakage paths 
that are detected only by Type A testing is small. The containment 
inspections performed in accordance with ASME Section XI, TS and the 
Maintenance Rule serve to provide a high degree of assurance that 
the containment would not degrade in a manner that is detectable 
only by Type A testing. The combination of these factors ensures 
that the margin of safety in the plant safety analysis is 
maintained. The design, operation, testing methods and acceptance 
criteria for Type A, B, and C containment leakage tests specified in 
applicable codes and standards would continue to be met, with the 
acceptance of this proposed change, since these are not affected by 
changes to the Type A test interval.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    Based on the NRC staff's 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: Nancy L. Salad.

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina

    Date of amendment request: December 16, 2011, as supplemented by 
letters dated January 20, March 1, March 16, and April 18, 2012.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications and the Updated Final Safety 
Analysis Report to add the new Protected Service Water (PSW) System to 
the plant's licensing basis as an additional method of achieving and 
maintaining safe shutdown of the reactors in the event of a high-energy 
line break or a fire in the turbine building, which is shared by all 
three units.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided 
its analysis of the issue of no significant hazards consideration. The 
Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's 
analysis against the standards of 10 CFR 50.92(c). The NRC staff's 
analysis of the no significant hazards consideration is presented 
below:

    Criterion 1:
    Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The changes proposed include the construction of a new PSW 
building, which will have the equipment to receive electrical power 
from two independent sources and provide electrical power to 
important equipment located in the auxiliary building or the reactor 
containment building without being routed through the turbine 
building. Since certain high-energy line breaks (HELBs) or fires in 
the turbine building could adversely affect the power supplies to 
equipment needed to maintain the reactors in safe shutdown, the PSW 
System provides added assurances that safe shutdown can be achieved 
and maintained. The PSW system does not have any failure modes that 
would initiate the type of accidents previously evaluated, so there 
will be no increase in the probability of an accident previously 
evaluated. The PSW System modifications will be designed and 
installed in accordance with applicable quality standards such that 
there will be no significant increase in the probability of failure 
or malfunction of existing structures, systems, or components (SSCs) 
used to mitigate accidents. Since there will be no significant 
increase in the probability of malfunction of these SSCs, there also 
will be no significant increase in the consequences of accidents 
previously evaluated.
    Criterion 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 modifications are designed to enhance the station's 
ability to achieve safe shutdown following a HELB or fire in the 
turbine building. As the new equipment will be designed and 
installed in accordance with applicable quality standards, there is 
reasonable assurance that it will not introduce new malfunctions or 
accident initiators different from the accidents that are already 
evaluated.
    Criterion 3:
    Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The addition of the PSW system improves the station's overall 
risk margin, therefore this change does not involve a significant 
reduction in a margin of safety.

    Based on the NRC staff's 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: Nancy L. Salgado.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
2, Pope County, Arkansas

    Date of amendment request: April 4, 2012.
    Description of amendment request: The proposed amendment addresses 
the Arkansas Nuclear One, Unit No. 2 (ANO-2) revised fuel handling 
accident (FHA) based on the U.S. Nuclear Regulatory Commission (NRC) 
staff approved license amendment request regarding use of Alternate 
Source Terms (AST) (NRC safety evaluation dated April 26, 2011 
(Agencywide Documents Access and Management System (ADAMS) Accession 
No. ML110980197)). As presented in the licensee's letter dated March 
31, 2010 (ADAMS Accession No. ML100910241), the original FHA analysis 
assumed failure of 60 fuel rods in a single fuel assembly. The revised 
analysis assumes the failure of all fuel rods in two fuel assemblies 
(472 rods). The revised analysis was provided in the licensee's letter 
dated June 23, 2010 (ADAMS Accession No. ML102000199).
    The changes necessary to support the revised FHA affect similar 
Technical Specifications (TSs) associated with NRC-approved Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Travelers TSTF-51, Revision 2, ``Revise Containment Requirements During 
Handling Irradiated Fuel and Core Alterations''; TSTF-272, Revision 1, 
``Refueling Boron Concentration Clarification''; TSTF-268, Revision 2, 
``Operations Involving Positive Reactivity Additions''; and TSTF-471, 
Revision 1, ``Eliminate use of Term Core Alterations in Actions and 
Notes.'' Therefore, the licensee proposes to adopt these TSTFs in 
conjunction with changes necessary to support the revised FHA. 
Additionally, administrative and/or editorial errors noted during the 
review are also corrected (in relation to the TS pages affected by the 
aforementioned proposed changes).
    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. Each of the five items described above is addressed 
individually under each of the three standards, as presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Revised FHA
    Response: No.
    TS changes associated with the FHA analysis ensure the initial 
assumptions of the FHA are maintained and, therefore, act to 
minimize the consequences of an accident by ensuring TS required 
features are operable during the movement of fuel assemblies. The 
FHA analysis was recently accepted by the NRC during adoption of 
Alternate Source

