[Federal Register Volume 80, Number 41 (Tuesday, March 3, 2015)]
[Notices]
[Pages 11472-11492]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-04298]


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NUCLEAR REGULATORY COMMISSION

[NRC-2015-0041]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of 
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission 
(NRC) is publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from February 5, 2015 to February 18, 2015. The 
last biweekly notice was published on February 17, 2015.

DATES: Comments must be filed by April 2, 2015. A request for a hearing 
must be filed by May 4, 2015.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
    <bullet> Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0041. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: <a href="/cdn-cgi/l/email-protection#abe8cad9c4c785eccac7c7caccc3ced9ebc5d9c885ccc4dd"><span class="__cf_email__" data-cfemail="f9ba988b9695d7be989595989e919c8bb9978b9ad79e968f">[email protected]</span></a>.
    <bullet> Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-5411, email: <a href="/cdn-cgi/l/email-protection#40132829322c25396e122f28322532002e32236e272f36"><span class="__cf_email__" data-cfemail="aefdc6c7dcc2cbd780fcc1c6dccbdceec0dccd80c9c1d8">[email protected]</span></a>.

[[Page 11473]]


SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2015-0041 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
    <bullet> Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0041.
    <bullet> NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to <a href="/cdn-cgi/l/email-protection#106074623e6275637f65627375507e62733e777f66"><span class="__cf_email__" data-cfemail="5d2d392f732f382e32282f3e381d332f3e733a322b">[email protected]</span></a>. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section.
    <bullet> 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-2015-0041, facility name, unit 
number(s), application date, and subject in your comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter 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 into ADAMS.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of Title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

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

    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC's regulations are accessible electronically from the NRC Library on 
the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed by the above date, the Commission or a presiding officer 
designated by the Commission or by the Chief Administrative Judge of 
the Atomic Safety and Licensing Board Panel, will rule on the request 
and/or petition; and the Secretary or the Chief Administrative Judge of 
the Atomic Safety and Licensing Board will issue a notice of a hearing 
or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends

[[Page 11474]]

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 unless the 
Commission finds an imminent danger to the health or safety of the 
public, in which case it will issue an appropriate order or rule under 
10 CFR part 2.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten 10 days prior to the filing deadline, the participant should 
contact the Office of the Secretary by email at <a href="/cdn-cgi/l/email-protection#a9c1ccc8dbc0c7ce87cdc6cac2ccdde9c7dbca87cec6df"><span class="__cf_email__" data-cfemail="e9818c889b80878ec78d868a828c9da9879b8ac78e869f">[email protected]</span></a>, 
or by telephone at 301-415-1677, to request (1) a digital 
identification (ID) certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at http://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to 
<a href="/cdn-cgi/l/email-protection#b3fee0fbf79de1d6c0dcc6c1d0d6f3ddc1d09dd4dcc5"><span class="__cf_email__" data-cfemail="dd908e9599f38fb8aeb2a8afbeb89db3afbef3bab2ab">[email protected]</span></a>, 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

[[Page 11475]]

or party to use E-Filing if the presiding officer subsequently 
determines that the reason for granting the exemption from use of E-
Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2 (ANO-2), Pope County, Arkansas
    Date of amendment request: February 6, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15041A068.
    Description of amendment request: The amendment would revise a Note 
to Technical Specification (TS) Surveillance Requirement (SR) 4.1.3.1.2 
to exclude Control Element Assembly (CEA) 18 from being exercised per 
the SR for the remainder of Cycle 24 due to a degrading upper gripper 
coil.
    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.
    One function of the CEAs is to provide a means of rapid negative 
reactivity addition into the core. This occurs upon receipt of a 
signal from the Reactor Protection System. This function will 
continue to be accomplished with the approval of the proposed 
change. Typically, once per 92 days each CEA is moved at least five 
inches to ensure the CEA is free to move. CEA 18 remains trippable 
(free to move) as illustrated by the last performance of SR 
4.1.3.1.2 in January 2015. However, due to abnormally high coil 
voltage and current measured on the CEA 18 Upper Gripper Coil (UGC), 
future exercising of the CEA could result in the CEA inadvertently 
inserting into the core, if the UGC were to fail during the exercise 
test. The mis-operation of a CEA, which includes a CEA drop event, 
is an abnormal occurrence and has been previously evaluated as part 
of the ANO-2 accident analysis. Inadvertent CEA insertion will 
result in a reactivity transient and power reduction, and could lead 
to a reactor shutdown if the CEA is deemed to be unrecoverable. The 
proposed change would minimize the potential for inadvertent 
insertion of CEA 18 into the core by maintaining the CEA in place 
using the Lower Gripper Coil (LGC), which is operating normally. The 
proposed change will not affect the CEAs ability to insert fully 
into the core upon receipt of a reactor trip signal.
    No modifications are proposed to the Reactor Protection System 
or associated Control Element Drive Mechanism Control System logic 
with regard to the ability of CEA 18 to remain available for 
immediate insertion. The accident mitigation features of the plant 
are not affected by the proposed amendment. Because CEA 18 remains 
trippable, no additional reactivity considerations need to be taken 
into consideration. Nevertheless, Entergy has evaluated the 
reactivity consequences associated with failure of CEA 18 to insert 
upon a reactor trip in accordance with TS requirements for Shutdown 
Margin (SDM) and has determined that SDM requirements would be met 
should such an event occur at any time during the remainder of Cycle 
24 operation.
    Therefore, 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.
    CEA 18 remains trippable. The proposed change will not introduce 
any new design changes or systems that can prevent the CEA from 
[performing] its specified safety function. As discussed previously, 
CEA mis-operation has been previously evaluated in the ANO-2 
accident analysis. Furthermore, SDM has been shown to remain within 
limits should an event occur at any time during the remainder of 
operating Cycle 24 such that CEA 18 fails to insert into the core 
upon receipt of a reactor trip signal.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from an accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    SR 4.1.3.1.2 is intended to verify CEAs are free to move (i.e., 
not mechanically bound). The physical and electrical design of the 
CEAs, and past operating experience, provides high confidence that 
CEAs remain trippable whether or not exercised during each SR 
interval. Eliminating further exercising of CEA 18 for the remainder 
of Cycle 24 operation does not directly relate to the potential for 
CEA binding to occur. No mechanical binding has been previously 
experienced at ANO-2. CEA 18 is contained within a Shutdown CEA 
Group and is not used for reactivity control during power maneuvers 
(the CEA must remain fully withdrawn at all times when the reactor 
is critical). In addition, Entergy has concluded that required SDM 
will be maintained should CEA 18 fail to insert following a reactor 
trip at any point during the remainder of Cycle 24 operation.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Acting Branch Chief: Eric R. Oesterle.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana
    Date of amendment request: October 1, 2014, as supplemented by 
letter dated February 2, 2015. Publicly-available versions are in ADAMS 
under Accession Nos. ML14275A374 and ML15033A482.
    Description of amendment request: The amendment would relocate 
Technical Specifications 3.9.6, ``Refuel Machine,'' and 3.9.7, ``Crane 
Travel,'' to the Technical Requirements Manual.
    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

[[Page 11476]]

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.
    This proposed change relocates Technical Specifications (TS) 3.9.6 
(Refuel Machine) and TS 3.9.7 (Crane Travel) to the Waterford 3 
Technical Requirements Manual (TRM). This is consistent with the 
requirements of [10 CFR 50.36(c)(2)(ii)] and aligns with NUREG-1432 
(Combustion Engineering Standard Technical Specifications).
    The applicable TS 3.9.6 and TS 3.9.7 design basis accident is the 
Fuel Handling Accident (FHA) described in the Updated Final Safety 
Analysis Report (UFSAR) Section 15.7.3.4. The limiting FHA results in 
all the fuel pins in the dropped and impacted fuel assemblies failing 
(472 pins or 236 per assembly). The analysis assumes that a fuel 
assembly is dropped as an initial condition and no equipment or 
intervention can prevent the initiating condition. The proposed change 
was evaluated against [10 CFR 50.36(c)(2)(ii)] criteria and shows no 
impact to the lowest functional capability or performance levels of 
equipment required for safe operation of the facility because the TS 
3.9.6 and TS 3.9.7 requirements do not prevent the accident conditions 
from occurring and do not limit the severity of the accident. Since, 
the dropped fuel assembly and the impacted fuel assembly are both 
already failed in the design basis accident scenario, this change could 
not result in a significant increase in the accident consequences. The 
TS 3.9.6 and TS 3.9.7 equipment are not required to respond, mitigate, 
or terminate any design basis accident, thus this change will not 
adversely impact the likelihood or probability of a design basis 
accident.
    The TS 3.9.6 and TS 3.9.7 requirements do not prevent the accident 
conditions from occurring and do not limit the severity of the 
accident.
    Therefore the TS 3.9.6 and TS 3.9.7 relocation to the TRM would not 
cause a significant increase in the accident probability or accident 
consequences.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This proposed change relocates TS 3.9.6 (Refuel Machine) and TS 
3.9.7 (Crane Travel) to the Waterford 3 TRM. In general, Technical 
Specifications are based upon the accident analyses. The accident 
analyses assumptions and initial conditions must be protected by the 
Technical Specifications. This is a requirement as outlined in [10 CFR 
50.36].
    [10 CFR 50.36(b)] states the technical specifications will be 
derived from the analyses and evaluation included in the safety 
analysis report.
    [10 CFR 50.36(c)(2)(i)] states that [``]the limiting conditions for 
operation are the lowest functional capability or performance levels of 
equipment required for safe operation of the facility[. . . .''] [10 
CFR 50.36(c)(2)(ii)] provides the four criteria in which any one met 
requires a limiting condition for operation. The proposed change 
demonstrated that the [10 CFR 50.36(c)(2)(ii)] criteria were not met 
and the relocation to the TRM is allowable. By not meeting the [10 CFR 
50.36(c)(2)(ii)] criteria for inclusion into the TS means that TS 3.9.6 
and TS 3.9.7 do not impact the accident analyses previously evaluated 
and would not create the possibility of a new or different kind of 
accident.
    Specifically, TS 3.9.6 and TS 3.9.7 equipment are not 
instrumentation used to detect, and indicate in the control room, a 
significant abnormal degradation of the reactor coolant pressure 
boundary (Criterion 1). TS 3.9.6 and TS 3.9.7 do not contain a process 
variable, design feature, or operating restriction that is an initial 
condition of a Design Basis Accident or Transient analysis that either 
assumes the failure of or presents a challenge to the integrity of a 
fission product barrier (Criterion 2). TS 3.9.6 and TS 3.9.7 does not 
contain a structure, system, or component that is part of the primary 
success path and which functions or actuates to mitigate a Design Basis 
Accident or Transient that either assumes the failure of or presents a 
challenge to the integrity of a fission product barrier (Criterion 3). 
Lastly, TS 3.9.6 and TS 3.9.7 do not contain a structure, system, or 
component which operating experience or probabilistic safety assessment 
has shown to be significant to public health and safety (Criterion 4).
    TS 3.9.6 and 3.9.7 are not required to meet the lowest functional 
capability or performance levels of equipment required for safe 
operation of the facility.
    Therefore, the accident analyses are not impacted and the 
possibility of a new or different kind of accident from any accident 
previously evaluated has not changed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed TS 3.9.6 (Refuel Machine) and TS 3.9.7 (Crane Travel) 
relocation to the Waterford 3 TRM is administrative in nature because 
all requirements will be relocated. Any changes after being relocated 
to the Waterford 3 TRM will require that the [10 CFR 50.59] process be 
entered ensuring the public health and safety is maintained. By using 
the [10 CFR 50.59] process for future changes, the regulatory 
requirements ensure that no significant reduction in the margin of 
safety occurs.
    In addition, the TS 3.9.6 and TS 3.9.7 requirements do not prevent 
the design basis accident conditions from occurring and do not limit 
the severity of the accident. Thus, TS 3.9.6 and TS 3.9.7 relocation 
will not adversely impact the accident analyses and will not cause a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Joseph A. Aluise, Associate General 
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Meena K. Khanna.
Exelon Generation Company, LLC (EGC), Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York
    Date of amendment request: November 17, 2014. A publicly available 
version is in ADAMS under Accession No. ML14321A744.
    Description of amendment request: The proposed amendment would 
revise the NMP2 Technical Specification (TS) Allowable Value for the 
Main Steam Line Tunnel Lead Enclosure Temperature-High instrumentation 
from an ambient temperature dependent (variable setpoint) to ambient 
temperature independent (constant Allowable Value). The changes would 
delete Surveillance Requirement (SR) 3.3.6.1.2 and revise the Allowable 
Value for Function 1.g on Table 3.3.6.1-1, ``Primary Containment 
Isolation Instrumentation.''
    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:


[[Page 11477]]


    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated because 
the performance of any equipment credited in the radiological 
consequences of an accident is not affected by the change in the leak 
detection capability.
    The Main Steam Line Tunnel Lead Enclosure Temperature--High is 
provided to detect a steam leak in the lead enclosure and provides 
diversity to the high flow instrumentation. This function provides a 
mitigating action for a steam leak in the Main Steam Line Tunnel Lead 
Enclosure, which could lead to a pipe break. This function does not 
affect any accident precursors, and the proposed changes do not affect 
the leak detection capability. Additionally, the proposed changes do 
not degrade the performance of or increase the challenges to any safety 
systems assumed to function in the accident analysis.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed changes do not add or remove equipment and do not 
physically alter the isolation instrumentation. In addition, the Main 
Steam Line Tunnel Lead Enclosure LDS [Leak Detection System] is not 
utilized in a different manner. The proposed changes do not introduce 
any new accident initiators and new failure modes, nor do they reduce 
or adversely affect the capabilities of any plant structure, system, or 
component to perform their safety function. The Main Steam Line Tunnel 
Lead Enclosure LDS will continue to be operated in the same manner.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not involve a significant reduction in a 
margin of safety because the changes eliminate the temperature setpoint 
dependency on lead enclosure temperature while maintaining the existing 
upper AV [Allowable Value] = 175.6[emsp14][deg]F, that was previously 
evaluated and approved. There is no adverse impact on the existing 
equipment capability as well as associated structures. The increase in 
the steam leak rate and associated crack size continues to be well 
below the leak rate associated with critical crack size that leads to 
pipe break. The proposed changes continue to provide the same level of 
protection against a main steam line break as the existing setpoint 
values.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Senior Vice President, 
Regulatory Affairs, Nuclear, and General Counsel, Exelon Generation 
Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Benjamin G. Beasley.
Florida Power and Light Company, et al. (FPL), Docket Nos. 50-335 and 
50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
    Date of amendment request: February 20, 2014, as supplemented by 
letters dated December 11, 2014, January 13 and January 28, 2015. 
Publicly-available in ADAMS under Accession Nos. ML14070A087, 
ML14349A333, ML15029A497 and ML15042A122.
    Description of amendment request: The NRC staff has previously made 
a proposed determination that the amendment request dated February 20, 
2014, involves no significant hazards consideration (see 79 FR 42550, 
July 22, 2014). Subsequently, by letter dated January 28, 2015, the 
licensee provided additional information that expanded the scope of the 
amendment request as originally noticed. Accordingly, this notice 
supersedes the previous notice in its entirety.
    The amendment would revise the Technical Specifications (TSs) by 
relocating specific surveillance frequency requirements to a licensee-
controlled program with implementation of Nuclear Energy Institute 
(NEI) 04-10 (Revision 1), ``Risk-Informed Technical Specifications 
Initiative 5b, Risk-Informed Method for Control of Surveillance 
Frequencies'' (ADAMS Accession No. ML071360456). The licensee stated 
that the NEI 04-10 methodology provides reasonable acceptance 
guidelines and methods for evaluating the risk increase of proposed 
changes to surveillance frequencies, consistent with Regulatory Guide 
1.177, ``An Approach for Plant-Specific, Risk-Informed Decisionmaking: 
Technical Specifications'' (ADAMS Accession No. ML003740176). The 
licensee stated that the changes are consistent with NRC-approved 
Technical Specification Task Force (TSTF) Improved Standard Technical 
Specifications Change Traveler TSTF-425, ``Relocate Surveillance 
Frequencies to Licensee Control--RITSTF [Risk-Informed Technical 
Specifications Task Force] Initiative 5b,'' Revision 3 (ADAMS Accession 
No. ML090850642). The Federal Register notice published on July 6, 2009 
(74 FR 31996), announced the availability of TSTF-425, Revision 3. In 
the supplement dated January 28, 2015, the licensee requested (1) 
additional surveillance frequencies be relocated to the licensee-
controlled program, (2) editorial changes, (3) administrative 
deviations from TSTF-425, and (4) other changes resulting from 
differences between the St. Lucie Plant TSs and the TSs on which TSTF-
425 was based.
    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 change relocates the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
significantly increased. The systems and components required by the 
technical specifications for which the surveillance frequencies are 
relocated are still required to be operable, meet the acceptance 
criteria for the surveillance requirements, and be capable of 
performing any mitigation function assumed in the accident analysis. 
As a result, the consequences of any accident previously evaluated 
are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the

[[Page 11478]]

probability or consequences of any 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.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis assumptions and current plant operating 
practice.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, FPL will 
perform a probabilistic risk evaluation using the guidance contained 
in NRC-approved NEI 04-10, Revision 1 in accordance with the TS 
Surveillance Frequency Control Program. NEI 04-10, Revision 1, 
methodology provides reasonable acceptance guidelines and methods 
for evaluating the risk increase of proposed changes to surveillance 
frequencies consistent with Regulatory Guide (RG) 1.177.