[[Page 40653]]

Terms for ANO-2. The probability of a fuel assembly drop (or any 
load drop) is unchanged by the revised analysis. Therefore, the 
revised FHA does not involve a significant increase in the 
probability of an accident previously evaluated.
    The FHA analysis was recently accepted by the NRC during 
adoption of Alternate Source Terms for ANO-2. In addition, 
Licensee's has reviewed station procedures and controls in order to 
verify that no other loads, other than a new or irradiated fuel 
assembly, need be addressed with regard to a FHA (i.e., no other 
known load carried over irradiated fuel assemblies exists which 
would be expected to cause fuel damage if dropped). The proposed TS 
changes simply ensure required systems will be operable during 
operations that could lead to an FHA. Based on the above, the 
proposed FHA-related changes to the TSs do not result in a 
significant increase in the consequences of an accident previously 
evaluated.

TSTF-51 and TSTF 471

    Response: No.
    The only design basis accident assumed for ANO-2 related to the 
proposed changes is the FHA. The boron dilution event is evaluated, 
but considered an unlikely event due to the time available for 
operator response and the administrative controls that permit early 
detection of the event. The loss of SDC [shutdown cooling] event has 
little relationship and minimal impact with regard to a FHA. TSTF-51 
and TSTF-471 simply replace the use of the previously defined ``core 
alterations'' term with requirements associated with the movement of 
fuel assemblies, since the drop of a fuel assembly is the only event 
that could reasonably lead to an FHA or a significant challenge to 
the plant.
    The removal of all references to ``core alterations'' in favor 
of restrictions associated with the movement of fuel assemblies 
eliminates current restrictions associated with the manipulation of 
other core components (i.e., sources or reactivity control 
components within the core) since such manipulation cannot result in 
an FHA, boron dilution event, or loss of SDC. In addition, 
manipulation of these other components cannot present a significant 
challenge to SDM [shutdown margin] because the TS required RCS 
[reactor coolant system] boron concentration for Mode 6 operation 
provides substantial margin to criticality.
    Changes associated with TSTF-51 and TSTF-471 do not modify 
limitations in such a way that the consequences of an FHA would be 
greater than that assumed in the FHA analysis (i.e., 10 CFR 50.67 
and General Design Criterion (GDC) 19 limitations are not exceeded 
following a FHA)).
    Based on the above, the proposed changes associated with the 
adoption of TSTF-51 and TSTF-471 do not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.

TSTF-272

    Response: No.
    Changes associated with TSTF-272 simply place additional 
restrictions on Mode 6 operations by ensuring the boron 
concentration of the water in the refueling canal meets the same TS 
limits required for the RCS when the RCS is in direct hydraulic 
communication with the refueling canal (i.e., reactor vessel head 
removed and refueling canal filled). These changes are unrelated to 
any accident initiator and further prohibit any challenge to the 
fuel in the reactor vessel by ensure sufficient boron concentration 
is maintained during Mode 6 operations. Therefore, these changes do 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.