    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: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Shana R. Helton.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Nuclear Generating, Unit Nos. 3 and 4, Miami-Dade County, Florida
    Date of amendment request: November 13, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14337A013.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3/4.5.2, ``ECCS [Emergency Core Cooling 
System] Subsystems--T<INF>avg</INF> [average temperature] Greater Than 
or Equal to 350[emsp14][deg]F [degrees Fahrenheit],'' to correct non-
conservative TS requirements. The licensee also requested editorial 
changes to the TS.
    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 as follows:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed TS changes involve TS 3.5.2 Action `a', new TS 
3.5.2 Action `h', and the provision in SR [Surveillance Requirement] 
4.5.2.a to address non-conservative TS requirements. Editorial changes 
are also proposed for consistency and clarity. These changes do not 
affect any precursors to any accident previously evaluated and 
subsequently, will not impact the probability or consequences of an 
accident previously evaluated. Furthermore, these changes do not 
adversely affect mitigation equipment or strategies.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed TS changes involve TS 3.5.2 Action `a', new TS 
3.5.2 Action `h', and the provision in SR 4.5.2.a to address non-
conservative TS requirements. Editorial changes are also proposed for 
consistency and clarity. The proposed changes provide better assurance 
that the ECCS systems, subsystems, and components are properly aligned 
to support safe reactor operation consistent with the licensing basis 
requirements. The proposed changes do not introduce new modes of plant 
operation and do not involve physical modifications to the plant (no 
new or different type of equipment will be installed). There are no 
changes in the method by which any safety related plant structure, 
system, or component (SSC) performs its specified safety function. As 
such, the plant conditions for which the design basis accident analyses 
were performed remain valid.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a result 
of the proposed change. There will be no adverse effect or challenges 
imposed on any SSC as a result of the proposed change.
    Therefore, the proposed changes do 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 the 
margin of safety?
    No. Margin of safety is related to confidence in the ability of the 
fission product barriers to perform their accident mitigation 
functions. The proposed TS changes involve TS 3.5.2 Action `a', new TS 
3.5.2 Action `h', and the provision in SR 4.5.2.a to address non-
conservative TS requirements. Editorial changes are also proposed for 
consistency and clarity. The proposed changes provide better assurance 
that the ECCS systems, subsystems, and components are properly aligned 
to support safe reactor operation consistent with the licensing basis 
requirements. The proposed changes do not physically alter any SSC. 
There will be no effect on those SSCs necessary to assure the 
accomplishment of specified functions. There will be no impact on the 
overpower limit, departure from nucleate boiling ratio (DNBR) limits, 
loss of cooling accident peak cladding temperature (LOCA PCT), or any 
other margin of safety. The applicable radiological dose consequence 
acceptance criteria will continue to be met. 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: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 
33408-0420.
    NRC Branch Chief: Shana R. Helton.
Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
    Date of amendment request: February 6, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15041A069.
    Description of amendment request: The proposed amendments would 
modify the technical specifications requirements for unavailable 
barriers by adding limiting condition for operation

[[Page 11479]]

(LCO) 3.0.8. The changes are consistent with the NRC approved Technical 
Specification Task Force (TSTF) Standard Technical Specification change 
TSTF-427, ``Allowance for Non-Technical Specification Barrier 
Degradation on Supported System OPERABILITY,'' Revision 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
affirmed the applicability of the model proposed no significant hazards 
consideration published on October 3, 2006 (71 FR 58444), ``Notice of 
Availability of the Model Safety Evaluation.'' The findings presented 
in that evaluation are 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 a delay time for entering a supported 
system technical specification (TS) when the inoperability is due 
solely to an unavailable barrier if risk is assessed and managed. The 
postulated initiating events which may require a functional barrier are 
limited to those with low frequencies of occurrence, and the overall TS 
system safety function would still be available for the majority of 
anticipated challenges. Therefore, the probability of an accident 
previously evaluated is not significantly increased, if at all. The 
consequences of an accident while relying on the allowance provided by 
proposed LCO 3.0.8 are no different than the consequences of an 
accident while relying on the TS required actions in effect without the 
allowance provided by proposed LCO 3.0.8. Therefore, the consequences 
of an accident previously evaluated are not significantly affected by 
this change. The addition of a requirement to assess and manage the 
risk introduced by this change will further minimize possible concerns. 
Therefore, this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From any Previously Evaluated
    The proposed change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed). 
Allowing delay times for entering supported system TS when 
inoperability is due solely to an unavailable barrier, if risk is 
assessed and managed, will not introduce new failure modes or effects 
and will not, in the absence of other unrelated failures, lead to an 
accident whose consequences exceed the consequences of accidents 
previously evaluated. The addition of a requirement to assess and 
manage the risk introduced by this change will further minimize 
possible concerns. Thus, this change does not create the possibility of 
a new or different kind of accident from an accident previously 
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety
    The proposed change allows a delay time for entering a supported 
system TS when the inoperability is due solely to an unavailable 
barrier, if risk is assessed and managed. The postulated initiating 
events which may require a functional barrier are limited to those with 
low frequencies of occurrence, and the overall TS system safety 
function would still be available for the majority of anticipated 
challenges. The risk impact of the proposed TS changes was assessed 
following the three-tiered approach recommended in RG [Regulatory 
Guide] 1.177. A bounding risk assessment was performed to justify the 
proposed TS changes. This application of LCO 3.0.8 is predicated upon 
the licensee's performance of a risk assessment and the management of 
plant risk. The net change to the margin of safety is insignificant as 
indicated by the anticipated low levels of associated risk (ICCDP 
[incremental conditional core damage probability] and ICLERP 
[incremental large early release probability]) as shown in Table 1 of 
Section 3.1.1 in the Safety Evaluation. Therefore, this change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, Michigan 49106.
    NRC Branch Chief: David L. Pelton.

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

    Date of amendment requests: October 27, 2014. A publicly-
available version is available in ADAMS under Accession No. 
ML14317A052.
    Description of amendment requests: The proposed amendments will 
modify the Susquehanna technical specifications (TS). Specifically, 
the proposed amendments will modify the TS by relocating specific 
surveillance frequencies to a licensee-controlled program, the 
Surveillance Frequency Control Program (SFCP), with implementation 
of Nuclear Energy Institute (NEI) 04-10, ``Risk-Informed Technical 
Specifications Initiative 5b, Risk-Informed Method for Control of 
Surveillance Frequencies'' (ADAMS Accession No. ML071360456). The 
changes are consistent with NRC-approved TS Task Force (TSTF) 
Standard TS change TSTF-425, ``Relocate Surveillance Frequencies to 
Licensee Control-Risk Informed Technical Specifications Task Force 
(RITSTF) Initiative 5b,'' Revision 3 (ADAMS Accession No. 
ML090850642). The Federal Register notice published on July 6, 2009 
(74 FR 31996), announced the availability of this TSTF improvement, 
and included a model no significant hazards consideration and safety 
evaluation.
    Basis for proposed no significant hazards consideration 
determination: An analysis of the no significant hazards 
consideration was presented in the TSTF-425. The licensee has 
affirmed its applicability of the model no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in the 
probability or consequences of any accident previously evaluated?
    Response: No.
    The proposed change relocates the specified frequencies for 
periodic surveillance requirements to licensee control under a new 
Surveillance Frequency Control Program. Surveillance frequencies are 
not an initiator to any accident previously evaluated. As a result, 
the probability of any accident previously evaluated is not 
significantly increased. The systems and components required by the 
technical specifications for which the surveillance frequencies are 
relocated are still required to be operable, meet the acceptance 
criteria for the surveillance requirements, and be capable of 
performing any mitigation function assumed in the accident analysis. 
As a result, the consequences of any accident previously evaluated 
are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different 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. In addition, the changes do not impose any new or 
different requirements. The changes do not alter assumptions made in 
the safety analysis. The proposed changes are consistent with the 
safety analysis assumptions and current plant operating practice.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

[[Page 11480]]

    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The design, operation, testing methods, and acceptance criteria 
for systems, structures, and components (SSCs), specified in 
applicable codes and standards (or alternatives approved for use by 
the NRC) will continue to be met as described in the plant licensing 
basis (including the final safety analysis report and bases to TS), 
since these are not affected by changes to the surveillance 
frequencies. Similarly, there is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis. To 
evaluate a change in the relocated surveillance frequency, PPL will 
perform a risk evaluation using the guidance contained in NRC 
approved NEI 04-10, Rev. 1 in accordance with the TS SFCP. NEI 04-
10, Rev. 1, methodology provides reasonable acceptance guidelines 
and methods for evaluating the risk increase of proposed changes to 
surveillance frequencies consistent with Regulatory Guide 1.177.
    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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Branch Chief: Douglas A. Broaddus.
Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-424 and 
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia
    Date of amendment request: July 18, 2014. A publicly-available 
version is in ADAMS under Accession Package No. ML14203A124.
    Description of amendment request: The licensee requested 23 
revisions to the Technical Specifications (TSs). These revisions adopt 
various previously NRC-approved Technical Specifications Task Force 
(TSTF) Travelers. A list of the requested revisions is included in 
Enclosure 1 of the application.
    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 for each of the 24 changes requested, which is presented 
below:

1: TSTF-2-A, Revision 1, ``Relocate the 10 Year Sediment Cleaning of 
the Fuel Oil Storage Tank to Licensee Control'' for TS pages 3.8.3-3 
and 3.8.3-4

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change removes the Surveillance Requirement for 
performing sediment cleaning of diesel fuel oil storage tanks every 
10 years from the Technical Specifications and places it under 
licensee control. Diesel fuel oil storage tank cleaning is not an 
initiator of any accident previously evaluated. This change will 
have no effect on diesel generator fuel oil quality, which is tested 
in accordance with other Technical Specifications requirements. 
Removing the diesel fuel oil storage tank sediment cleaning 
requirements from the Technical Specifications will have no effect 
on the ability to mitigate an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change removes the requirement to clean sediment 
from the diesel fuel oil storage tank from the Technical 
Specifications and places it under licensee control. The margin of 
safety provided by the fuel oil storage tank sediment cleaning is 
unaffected by this relocation because the quality of diesel fuel oil 
is tested in accordance with other Technical Specifications 
requirements.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2: TSTF-27-A, Revision 3, ``Revise SR [Surveillance Requirement] 
Frequency for Minimum Temperature for Criticality'' for TS 3.4.2, TS 
Page 3.4.2-1

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the Surveillance Frequency for 
monitoring [reactor coolant system] RCS temperature to ensure the 
minimum temperature for criticality is met. The Frequency is changed 
from a 30 minute Frequency when certain conditions are met to a 
periodic Frequency that it is controlled in accordance with the 
Surveillance Frequency Control Program. The initial Frequency for 
this Surveillance will be 12 hours. This will ensure that 
T<INF>avg</INF> [average temperature] is logged at appropriate 
intervals (in addition to strip chart recorders and computer logging 
of temperature). The measurement of RCS temperature is not an 
initiator of any accident previously evaluated. The minimum RCS 
temperature for criticality is not changed. As a result, the 
mitigation of any accident previously evaluated is not affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises the Surveillance Frequency for 
monitoring RCS temperature to ensure the minimum temperature for 
criticality is met. The current, condition based Frequency 
represents a distraction to the control room operator during the 
critical period of plant startup. RCS temperature is closely 
monitored by the operator during the approach to criticality, and 
temperature is recorded on charts and computer logs. Allowing the 
operator to monitor temperature as needed by the situation and 
logging RCS temperature at a periodic Frequency that it is 
controlled in accordance with the Surveillance Frequency Control 
Program is sufficient to ensure that the LCO [Limiting Condition for 
Operation] is met while eliminating a diversion of the operator's 
attention.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

[[Page 11481]]

3: TSTF-28-A, Revision 0, ``Delete Unnecessary Action to Measure Gross 
Specific Activity, TS 3.4.16,'' TS page 3.4-16

    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 Required Action B.1 of 
Specification 3.4.16, ``RCS Specific Activity,'' which requires 
verifying that Dose Equivalent I-131 specific activity is within 
limits. Determination of Dose Equivalent I-131 is not an initiator 
of any accident previously evaluated. Determination of Dose 
Equivalent I-131 has no effect on the mitigation of any accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change eliminates a Required Action. The activities 
performed under the Required Action will still be performed to 
determine if the LCO is met or the plant will exit the Applicability 
of the Specification. In either case, the presence of the Required 
Action does not provide any significant margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

4: TSTF-45-A, Revision 2, ``Exempt Verification of CIVs that are 
Locked, Sealed or Otherwise Secured,'' TS 3.6.3, TS pages 3.6.3-4, 
3.6.3-5

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change exempts containment isolation valves (CIVs) 
located inside and outside of containment that are locked, sealed, 
or otherwise secured in position from the periodic verification of 
valve position required by Surveillance Requirements 3.6.3.3 and 
3.6.2.4. The exempted valves are verified to be in the correct 
position upon being locked, sealed, or secured. Because the valves 
are in the condition assumed in the accident analysis, the proposed 
change will not affect the initiators or mitigation of any accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change replaces the periodic verification of valve 
position with verification of valve position followed by locking, 
sealing, or otherwise securing the valve in position. Periodic 
verification is also effective in detecting valve mispositioning. 
However, verification followed by securing the valve in position is 
effective in preventing valve mispositioning. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

5: TSTF-46-A, Revision 1, ``Clarify the CIV Surveillance to Apply Only 
to Automatic Isolation Valves,'' TS 3.6.3, TS page 3.6.3.5

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the requirements in Technical 
Specification SR 3.6.3.5, and the associated Bases, to delete the 
requirement to verify the isolation time of ``each power operated'' 
containment isolation valve (CIV) and only require verification of 
closure time for each ``automatic power operated isolation valve.'' 
The closure times for CIVs that do not receive an automatic closure 
signal are not an initiator of any design basis accident or event, 
and therefore the proposed change does not increase the probability 
of any accident previously evaluated. The CIVs are used to respond 
to accidents previously evaluated. Power operated CIVs that do not 
receive an automatic closure signal are not assumed to close in a 
specified time. The proposed change does not change how the plant 
would mitigate an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not result in a change in the manner in 
which the CIVs provide plant protection or introduce any new or 
different operational conditions. Periodic verification that the 
closure times for CIVs that receive an automatic closure signal are 
within the limits established by the accident analysis will continue 
to be performed under SR 3.6.3.5. The change does not alter 
assumptions made in the safety analysis, and is consistent with the 
safety analysis assumptions and current plant operating practice. 
There are also no design changes associated with the proposed 
changes, and the change does not involve a physical alteration of 
the plant (i.e., no new or different type of equipment will be 
installed).
    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 provides clarification that only CIVs that 
receive an automatic isolation signal are within the scope of the SR 
3.6.3.5. The proposed change does not result in a change in the 
manner in which the CIVs provide plant protection. Periodic 
verification that closure times for CIVs that receive an automatic 
isolation signal are within the limits established by the accident 
analysis will continue to be performed. The proposed change does not 
affect the safety analysis acceptance criteria for any analyzed 
event, nor is there a change to any safety analysis limit. The 
proposed change does not alter the manner in which safety limits, 
limiting safety system settings or limiting conditions for operation 
are determined, nor is there any adverse effect on those plant 
systems necessary to assure the accomplishment of protection 
functions. The proposed change will not result in plant operation in 
a configuration outside the design basis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

6: TSTF-87-A, Revision 2, ``Revise `RTBs [Reactor Trip Breaker] Open' 
and `CRDM [Control Rod Drive Mechanism] De-energized' Actions to 
`Incapable of Rod Withdrawal,''' TS 3.4.5, TS Pages 3.4.5-2, 3.4.9-1

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

[[Page 11482]]

    This change revises the Required Actions for LCO 3.4.5, ``RCS 
Loops--Mode 3,'' Conditions C.2 and D.1, from ``De-energize all 
control rod drive mechanisms,'' to ``Place the Rod Control System in 
a condition incapable of rod withdrawal.'' It also revises LCO 
3.4.9, ``Pressurizer,'' Required Action A.1, from requiring Reactor 
Trip Breakers to be open after reaching MODE 3 to ``Place the Rod 
Control System in a condition incapable of rod withdrawal,'' and to 
require full insertion of all rods. Inadvertent rod withdrawal can 
be an initiator for design basis accidents or events during certain 
plant conditions, and therefore must be prevented under those 
conditions. The proposed Required Actions for LCO 3.4.5 and LCO 
3.4.9 satisfy the same intent as the current Required Actions, which 
is to prevent inadvertent rod withdrawal when an applicable 
Condition is not met, and is consistent with the assumptions of the 
accident analysis. As a result, the proposed change does not 
increase the probability of any accident previously evaluated. The 
proposed change does not change how the plant would mitigate an 
accident previously evaluated, as in both the current and proposed 
requirements, rod withdrawal is prohibited.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change provides less specific, but equivalent, 
direction on the manner in which inadvertent control rod withdrawal 
is to be prevented when the Conditions of LCO 3.4.5 and LCO 3.4.9 
are not met. Rod withdrawal will continue to be prevented when the 
applicable Conditions of LCO 3.4.5 and LCO 3.4.9 are met. There are 
no design changes associated with the proposed changes, and the 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed). The change 
does not alter assumptions made in the safety analysis, and is 
consistent with the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change provides the operational flexibility of 
allowing alternate, but equivalent, methods of preventing rod 
withdrawal when the applicable Conditions of LCO 3.4.5 and LCO 3.4.9 
are met. The proposed change does not affect the safety analysis 
acceptance criteria for any analyzed event, nor is there a change to 
any safety analysis limit. The proposed change does not alter the 
manner in which safety limits, limiting safety system settings or 
limiting conditions for operation are determined, nor is there any 
adverse effect on those plant systems necessary to assure the 
accomplishment of protection functions. The proposed change will not 
result in plant operation in a configuration outside the design 
basis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

7: TSTF-95-A, Revision 0, ``Revise Completion Time for Reducing Power 
Range High trip Setpoint from 8 to 72 Hours,'' TS 3.2.1, TS Pages 
3.2.1-1 and 3.2.2-1

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the time allowed to reduce the Power 
Range Neutron Flux--High trip setpoint when Specification 3.2.1, 
``Heat Flux Hot Channel Factor,'' or Specification 3.2.2, ``Nuclear 
Enthalpy Rise Hot Channel Factor,'' are not within their limits. 
Both specifications require a power reduction followed by a 
reduction in the Power Range Neutron Flux--High trip setpoint. 
Because reactor power has been reduced, the reactor core power 
distribution limits are within the assumptions of the accident 
analysis. Reducing the Power Range Neutron Flux--High trip setpoints 
ensures that reactor power is not inadvertently increased. Reducing 
the Power Range Neutron Flux--High trip setpoints is not an 
initiator to any accident previously evaluated. The consequences of 
any accident previously evaluated with the Power Range Neutron 
Flux--High trip setpoints not reduced are no different under the 
proposed Completion Time than under the existing Completion Time. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change provides additional time before requiring 
the Power Range Neutron Flux--High trip setpoint be reduced when the 
reactor core power distribution limits are not met. The manual 
reduction in reactor power required by the specifications provides 
the necessary margin of safety for this condition. Reducing the 
Power Range Neutron Flux--High trip setpoints carries an increased 
risk of a reactor trip. Delaying the trip setpoint reduction until 
the power reduction has been completed and the condition is verified 
will minimize overall plant risk.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

8: TSTF-110-A, Revision 2, ``Delete SR Frequencies Based on Inoperable 
Alarms,'' TS 3.1, TS pages 3.1.4-3, 3.1.6-3, 3.2.3-1, 3.2.4-4

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change removes surveillance Frequencies associated 
with inoperable alarms (rod position deviation monitor, rod 
insertion limit monitor, AFD [Axial Flux Difference] monitor and 
QPTR [Quadrant Power Tilt Ratio] alarm) from the Technical 
Specifications and places the actions in plant administrative 
procedures. The subject plant alarms are not an initiator of any 
accident previously evaluated. The subject plant alarms are not used 
to mitigate any accident previously evaluated, as the control room 
indications of these parameters are sufficient to alert the operator 
of an abnormal condition without the alarms. The alarms are not 
credited in the accident analysis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change removes surveillance Frequencies associated 
with inoperable alarms (rod position deviation monitor, rod 
insertion limit monitor, AFD monitor and QPTR alarm) from the 
Technical Specifications and places the actions in plant 
administrative procedures. The alarms are not being removed from the 
plant. The actions to be taken when the alarms are not available are 
proposed to be controlled under licensee administrative procedures. 
As a result, plant operation is unaffected by this change and there 
is no effect on a margin of safety.