TSTF-286

    Response: No.
    Changes associated with TSTF-286 permit operator control of RCS 
inventory and temperature when certain TS requirements are not met, 
provide the overall required SDM of the RCS is maintained. The 
activities that involve inventory makeup from sources with boron 
concentrations less than the current RCS concentration (i.e., boron 
dilution) need not be precluded in the TSs provided the required SDM 
is maintained for the worst-case overall effect on the core. Note 
that an unexpected boron dilution event is considered unlikely for 
ANO-2 due to the significant period of time for operator detection 
and response before SDM would be significantly challenged (reference 
ANO-2 SAR Section 15.1.4.3). In addition, while a boron dilution 
event is evaluated in the safety analysis, the only ``accident'' 
assumed for ANO-2 during Mode 6 operations is the FHA. Permitting 
RCS inventory and temperature adjustments is unrelated to any 
assumptions associated with a FHA. Therefore, these changes do not 
result in a significant increase in the probability an accident (or 
a boron dilution event) previously evaluated. Because an unexpected 
boron dilution event provides sufficient opportunity for detection 
and recovery, the proposed changes associated with TSTF-286 likewise 
do not result in a significant increase in the consequences of an 
accident (or boron dilution event) previously evaluated.

Enhancements and Administrative Changes

    Response: No.
    Enhancements and administrative changes proposed for 
specifications affected by the above revised FHA or TSTF adoptions 
are unrelated to any accident initiator. Administrative changes 
likewise cannot impact the consequences of any accident previously 
evaluated.
    Enhancements associated with the Containment Purge system 
radiation instrumentation ensure Surveillance testing is performed 
when the system is in service, regardless if an actual Purge is 
taking place. In addition, the proposed changes ensure appropriate 
testing is performed prior to placing the system in service each 
refueling outage. The proposed changes are neutral or more 
restrictive and, therefore, cannot increase the consequences of an 
accident previously evaluated.
    Based on the above, the proposed changes do not represent 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?
    Revised FHA
    Response: No.
    TS changes associated with the revised FHA involve no physical 
changes to the plant. These changes act to ensure required SSCs are 
operable when moving irradiated fuel assemblies or new fuel 
assemblies over irradiated fuel assemblies to limit any Control Room 
or offsite dose consequences to within acceptable limits. Therefore, 
these changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

TSTF-51 and TSTF-471

    Response: No.
    TS changes associated with the adoption of these TSTFs involve 
no physical changes to the plant. The removal of all references to 
``core alterations'' in favor of restrictions associated with the 
movement of fuel assemblies eliminates current restrictions 
associated with the manipulation of other core components (i.e., 
sources or reactivity control components within the core). Such 
manipulations cannot result in an FHA, boron dilution event, or loss 
of SDC. In addition, such manipulations cannot result in an 
appreciable change in core reactivity due to the high RCS boron 
concentration required during refueling operations by the TSs. The 
proposed changes do not introduce a new accident initiator, accident 
precursor, or accident-related malfunction mechanism.
    Therefore, these changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

TSTF-272

    Response: No.
    Changes associated with TSTF-272 place additional restrictions 
on Mode 6 operations by ensuring the boron concentration of the 
water in the refueling canal meets the same TS limits required for 
the RCS when the RCS is in direct hydraulic communication with the 
refueling canal (i.e., reactor vessel head removed and refueling 
canal filled). These changes are unrelated to any accident initiator 
and further prohibit any challenge to the fuel in the reactor vessel 
by ensure sufficient boron concentration is maintained during Mode 6 
operations. The proposed changes do not introduce a new accident 
initiator, accident precursor, or accident-related malfunction 
mechanism. Therefore, these changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

TSTF-286

    Response: No.
    Changes associated with TSTF-286 permit operator control of RCS 
inventory and temperature when certain TS requirements are not met, 
provide the overall required SDM of the RCS is maintained. No 
physical plant changes are related to these TS changes. The only 
accident or event that could be affected by this change is the boron 
dilution event, which has been previously evaluated. The proposed 
changes do not introduce a new accident initiator, accident 
precursor, or accident-related malfunction

[[Page 40654]]

mechanism. Therefore, these changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

Enhancements and Administrative Changes

    Response: No.
    Enhancements and administrative changes proposed for 
specifications affected by the above revised FHA or TSTF adoptions 
are unrelated to any accident initiator and involve no physical 
changes to the plant.
    Enhancements associated with the Containment Purge system 
radiation instrumentation ensure Surveillance testing is performed 
when the system is in service, regardless if an actual Purge is 
taking place. In addition, the proposed changes ensure appropriate 
testing is performed prior to placing the system in service each 
refueling outage.
    The proposed changes do not introduce a new accident initiator, 
accident precursor, or accident-related malfunction mechanism. Based 
on the above, these changes do 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?