[[Page 11483]]

    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

9: TSTF-142-A, Revision 0, ``Increase the Completion Time When the Core 
Reactivity Balance is Not Within Limit,'' TS 3.1.2, TS Page 3.1.2-1

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the Completion Time to take the 
Required Actions when measured core reactivity is not within the 
specified limit of the predicted values. The Completion Time to 
respond to a difference between predicted and measured core 
reactivity is not an initiator to any accident previously evaluated. 
The consequences of an accident during the proposed Completion Time 
are no different from the consequences of an accident during the 
existing Completion Time. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change provides additional time to investigate and 
to implement appropriate operating restrictions when measured core 
reactivity is not within the specified limit of the predicted 
values. The additional time will not have a significant effect on 
plant safety due to the conservatisms used in designing the reactor 
core and performing the safety analyses and the low probability of 
an accident or transient which would approach the core design limits 
during the additional time. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

10: TSTF-234-A, Revision 1, ``Add Action for More Than One [D]RPI 
Inoperable,'' TS 3.1.7, TS Pages 3.1.7-1 and 3.1.7-2.

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change provides a Condition and Required Actions 
for more than one inoperable digital rod position indicator (DRPI) 
per rod group. The DRPIs are not an initiator of any accident 
previously evaluated. The DRPIs are one indication used by operators 
to verify control rod insertion following an accident, however other 
indications are available. Therefore, allowing a finite period to 
time to correct more than one inoperable DRPI prior to requiring a 
plant shutdown will not result in a significant increase in the 
consequences of any accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change provides time to correct the condition of 
more than one DRPI inoperable in a rod group. Compensatory measures 
are required to verify that the rods monitored by the inoperable 
DRPIs are not moved to ensure that there is no effect on core 
reactivity. Requiring a plant shutdown with inoperable rod position 
indications introduces plant risk and should not be initiated unless 
the rod position indication cannot be repaired in a reasonable 
period of time. As a result, the safety benefit provided by the 
proposed Condition offsets the small decrease in safety resulting 
from continued operation with more than one inoperable DRPI.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

11: TSTF-245-A, Revision 1, ``AFW Train Operable When in Service,'' TS 
3.7.5, TS Page 3.7.5-3

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the requirements in Technical 
Specification 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to 
clarify the operability of an AFW train when it is aligned for 
manual steam generator level control. The AFW System is not an 
initiator of any design basis accident or event, and therefore the 
proposed change does not increase the probability of any accident 
previously evaluated. The AFW System is used to respond to accidents 
previously evaluated. The proposed change does not affect the design 
of the AFW System, and no physical changes are made to the plant. 
The proposed change does not significantly change how the plant 
would mitigate an accident previously evaluated. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not result in a change in the manner in 
which the AFW System provides plant protection. The AFW System will 
continue to supply water to the steam generators to remove decay 
heat and other residual heat by delivering at least the minimum 
required flow rate to the steam generators. There are no design 
changes associated with the proposed changes, and the change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed). The change does not 
alter assumptions made in the safety analysis, and is consistent 
with the safety analysis assumptions and current plant operating 
practice. Manual control of AFW level control valves is not an 
accident initiator.
    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?
    Responses: No.
    The proposed change provides the operational flexibility of 
allowing an AFW train(s) to be considered operable when it is not in 
the normal standby alignment and is temporarily incapable of 
automatic initiation, such as during alignment and operation for 
manual steam generator level control, provided it is capable of 
being manually realigned to the AFW heat removal mode of operation. 
The proposed change does not result in a change in the manner in 
which the AFW System provides plant protection. The AFW System will 
continue to supply water to the steam generators to remove decay 
heat and other residual heat by delivering at least the minimum 
required flow rate to the steam generators. The proposed change does 
not affect the safety analysis acceptance criteria for any analyzed 
event, nor is there a change to any safety analysis limit. The 
proposed change does not alter the manner in which safety limits, 
limiting safety system settings

[[Page 11484]]

or limiting conditions for operation are determined, nor is there 
any adverse effect on those plant systems necessary to assure the 
accomplishment of protection functions. The proposed change will not 
result in plant operation in a configuration outside the design 
basis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

12: TSTF-247-A, Revision 0, ``Provide Separate Condition Entry for Each 
[Power Operated Relief Valve] PORV and Block Valve,'' TS 3.4.11, TS 
Pages 3.4.11-1, 3.4.11-2, 3.4.11-3

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the requirements in Technical 
Specification 3.4.11, ``Pressurizer PORVs,'' to clarify that 
separate Condition entry is allowed for each block valve. 
Additionally, the Actions are modified to no longer require that the 
PORVs be placed in manual operation when both block valves are 
inoperable and cannot be restored to operable status within the 
specified Completion Time. This preserves the overpressure 
protection capabilities of the PORVs. The pressurizer block valves 
are used to isolate their respective PORV in the event it is 
experiencing excessive leakage, and are not an initiator of any 
design basis accident or event. Therefore the proposed change does 
not increase the probability of any accident previously evaluated. 
The PORV and block valves are used to respond to accidents 
previously evaluated. The proposed change does not affect the design 
of the PORV and block valves, and no physical changes are made to 
the plant. The proposed change does not change how the plant would 
mitigate an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not result in a change in the manner in 
which the PORV and block valves provide plant protection. The PORVs 
will continue to provide overpressure protection, and the block 
valves will continue to provide isolation capability in the event a 
PORV is experiencing excessive leakage. There are no design changes 
associated with the proposed changes, and the change does not 
involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed). The change does not 
alter assumptions made in the safety analysis, and is consistent 
with the safety analysis assumptions and current plant operating 
practice. Operation of the PORV block valves is not an accident 
initiator.
    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 changes provide clarification that separate 
Condition entry is allowed for each block valve. Additionally, the 
Actions are modified to no longer require that the PORVs be placed 
in manual operation when both block valves are inoperable and cannot 
be restored to operable status within the specified Completion Time. 
This preserves the overpressure protection capabilities of the 
PORVs. The proposed change does not result in a change in the manner 
in which the PORV and block valves provide plant protection. The 
PORVs will continue to provide overpressure protection, and the 
block valves will continue to provide isolation capability in the 
event a PORV is experiencing excessive leakage. The proposed change 
does not affect the safety analysis acceptance criteria for any 
analyzed event, nor is there a change to any safety analysis limit. 
The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined, nor is there any adverse effect on those 
plant systems necessary to assure the accomplishment of protection 
functions. The proposed change will not result in plant operation in 
a configuration outside the design basis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

13: TSTF-248-A, Revision 0, ``Revise Shutdown Margin Definition for 
Stuck Rod Exception,'' TS 1.1, TS Page 1.1-6

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change modifies the definition of Shutdown Margin 
to eliminate the requirement to assume the highest worth control rod 
is fully withdrawn when calculating Shutdown Margin if it can be 
verified by two independent means that all control rods are 
inserted. The method for calculating shutdown margin is not an 
initiator of any accident previously evaluated. If it can be 
verified by two independent means that all control rods are 
inserted, the calculated Shutdown Margin without the conservatism of 
assuming the highest worth control rod is withdrawn is accurate and 
consistent with the assumptions in the accident analysis. As a 
result, the mitigation of any accident previously evaluated is not 
affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change modifies the definition of Shutdown Margin 
to eliminate the requirement to assume the highest worth control rod 
is fully withdrawn when calculating Shutdown Margin if it can be 
verified by two independent means that all control rods are 
inserted. The additional margin of safety provided by the assumption 
that the highest worth control rod is fully withdrawn is unnecessary 
if it can be independently verified that all controls rods are 
inserted.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

14: TSTF-266-A, Revision 3, ``Eliminate the Remote Shutdown System 
Table of Instrumentation and Controls,'' TS 3.3.4, TS Pages 3.3.4-1, 
3.3.4-3

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change removes the list of Remote Shutdown System 
instrumentation and controls from the Technical Specifications and 
places them in the Bases. The Technical Specifications continue to 
require that the instrumentation and controls be operable. The 
location of the list of Remote Shutdown System instrumentation and 
controls is not an initiator to any accident previously evaluated. 
The proposed change will have no effect on the mitigation of any 
accident previously evaluated because the instrumentation and 
controls continue to be required to be operable.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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.

[[Page 11485]]

    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change removes the list of Remote Shutdown System 
instrumentation and controls from the Technical Specifications and 
places it in the Bases. The review performed by the NRC when the 
list of Remote Shutdown System instrumentation and controls is 
revised will no longer be needed unless the criteria in 10 CFR 50.59 
are not met such that prior NRC review is required. The Technical 
Specification requirement that the Remote Shutdown System be 
operable, the definition of operability, the requirements of 10 CFR 
50.59, and the Technical Specifications Bases Control Program are 
sufficient to ensure that revision of the list without prior NRC 
review and approval does not introduce a significant safety risk.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

15: TSTF-272-A, Revision 1, ``Refueling Boron Concentration 
Clarification,'' TS 3.9.1, TS Page 3.9.1-1

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change modifies the Applicability of Specification 
3.9.1, ``Boron Concentration,'' to clarify that the boron 
concentration limits are only applicable to the refueling canal and 
the refueling cavity when those volumes are attached to the Reactor 
Coolant System (RCS). The boron concentration of water volumes not 
connected to the RCS are not an initiator of an accident previously 
evaluated. The ability to mitigate any accident previously evaluated 
is not affected by the boron concentration of water volumes not 
connected to the RCS.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change modifies the Applicability of Specification 
3.9.1, ``Boron Concentration,'' to clarify that the boron 
concentration limits are only applicable to the refueling canal and 
the refueling cavity when those volumes are attached to the RCS. 
Technical Specification SR 3.0.4 requires that Surveillances be met 
prior to entering the Applicability of a Specification. As a result, 
the boron concentration of the refueling cavity or the refueling 
canal must be verified to satisfy the LCO prior to connecting those 
volumes to the RCS. The margin of safety provided by the refueling 
boron concentration is not affected by this change as the RCS boron 
concentration will continue to satisfy the LCO.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

16: TSTF-273-A, Revision 2, ``Safety Function Determination Program 
Clarifications,'' TS 5.5.15, TS Page 5.5-15

    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 add explanatory text to the programmatic 
description of the Safety Function Determination Program (SFDP) in 
Specification 5.5.15 to clarify in the requirements that 
consideration does not have to be made for a loss of power in 
determining loss of function. The Bases for LCO 3.0.6 is revised to 
provide clarification of the ``appropriate LCO for loss of 
function,'' and that consideration does not have to be made for a 
loss of power in determining loss of function. The changes are 
editorial and administrative in nature, and therefore do not 
increase the probability of any accident previously evaluated. No 
physical or operational changes are made to the plant. The proposed 
change does not change how the plant would mitigate an accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes are editorial and administrative in nature 
and do not result in a change in the manner in which the plant 
operates. The loss of function of any specific component will 
continue to be addressed in its specific TS LCO and plant 
configuration will be governed by the required actions of those 
LCOs. The proposed changes are clarifications that do not degrade 
the availability or capability of safety related equipment, and 
therefore do not create the possibility of a new or different kind 
of accident from any accident previously evaluated. There are no 
design changes associated with the proposed changes, and the changes 
do not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed). The changes do not 
alter assumptions made in the safety analysis, and are consistent 
with the safety analysis assumptions and current plant operating 
practice. Due to the administrative nature of the changes, they 
cannot be an accident initiator.
    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 changes to TS 5.5.15 are clarifications and are 
editorial and administrative in nature. No changes are made the LCOs 
for plant equipment, the time required for the TS Required Actions 
to be completed, or the out of service time for the components 
involved. The proposed changes do not affect the safety analysis 
acceptance criteria for any analyzed event, nor is there a change to 
any safety analysis limit. The proposed changes do not alter the 
manner in which safety limits, limiting safety system settings or 
limiting conditions for operation are determined, nor is there any 
adverse effect on those plant systems necessary to assure the 
accomplishment of protection functions. The proposed changes will 
not result in plant operation in a configuration outside the design 
basis.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

17: TSTF-284-A, Revision 3, ``Add `Met vs. Perform' to Technical 
Specification 1.4, Frequency,'' TS 1.4, TS 3.4, TS 3.9, TS Pages 1.4-1, 
1.4-4, 3.4.11-3, 3.4.12-4 and 3.9.4-2

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes insert a discussion paragraph into 
Specification 1.4, and several new examples are added to facilitate 
the use and application of SR Notes that utilize the terms ``met'' 
and ``perform.'' The changes also modify SRs in multiple 
Specifications to appropriately use ``met'' and ``perform'' 
exceptions. The changes are administrative in nature because they 
provide clarification and correction of existing expectations, and 
therefore the proposed change does not increase the probability of 
any accident previously evaluated. No physical or

[[Page 11486]]

operational changes are made to the plant. The proposed change does 
not significantly change how the plant would mitigate an accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes are administrative in nature and do not 
result in a change in the manner in which the plant operates. The 
proposed changes provide clarification and correction of existing 
expectations that do not degrade the availability or capability of 
safety related equipment, and therefore do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. There are no design changes associated with 
the proposed changes, and the changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed). The changes do not alter assumptions made in the 
safety analysis, and are consistent with the safety analysis 
assumptions and current plant operating practice. Due to the 
administrative nature of the changes, they cannot be an accident 
initiator.
    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 changes are administrative in nature and do not 
result in a change in the manner in which the plant operates. The 
proposed changes provide clarification and correction of existing 
expectations that do not degrade the availability or capability of 
safety related equipment, or alter their operation. The proposed 
changes do not affect the safety analysis acceptance criteria for 
any analyzed event, nor is there a change to any safety analysis 
limit. The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined, nor is there any adverse effect on those 
plant systems necessary to assure the accomplishment of protection 
functions. The proposed changes will not result in plant operation 
in a configuration outside the design basis.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

18: TSTF-308-A, Revision 1, ``Determination of Cumulative and Projected 
Dose Contributions in RECP [Radioactive Effluent Controls Program],'' 
TS 5.5.4, TS Page 5.5-3

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises Specification 5.5.4, ``Radioactive 
Effluent Controls Program,'' paragraph e, to describe the original 
intent of the dose projections. The cumulative and projection of 
doses due to liquid releases are not an assumption in any accident 
previously evaluated and have no effect on the mitigation of any 
accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises Specification 5.5.4, ``Radioactive 
Effluent Controls Program,'' paragraph e, to describe the original 
intent of the dose projections. The cumulative and projection of 
doses due to liquid releases are administrative tools to assure 
compliance with regulatory limits. The proposed change revises the 
requirement to clarify the intent, thereby improving the 
administrative control over this process. As a result, any effect on 
the margin of safety should be minimal.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

19: TSTF-312-A, Revision 1, ``Administrative Control of Containment 
Penetrations,'' TS 3.9.4, TS Page 3.9.4-1

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would allow containment penetrations to be 
unisolated under administrative controls during core alterations or 
movement of irradiated fuel assemblies within containment. The 
status of containment penetration flow paths (i.e., open or closed) 
is not an initiator for any design basis accident or event, and 
therefore the proposed change does not increase the probability of 
any accident previously evaluated. The proposed change does not 
affect the design of the primary containment, or alter plant 
operating practices such that the probability of an accident 
previously evaluated would be significantly increased. The proposed 
change does not significantly change how the plant would mitigate an 
accident previously evaluated, and is bounded by the fuel handling 
accident (FHA) accident analysis.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Allowing penetration flow paths to be open is not an initiator 
for any accident. The proposed change to allow open penetration flow 
paths will not affect plant safety functions or plant operating 
practices such that a new or different accident could be created. 
There are no design changes associated with the proposed changes, 
and the change does not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed). The 
change does not alter assumptions made in the safety analysis, and 
is consistent with the safety analysis assumptions and current plant 
operating practice.
    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.
    TS 3.9.4 provides measures to ensure that the dose consequences 
of a postulated FHA inside containment are minimized. The proposed 
change to LCO 3.9.4 will allow penetration flow path(s) to be open 
during refueling operations under administrative control. These 
administrative controls will can and will be achieved in the event 
of an FHA inside containment, and will minimize dose consequences. 
The proposed change is bounded by the existing FHA analysis. The 
proposed change does not affect the safety analysis acceptance 
criteria for any analyzed event, nor is there a change to any safety 
analysis limit. The proposed change does not alter the manner in 
which safety limits, limiting safety system settings or limiting 
conditions for operation are determined, nor is there any adverse 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. The proposed change will not result in 
plant operation in a configuration outside the design basis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

[[Page 11487]]

20: TSTF-314-A, Revision 0, ``Require Static and Transient 
F<INF>Q</INF> Measurement,'' TS 3.1.4, 3.2.4, TS Pages 3.1.4-2, 3.2.4-
1, 3.2.4-3

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the Required Actions of 
Specification 3.1.4, ``Rod Group Alignment Limits,'' and 
Specification 3.2.4, ``Quadrant Power Tilt Ratio,'' to require 
measurement of both the steady state and transient portions of the 
Heat Flux Hot Channel Factor, FQ(Z). This change will ensure that 
the hot channel factors are within their limits when the rod 
alignment limits or quadrant power tilt ratio are not within their 
limits. The verification of hot channel factors is not an initiator 
of any accident previously evaluated. The verification that both the 
steady state and transient portion of FQ(Z) are within their limits 
will ensure this initial assumption of the accident analysis is met 
should a previously evaluated accident occur.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises the Required Actions in the 
Specifications for Rod Group Alignment Limits and Quadrant Power 
Tilt Ratio to require measurement of both the steady state and 
transient portions of the Heat Flux Hot Channel Factor, 
F<INF>Q</INF>(Z). This change is a correction that ensures that the 
plant conditions are as assumed in the accident analysis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