Revised FHA

    Response: No.
    TS changes associated with the revised FHA act to ensure 
required SSCs [structures, systems, and components] are operable 
when moving irradiated fuel assemblies or new fuel assemblies over 
irradiated fuel assemblies to limit any Control Room or offsite dose 
consequences to within acceptable limits. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.

TSTF-51 and TSTF-471

    Response: No.
    The removal of all references to ``core alterations'' in favor 
of restrictions associated with the movement of fuel assemblies 
eliminates current restrictions associated with the manipulation of 
other core components (i.e., sources or reactivity control 
components within the core). Such manipulations cannot result in an 
FHA, boron dilution event, or loss of SDC. In addition, such 
manipulations cannot result in an appreciable change in core 
reactivity due to the high RCS boron concentration required during 
refueling operations by the TSs. Changes associated with TSTF-51 and 
TSTF-471 do not modify limitations in such a way that the 
consequences of an FHA would be greater than that assumed in the FHA 
analysis (i.e., 10 CFR 50.67 and GDC 19 limitations are not exceeded 
following a FHA). Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

TSTF-272

    Response: No.
    Changes associated with TSTF-272 place additional restrictions 
on Mode 6 operations by ensuring the boron concentration of the 
water in the refueling canal meets the same TS limits required for 
the RCS when the RCS is in direct hydraulic communication with the 
refueling canal (i.e., reactor vessel head removed and refueling 
canal filled). These changes are more restrictive than the current 
specification and therefore do not involve a significant reduction 
in a margin of safety.

TSTF-286

    Response: No.
    Changes associated with TSTF-286 permit operator control of RCS 
inventory and temperature when certain TS requirements are not met, 
provide the overall required SDM of the RCS is maintained. The only 
accident or event that could be affected by this change is the boron 
dilution event, which has been previously evaluated. While the 
margin between existing boron concentration and that required to 
meet SDM requirements may be reduced, margin is gained by permitting 
operators to take corrective action to maintain RCS inventory and 
temperature within limits during periods when such operations are 
otherwise prohibited. While not quantifiable, the changes associated 
with TSTF-286 have a general balanced effect in relation to the 
margin of safety. Because an unexpected boron dilution event 
provides sufficient opportunity for detection and recovery, the 
proposed changes associated with TSTF-286 do not involve a 
significant reduction in a margin of safety.

Enhancements and Administrative Changes

    Response: No.
    Enhancements and administrative changes proposed for 
specifications affected by the above revised FHA or TSTF adoptions 
are unrelated to any accident initiator or mitigation strategy. 
Enhancements associated with the Containment Purge system radiation 
instrumentation ensure Surveillance testing is performed when the 
system is in service, regardless if an actual Purge is taking place. 
In addition, the proposed changes ensure appropriate testing is 
performed prior to placing the system in service each refueling 
outage. Based on the above, these 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: Joseph A. Aluise, Associate General 
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Michael T. Markley.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center (DAEC), Linn County, Iowa