21: TSTF-340-A, Revision 3, ``Allow 7 Day Completion Time for a 
Turbine--Driven AFW Pump Inoperable,'' TS 3.7.5, TS Page 3.7.5-1

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises Specification 3.7.5, ``Auxiliary 
Feedwater (AFW) System,'' to allow a 7 day Completion Time to 
restore an inoperable AFW turbine-driven pump in Mode 3 immediately 
following a refueling outage, if Mode 2 has not been entered. An 
inoperable AFW turbine-driven pump is not an initiator of any 
accident previously evaluated. The ability of the plant to mitigate 
an accident is no different while in the extended Completion Time 
than during the existing Completion Time.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in of safety?
    Response: No.
    The proposed change revises Specification 3.7.5, ``Auxiliary 
Feedwater (AFW) System,'' to allow a 7-day Completion Time to 
restore an inoperable turbine-driven AFW pump in Mode 3 immediately 
following a refueling outage if Mode 2 has not been entered. In Mode 
3 immediately following a refueling outage, core decay heat is low 
and the need for AFW is also diminished. The two operable motor 
driven AFW pumps are available and there are alternate means of 
decay heat removal if needed. As a result, the risk presented by the 
extended Completion Time is minimal.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

22: TSTF-343-A, Revision 1, ``Containment Structural Integrity,'' TS 
5.5, TS Page 5.5-16

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the Technical Specifications (TS) 
Administrative Controls programs for consistency with the 
requirements of 10 CFR 50, paragraph 55a(g)(4) for components 
classified as Code Class CC. The proposed changes affect the 
frequency of visual examinations that will be performed for the 
steel containment liner plate for the purpose of the Containment 
Leakage Rate Testing Program.
    The frequency of visual examinations of the containment and the 
mode of operation during which those examinations are performed does 
not affect the initiation of any accident previously evaluated. The 
use of NRC approved methods and frequencies for performing the 
inspections will ensure the containment continues to perform the 
mitigating function assumed for accidents previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change revises the TS Administrative Controls 
programs for consistency with the requirements of 10 CFR 50, 
paragraph 55a(g)(4) for components classified as Code Class CC. The 
proposed change affects the frequency of visual examinations that 
will be performed for the steel containment liner plate for the 
purpose of the Containment Leakage Rate Testing Program.
    The proposed changes do not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed changes will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released off-site and there is no increase in individual or 
cumulative occupational exposure.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes revise the Technical Specifications (TS) 
Administrative Controls programs for consistency with the 
requirements of 10 CFR 50, paragraph 55a(g)(4) for components 
classified as Code Class CC. The proposed change affects the 
frequency of visual examinations that will be performed for the 
steel containment liner plate for the purpose of the Containment 
Leakage Rate Testing Program. The safety function of the containment 
as a fission product barrier will be maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

23: TSTF-349-A, Revision 1, ``Add Note to LCO 3.9.5 Allowing Shutdown 
Cooling Loops Removal From Operation,'' TS 3.9.6, TS Page 3.9.6-1

    1. Does the proposed amendment involve a significant increase in 
the probability or

[[Page 11488]]

consequences of an accident previously evaluated?
    Response: No.
    The proposed change adds an LCO Note to LCO 3.9.6, ``RHR and 
Coolant Circulation--Low Water Level,'' to allow securing the 
operating train of Residual Heat Removal (RHR) for up to 15 minutes 
to support switching operating trains. The allowance is restricted 
to conditions in which core outlet temperature is maintained at 
least 10 degrees F below the saturation temperature, when there are 
no draining operations, and when operations that could reduce the 
reactor coolant system (RCS) boron concentration are prohibited. 
Securing an RHR train to facilitate the changing of the operating 
train is not an initiator to any accident previously evaluated. The 
restrictions on the use of the allowance ensure that an RHR train 
will not be needed during the 15 minute period to mitigate any 
accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any 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 a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change adds an LCO Note to LCO 3.9.6, ``RHR and 
Coolant Circulation--Low Water Level,'' to allow securing the 
operating train of RHR to support switching operating trains. The 
allowance is restricted to conditions in which core outlet 
temperature is maintained at least 10 degrees F below the saturation 
temperature, when there are no draining operations, and when 
operations that could reduce the reactor coolant system (RCS) boron 
concentration are prohibited. With these restrictions, combined with 
the short time frame allowed to swap operating RHR trains and the 
ability to start an operating RHR train if needed, the occurrence of 
an event that would require immediate operation of an RHR train is 
extremely remote.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Leigh D. Perry, SVP & General Counsel of 
Operations and Nuclear, Southern Nuclear Operating Company, 40 Iverness 
Center Parkway, Birmingham, AL 35201.
    NRC Branch Chief: Robert J. Pascarelli.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power, Unit Nos. 1 and 2, Louisa County, Virginia
    Date of amendment request: February 4, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15041A667.
    Description of amendment request: The proposed license amendment 
requests the changes to the Technical Specification (TS) TS 3.1.7, Rod 
Position Indication, to provide an additional monitoring option for an 
inoperable control rod position indicator. Specifically, the proposed 
changes would allow monitoring of control rod drive mechanism 
stationary gripper coil voltage every eight hours as an alternative to 
using the movable in core detectors every eight hours to verify control 
rod position.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change provides an alternative method for verifying 
rod position of one rod. The proposed change meets the intent of the 
current specification in that it ensures verification of position of 
the rod once every 8 hours. The proposed change provides only an 
alternative method of monitoring rod position and does not change 
the assumptions or results of any previously evaluated accident.
    Therefore, operation of the facility in accordance with the 
proposed amendment would 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 only an alternative method of 
determining the position of one rod. No new accident initiators are 
introduced by the proposed alternative manner of performing rod 
position verification. The proposed change does not affect the 
reactor protection system. Hence, no new failure modes are created 
that would cause a new or different kind of accidents from any 
accident previously evaluated.
    Therefore, operation of the facility in accordance with the 
proposed amendments 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 basis of TS 3.1.7 states that the operability of the rod 
position indicators is required to determine control rod positions 
and thereby ensure compliance with the control rod alignment and 
insertion limits. The proposed change does not alter the requirement 
to determine rod position but provides an alternative method for 
determining the position of the affected rod. As a result, the 
initial conditions of the accident analysis are preserved and the 
consequences of previously analyzed accidents are unaffected.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.
    Based on the above, Dominion concludes 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Branch Chief: Robert Pascarelli.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating

[[Page 11489]]

license or combined license, as applicable, proposed no significant 
hazards consideration determination, and opportunity for a hearing in 
connection with these actions, was published in the Federal Register as 
indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin
    Date of application for amendment: May 29, 2013, as supplemented by 
letters dated September 23, October 15, October 17, October 31, and 
November 7, 2013, and January 7, March 13, April 29, and October 6, 
2014, and January 15, 2015.
    Brief description of amendment: The amendment revised the Renewed 
Facility Operating License and associated Technical Specifications to 
conform to the permanent shutdown and defueled status of the facility. 
It also denied a proposal to delete paragraphs 1.B, 1.I, and 1.J of the 
Kewaunee Operating License.
    Date of issuance: February 13, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 215. A publicly-available version is in ADAMS under 
Accession No. ML14237A045; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-43: The amendment 
revised the renewed facility operating license and Technical 
Specifications.
    Date of initial notice in Federal Register: August 20, 2013 (78 FR 
51224). The supplemental letters dated September 23, October 15, 
October 17, October 31, and November 7, 2013, and January 7, March 13, 
April 29, and October 6, 2014, and January 15, 2015, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 13, 2015.
    No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 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, 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 County, South 
Carolina
    Date of application for amendments: July 21, 2014.
    Brief description of amendments: The amendment revises the licensed 
operator training requirements to be consistent with the National 
Academy for Nuclear Training (NANT) program. Additionally, the 
amendment makes administrative changes to Technical Specification 
Sections 5.1, ``Responsibility;'' 5.2, ``Organization;'' 5.3, ``Unit 
Staff Qualifications;'' 5.5, ``Programs and Manuals;'' and for Catawba 
and McGuire, Section 5.7, ``High Radiation Area.''
    Date of issuance: February 12, 2015.
    Effective date: This license amendment is effective as of its date 
of issuance and shall be implemented within 120 days of issuance.
    Amendment Nos.: 273, 269, 276, 256, 389, 391, and 390. A publicly-
available version is available in ADAMS under Accession No. 
ML15002A324.
    Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, NPF-
17, DPR-38, DPR-47, and DPR-55: Amendments revised the licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: November 12, 2014 (79 
FR 67199).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 12, 2015.
    No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas
    Date of application for amendment: December 17, 2012, as 
supplemented by letters dated November 7, and December 4, 2013; January 
6, May 22, June 30, August 7, September 24, and December 9, 2014.
    Brief description of amendment: The amendment authorized the 
transition of the Arkansas Nuclear One, Unit No. 2, fire protection 
program to a risk-informed, performance-based program based on National 
Fire Protection Association (NFPA) 805, in accordance with 10 CFR 
50.48(c). NFPA 805 allows the use of performance-based methods such as 
fire modeling and risk-informed methods such as fire probabilistic risk 
assessment to demonstrate compliance with the nuclear safety 
performance criteria.
    Date of issuance: February 18, 2015.
    Effective date: As of its date of issuance and shall be implemented 
by 6 months from the date of issuance.
    Amendment No.: 300. A publicly-available version is in ADAMS under 
Accession No. ML14356A227; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPR-6: Amendment revised the 
License and Technical Specifications.
    Date of initial notice in Federal Register: July 23, 2013 (78 FR 
44171). The supplemental letters dated November 7 and December 4, 2013; 
and January 6, May 22, June 30, August 7, September 24, and December 9, 
2014, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 18, 2015.
    No significant hazards consideration comments received: No.
Entergy Nuclear FitzPatrick, LLC and Entergy Nuclear Operations, Inc., 
Docket No. 50-333, James A. FitzPatrick Nuclear Power Plant, Oswego 
County, New York
    Date of amendment request: October 8, 2013, as supplemented by a 
letter dated November 18, 2014.
    Brief description of amendment: The amendment modifies the 
Technical Specifications (TSs) to reduce the reactor steam dome 
pressure associated