    Date of amendment request: May 1, 2012.
    Description of amendment request: The proposed amendment would 
revise the Duane Arnold Energy Center (DAEC) Technical Specifications 
(TS) on a one-time basis by adding a note to TS Table 3.3.5.1-1, 
Function 1d, Modes 4 and 5, specifying that Function 1d is not required 
to be met during Refueling Outage (RFO) 23 in Modes 4 and 5.
    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 any accident previously 
evaluated?
    Response: No.
    The proposed amendment would revise the DAEC TS on a one-time 
basis by adding a note to TS Table 3.3.5.1-1, Function 1d, Modes 4 
and 5, specifying that Function 1d is not required to be met during 
RFO 23 in Modes 4 and 5. Accidents are initiated by the malfunction 
of plant equipment, or the catastrophic failure of plant structures, 
systems, or components.
    The low pressure Emergency Core Cooling System (ECCS) subsystems 
are designed to inject to reflood or to spray the core after any 
size break up to and including a design basis Loss of Coolant 
Accident (LOCA). The proposed change to the Core Spray System 
Operability requirements does not change the operating 
configurations or minimum amount of operating equipment assumed in 
the safety analysis for accident mitigation. The change does not 
require any change in safety analysis methods or results. Also, it 
does not change the amount of core spray provided to the core in the 
accident analyses. No changes are proposed to the manner in which 
the ECCS provides plant protection or which would create new modes 
of plant operation. The proposed change does not result in any new 
or affect the probability of any accident initiators. There will be 
no degradation in the performance of, or an increase in the number 
of challenges imposed on, safety related equipment assumed to 
function during an accident situation. There will be no change to 
normal plant operating parameters or accident mitigation 
performance. This change will only apply when the plant is in MODES 
4 and 5 where LOCAs are not postulated to occur. In MODES 4 and 5, 
the CS function is to mitigate OPDRVs [Operations with the Potential 
for Draining the Reactor Vessel].
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    This change does not affect the method by which any plant 
systems perform a safety function. It does not introduce any new 
equipment, or hardware changes, which could create a new or 
different kind of accident. No new release pathways or equipment 
failure modes are created. No new

[[Page 40655]]

accident scenarios failure mechanisms or limiting single failures 
are introduced as a result of this request. This request does not 
affect the normal methods of plant operation. The Core Spray System 
retains its ability to function following any accident previously 
evaluated and provide the proper flow rate to the core. This change 
will only apply when the plant is in MODES 4 and 5 where LOCAs are 
not postulated to occur. In MODES 4 and 5, the CS function is to 
mitigate OPDRVs. Strict administrative and procedural controls, 
operator training, and use of human performance tools will be 
essential to preventing these types of consequential human errors. 
Furthermore, both CS subsystems will be guarded and no work or 
testing will be permitted on either of the CS subsystems during RFO 
23 when both CS subsystems are needed to be Operable to meet the 
requirements of LCO 3.5.2.
    Therefore, the implementation of the proposed change will not 
create a possibility for an accident of a new or different type than 
those previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The ECCS are designed with sufficient redundancy such that if a 
Core Spray subsystem were unavailable, or did not provide the 
required flowrate, the remaining Core Spray subsystem is capable of 
providing water and removing heat loads to satisfy the Updated Final 
Safety Analysis Report requirements for accident mitigation. A 
minimum of two low pressure ECCS subsystems continue to be required 
to be OPERABLE in MODES 4 and 5, except with the spent fuel storage 
pool gates removed and water level >= 21 ft 1 inch over the top of 
the reactor pressure vessel flange. There is no change in the 
Limiting Conditions for Operation. For these reasons, the proposed 
amendment does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Mitchell S. Ross, P.O. Box 14000 Juno 
Beach, FL 33408-0420.
    NRC Acting Branch Chief: Istvan Frankl.

Previously Published Notices 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 following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of amendment request: March 22, 2012.
    Brief description of amendment request: The proposed amendments 
would revise the technical specification for the Vogtle Electric 
Generating Plant, Units 1 and 2, associated with the ``Steam Generator 
(SG) Program'' allowing the exclusion of portions of the SG tubes below 
the top of the tube sheet from periodic SG tube inspections during the 
remaining licensed operations of the plant. Furthermore, the amendment 
requests to remove the interim SG alternative inspection criteria that 
had been previously approved.
    Date of publication of individual notice in Federal Register: May 
25, 2012 (77 FR 31402).
    Expiration date of individual notice: July 24, 2012.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the NRC's Public Document Room (PDR), located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. Publicly available documents created or received at the NRC are 
accessible electronically through the Agencywide Documents Access and 
Management System (ADAMS) in the NRC Library at http://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 
[email protected].