[[Page 11490]]

with the Reactor Core Safety Limit from 785 psig to 685 psig in TS 
2.1.1.1 and TS 2.1.1.2. This change addresses the potential to not meet 
the pressure/thermal power/minimal critical power ratio TS safety limit 
during a pressure regulator failure-maximum demand (open) (PRFO) 
transient. The PRFO transient was reported by General Electric as a 
notification pursuant to Title 10 of the Code of Federal Regulations, 
Part 21, ``Reporting of Defects and Noncompliance.''
    Date of issuance: February 9, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 309. A publicly-available version is in ADAMS under 
Accession No. ML15014A277; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-59: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2014 (79 FR 
38589). The supplemental letter dated November 18, 2014, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 9, 2015.
    No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont
    Date of amendment request: November 14, 2013, as supplemented by 
letters dated June 9, 2014, August 6, 2014, and October 9, 2014.
    Description of amendment request: The amendment eliminates 
operability requirements for secondary containment when handling 
sufficiently decayed irradiated fuel or a fuel cask following a minimum 
of 13 days after the permanent cessation of reactor operation.
    Date of Issuance: February 12, 2015.
    Effective date: The license amendment becomes effective 13 days 
after the licensee's submittal of the certifications, as required by 10 
CFR 50.82(a)(1)(i) and (ii).
    Amendment No.: 262. A publicly-available version is in ADAMS under 
Accession No. ML14304A588; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-28: The amendment revised the 
Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 16, 2014 (79 
FR 55511).
    The supplemental letters dated June 9, 2014, August 6, 2014, and 
October 9, 2014, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the NRC staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated February 12, 2015.
    No significant hazards consideration comments received: No.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center, Linn County, Iowa
    Date of amendment request: June 23, 2014.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) requirements to address NRC Generic Letter (GL) 
2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay 
Heat Removal, and Containment Spray Systems,'' as described in TSTF-
523, Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation.''
    Date of issuance: February 10, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 290. A publicly-available version is in ADAMS under 
Accession No. ML15014A200; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-49: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: September 30, 2014 (79 
FR 58820).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 10, 2015.
    No significant hazards consideration comments received: No
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
No. 1, Rockingham County, New Hampshire
    Date of amendment request: June 24, 2014, as supplemented by letter 
dated December 11, 2014.
    Brief description of amendment: The amendment revised the Seabrook 
Technical Specifications (TSs). Specifically, the amendment modifies 
Seabrook TSs to address U.S. Nuclear Regulatory Commission Generic 
Letter (GL) 2008-01, ``Managing Gas Accumulation in Emergency Core 
Cooling, Decay Heat Removal, and Containment Spray Systems,'' as 
described in TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing 
Gas Accumulation.''
    Date of issuance: February 6, 2015.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 144. A publicly-available version is in ADAMS under 
Accession No. ML14345A288; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-86: The amendment revised the 
License and TS.
    Date of initial notice in Federal Register: September 2, 2014 (79 
FR 52066). The supplemental letter dated December 11, 2014, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 6, 2015.
    No significant hazards consideration comments received: No.
South Carolina Electric and Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, 
Fairfield County, South Carolina
    Date of amendment request: November 15, 2011, as supplemented by 
letters dated November 22, 2011; January 26 and October 10, 2012; 
February 1, April 1, October 14, and November 26, 2013; January 9, 
February 25, May 2, May 11, August 14, October 9, and December 11, 
2014.
    Brief description of amendment: The amendment authorizes the 
transition of the V.C. Summer fire protection program to a risk-
informed, performance-based program based on

[[Page 11491]]

National Fire Protection Association (NFPA) 805, ``Performance-Based 
Standard for Fire Protection for Light Water Reactor Electric 
Generating Plants, 2001 Edition'' (NFPA 805), in accordance with 10 CFR 
50.48(c).
    Date of issuance: February 11, 2015.
    Effective date: This amendment is effective as of its date of 
issuance and shall be implemented per the December 11, 2014, 
supplement, Attachment S, Table S-2 ``Implementation Items'', requiring 
full implementation by March 31, 2016.
    Amendment No.: 199. A publicly-available version is in ADAMS under 
Accession No. ML14287A289; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-12: Amendment revised 
the Facility Operating License.
    Date of initial notice in Federal Register: August 14, 2012 (77 FR 
48561). The supplemental letters dated November 22, 2011; October 10, 
2012; February 1, April 1, October 14, and November 26, 2013; January 
9, February 25, May 2, May 11, August 14, October 9, and December 11, 
2014, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 11, 2015.
    No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear 
Plant (HNP), Unit No. 2, Appling County, Georgia
    Date of amendment request: August 8, 2014, as supplemented by 
letters dated September 8 and October 24, 2014.
    Brief description of amendments: The amendment revises the 
Technical Specification value of the Safety Limit Minimum Critical 
Power Ratio to support operation in the next fuel cycle.
    Date of issuance: February 18, 2015.
    Effective date: As of the date of issuance and shall be implemented 
prior to reactor startup following the HNP, Unit 2, spring 2015 
refueling outage.
    Amendment No(s).: 218. A publicly-available version is in ADAMS 
under Accession No. ML15020A434; documents related to this amendment 
are listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: Amendment 
revised the licenses and the Technical Specifications.
    Date of initial notice in Federal Register: January 6, 2015, (80 FR 
536).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 18, 2015.
    No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project (STP), Units 1 and 2, Matagorda County, Texas
    Date of amendment request: July 23, 2013, as supplemented by 
letters dated May 12 (two letters), May 19, and December 17, 2014.
    Brief description of amendments: The amendments revised the STP, 
Units 1 and 2, Fire Protection Program (FPP) related to the alternate 
shutdown capability. Specifically, it approves the following operator 
actions in the control room prior to evacuation due to a fire for 
meeting the alternate shutdown capability, in addition to manually 
tripping the reactor that is currently credited in the STP, Units 1 and 
2, FPP licensing basis:
    <bullet> Initiate main steam line isolation
    <bullet> Closing the pressurizer power-operated relief valves block 
valves
    <bullet> Securing all reactor coolant pumps
    <bullet> Closing feedwater isolation valves
    <bullet> Securing the startup feedwater pump
    <bullet> Isolating reactor coolant system letdown
    <bullet> Securing the centrifugal charging pumps
    In addition, the licensee credits the automatic trip of the main 
turbine upon the initiation of a manual reactor trip for meeting the 
alternate shutdown capability.
    Date of issuance: February 13, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment Nos.: Unit 1--203; Unit 2--191. A publicly-available 
version is in ADAMS under Accession No. ML14339A170; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: October 29, 2013 (78 FR 
64546). The supplements dated May 12 (two letters), May 19, and 
December 17, 2014, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 13, 2015.
    No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear 
Plant, Unit 1, Limestone County, Alabama
    Date of amendment request: December 18, 2013, as supplemented by 
letter dated June 13, 2014.
    Brief description of amendment: The amendment revised the Technical 
Specification (TS) 3.4.9, ``RCS [Reactor Coolant System] Pressure and 
Temperature (P/T) Limits,'' Figures 3.4.9-1 through 3.4.9-2. The P/T 
limits are based on proprietary topical report NEDC-33178P-A, Revision 
1, ``GE [General Electric] Hitachi Nuclear Energy Methodology for 
Development of Reactor Pressure Vessel Pressure-Temperature Curves.'' 
NEDO-33178-A, Revision 1 is the non-proprietary version of the NRC-
approved topical report.
    Date of issuance: February 2, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 287. A publicly available version is in ADAMS under 
Accession No. ML14325A501; documents related to this amendment are 
listed in the Safety Evaluation (SE) enclosed with the amendment.
    Renewed Facility Operating License No. DPR-33: Amendment revised 
the TSs and the Operating License.
    Date of initial notice in Federal Register: May 6, 2014 (79 FR 
25902). The supplemental letter dated June 13, 2014, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in the SE dated February 2, 2015.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 23rd day of February 2015.


[[Page 11492]]


    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2015-04298 Filed 3-2-15; 8:45 am]
BILLING CODE 7590-01-P


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