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: November 22, 2011, as 
supplemented by letter dated May 11, 2012.
    Brief description of amendment: The amendments remove duplicate 
Technical Specification (TS) requirements and unit-specific references 
that are no longer needed. In addition, the administrative changes 
correct typographical errors and provide clarification to ensure 
understanding of the required actions of some of the TSs. The changes 
include corrective actions from the Unit 2 event described in Licensee 
Event Report (LER) 50-529/2011-001. The changes are administrative or 
editorial in nature, and would not result in any change to operating 
requirements. These administrative changes are for TS 3.3.1, ``Reactor 
Protective System (RPS) Instrumentation--Operating''; TS 3.3.2, 
``Reactor Protective System (RPS) Instrumentation--Shutdown''; TS 
3.3.5, ``Engineered Safety Features Actuation System (ESFAS) 
Instrumentation''; TS 3.5.5, ``Refueling Water Tank (RWT)''; TS 3.3.9, 
``Control Room Essential

[[Page 40656]]

Filtration Actuation Signal (CREFAS)''; TS 3.7.11, ``Control Room 
Essential Filtration System (CREFS)''; TS 5.4, ``Procedures''; and TS 
5.5.16, ``Containment Leakage Rate Testing Program.''
    Date of issuance: June 18, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: Unit 1--189; Unit 2--189; Unit 3--189.
    Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
The amendment revised the Operating Licenses and Technical 
Specifications.
    Date of initial notice in Federal Register: January 24, 2012 (77 FR 
3510). The supplemental letter dated May 11, 2012, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register on January 24, 2012 (77 FR 3510).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 18, 2012.
    No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2 (Catawba 1 and 2), York County, 
South Carolina; Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-
370, McGuire Nuclear Station, Units 1 and 2 (McGuire 1 and 2), 
Mecklenburg County, North Carolina; Duke Energy Carolinas, LLC, Docket 
Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, 
and 3 (Oconee 1, 2, and 3), Oconee County, South Carolina

    Date of application for amendments: December 15, 2009, as 
supplemented by letter dated September 22, 2011.
    Brief description of amendments: The amendments consist of changes 
to the Technical Specifications (TSs) associated with Reactor Coolant 
System (RCS) Specific Activity and the deletion of the TS definition of 
E Bar (average disintegration energy) consistent with Revision 0 to TS 
Task Force (TSTF) Standard Technical Specification Change Document 
TSTF-490, ``Deletion of E Bar Definition and Revision to RCS Specific 
Activity Tech Spec.''
    Date of issuance: June 25, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: Catawba: Unit 1--268 and Unit 2--264; McGuire: Unit 
1--266 and Unit 2--246; Oconee: Unit 1--380, Unit 2--382, and Unit 3--
381.
    Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, NPF-
17, DPR-38, DPR-47, and DPR-55: Amendments revised the licenses.
    Date of initial notice in Federal Register: March 23, 2010 (75 FR 
13789). The September 22, 2011, supplement did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 25, 2012.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of application for amendments: October 21 and December 14, 
2010, as supplemented by letters dated December 21, 2010, January 7, 
2011, January 28, February 22, March 3, March 9 (two letters), March 16 
(two letters), March 23, March 25, March 31 (two letters), April 14 
(two letters), April 22 (2 letters), April 26, April 28 (2 letters), 
April 29, May 11, May 18, May 19 (two letters), May 26 (two letters), 
June 7, June 9, June 21 (two letters), July 7 (two letters), July 22, 
July 29, August 5, August 11, August 16 (two letters), August 19, 
August 25 (two letters), August 29, September 14, September 16, 
September 30 (two letters), October 6, October 12 (two letters), 
October 14, October 15, November 9, December 22 (2 letters), December 
31, 2011, January 10, 2012, January 16 (two letters), January 17, 
January 19, January 23 (two letters), January 25, January 31, February 
3, February 15, February 23 (two letters), and March 15, 2012.
    Brief description of amendments: The proposed amendments would 
increase the licensed core power level for Turkey Point, Units 3 and 4 
from 2300 megawatts thermal (MWt) to 2644 MWt. This represents a net 
increase in the core thermal power of approximately 15 percent, 
including a 13-percent power uprate and a 1.7 percent measurement 
uncertainty recapture, over the current licensed thermal power level 
and is defined as an extended power uprate. The proposed amendments 
would change the renewed facility operating licenses, the technical 
specifications (TSs) and licensing bases to support operation at the 
increased core thermal power level, including changes to the maximum 
licensed reactor core thermal power, reactor core safety limits, 
reactor protection system and engineered safety feature actuation 
system limiting safety system settings, and emergency diesel generator 
surveillance start voltage and frequency. Additional TS changes include 
reactor coolant system heatup and cooldown limitations, pressurizer 
safety valve settings, accumulator and refueling water storage tank 
boron concentrations, main steam safety valve maximum allowable power 
level and lift settings, new main feedwater isolation valves, and core 
operating limits report references. A complete list of the proposed TS 
changes and the licensee's basis for change can be found in Attachment 
1 of the licensee's application (Agencywide Documents and Management 
System Accession No. ML103560167).
    Date of issuance: June 15, 2012.
    Effective date: Unit 3--This license amendment is effective as of 
its date of issuance and shall be implemented prior to Unit 3 startup 
from the spring 2012 refueling outage. Unit 4--This license amendment 
is effective as of its date of issuance and shall be implemented prior 
to Unit 4 startup from the fall 2012 refueling outage.
    Amendment Nos.: Unit 3--249 and Unit 4--245.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the License and Technical Specifications.
    Date of initial notice in Federal Register: May 9, 2011 (76 FR 
26771). The supplemental letters provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the 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 June 15, 2012.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of application for amendments: August 17, 2011, as 
supplemented by letters dated October 14, and December 1, 2011.
    Brief description of amendments: The amendments revised items in 
Technical Specification (TS) 3.3.3.3, Table 3.3-5, Accident Monitoring 
Instrumentation, High Range-Noble Gas Effluent Monitors, Main Steam 
Lines, Instrument 19d, and TS 4.3.3.3, Table 4.3-4 related

[[Page 40657]]

to the need to have High Range-Noble Gas Effluent Monitors for the Main 
Steam Lines. The changes relocated the TSs and surveillance 
requirements for this instrument to the Updated Final Safety Analysis 
Report and related procedures.
    Date of issuance: June 15, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 3--250 and Unit 4--246.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the TSs and Surveillance Requirements.
    Date of initial notice in Federal Register. October 18, 2011 (76 FR 
64393). The supplements dated October 14 and December 1, 2011, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 15, 2012.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of application for amendments: May 25, 2011.
    Brief description of amendments: The amendments relocate Technical 
Specifications (TSs) in Section 5.2--``Containment,'' Section 5.4--
``Reactor Coolant System,'' and Section 5.6--``Component Cyclic or 
Transient Limit,'' to the Updated Final Safety Analysis Report. TS 
5.3.3 regarding spent fuel storage pool capacity would be revised to a 
total pool capacity limit only.
    Date of issuance: June 21, 2012.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: Unit 3-251 and Unit 4-247.
    Renewed Facility Operating License Nos. DPR-31 and DPR-41: 
Amendments revised the TSs.
    Date of initial notice in Federal Register: October 18, 2011 (76 FR 
64392).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 21, 2012.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 19, 2012.
    Brief description of amendment: The NRC issued Amendment No. 239, 
Departure from a Method of Evaluation for the Auxiliary Building 
Overhead Crane (FHCR-5), on December 27, 2011. Amendment No. 239 was 
approved to be implemented within 180 days of issuance of the 
amendment. By letter dated March 19, 2012, the licensee requested 
extending the implementation period for Amendment 239 to allow for 
installation and testing of the new single failure proof FHCR-5. This 
amendment approved additional time to complete the implementation of 
Amendment No. 239 from 180 days to, ``Implementation shall be completed 
90 days prior to moving a spent fuel shipping cask with FHCR-5.''
    Date of issuance: June 26, 2012.
    Effective date: As of the date of issuance.
    Amendment No.: 241.
    Facility Operating License No. DPR-72: Amendment approved a 
revision to the Amendment No. 239 implementation schedule.
    Date of initial notice in Federal Register: April 17, 2012 (77 FR 
22814).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 26, 2012.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 29th day of June 2012.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2012-16656 Filed 7-9-12; 8:45 am]
BILLING CODE 7590-01-P