Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 77565-77574 [2011-31901]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 76, Number 239 (Tuesday, December 13, 2011)]
[Notices]
[Pages 77565-77574]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2011-31901]


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NUCLEAR REGULATORY COMMISSION

[NRC-2011-0285]


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

Background

    Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC) is publishing this regular biweekly notice. The Act 
requires the Commission publish notice of any amendments issued, or 
proposed to be issued and grants the Commission the authority to issue 
and make immediately effective any amendment to an operating license 
upon a determination by the Commission that such amendment involves no 
significant hazards consideration, notwithstanding the pendency before 
the Commission of a request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from November 17 to November 30, 2011. The last 
biweekly notice was published on November 29, 2011 (76 FR 73727).

ADDRESSES: Please include Docket ID NRC-2011-0285 in the subject line 
of your comments. Comments submitted in writing or in electronic form 
will be posted on the NRC Web site and on the Federal rulemaking Web 
site https://www.regulations.gov. Because your comments will not be 
edited to remove any identifying or contact information, the NRC 
cautions you against including any information in your submission that 
you do not want to be publicly disclosed.
    The NRC requests that any party soliciting or aggregating comments 
received from other persons for submission to the NRC inform those 
persons that the NRC will not edit their comments to remove any 
identifying or contact information, and therefore, they should not 
include any information in their comments that they do not want 
publicly disclosed.
    You may submit comments by any one of the following methods.
    Federal Rulemaking Web Site: Go to https://www.regulations.gov and 
search for documents filed under Docket ID NRC-2011-0285. Address 
questions about NRC dockets to Carol Gallagher (301) 492-3668; email 
Carol.Gallagher@nrc.gov.
    Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and 
Directives Branch (RADB), Office of Administration, Mail Stop: TWB-05-
B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
    Fax comments to: RADB at (301) 492-3446.
    You can access publicly available documents related to this notice 
using the following methods:
    NRC's Public Document Room (PDR): The public may examine and have 
copied for a fee publicly available documents at the NRC's PDR, Room 
O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland 20852.
    NRC's Agencywide Documents Access and Management System (ADAMS): 
Publicly available documents created or received at the NRC are 
accessible electronically through ADAMS in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. From this page, the public can gain 
entry into ADAMS, which provides text and image files of NRC's public 
documents. If you do not have access to ADAMS or if there are problems 
in accessing the documents located in ADAMS, contact the NRC's PDR 
reference staff at 1-(800) 397-4209, (301) 415-4737, or by email to 
pdr.resource@nrc.gov.
    Federal Rulemaking Web Site: Public comments and supporting 
materials related to this notice can be found at https://www.regulations.gov by searching on Docket ID: NRC-2011-0285.

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

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

[[Page 77566]]

will publish in the Federal Register a notice of issuance. Should the 
Commission make a final No Significant Hazards Consideration 
Determination, any hearing will take place after issuance. The 
Commission expects that the need to take this action will occur very 
infrequently.
    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license. 
Requests for a hearing and a petition for leave to intervene shall be 
filed in accordance with the Commission's ''Rules of Practice for 
Domestic Licensing Proceedings'' in 10 CFR part 2. Interested person(s) 
should consult a current copy of 10 CFR 2.309, which is available at 
the NRC's PDR, located at One White Flint North, Room O1-F21, 11555 
Rockville Pike (first floor), Rockville, Maryland 20874. NRC 
regulations are accessible electronically from the NRC Library on the 
NRC Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. If 
a request for a hearing or petition for leave to intervene is filed by 
the above date, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also identify the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment.
    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC E-Filing rule (72 FR 49139, 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at hearing.docket@nrc.gov, or by 
telephone at (301) 415-1677, to request (1) A digital identification 
(ID) certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions

[[Page 77567]]

should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the 
time the documents are submitted through the NRC's E-Filing system. To 
be timely, an electronic filing must be submitted to the E-Filing 
system no later than 11:59 p.m. Eastern Time on the due date. Upon 
receipt of a transmission, the E-Filing system time-stamps the document 
and sends the submitter an email notice confirming receipt of the 
document. The E-Filing system also distributes an email notice that 
provides access to the document to the NRC Office of the General 
Counsel and any others who have advised the Office of the Secretary 
that they wish to participate in the proceeding, so that the filer need 
not serve the documents on those participants separately. Therefore, 
applicants and other participants (or their counsel or representative) 
must apply for and receive a digital ID certificate before a hearing 
request/petition to intervene is filed so that they can obtain access 
to the document via the E-Filing system.
    A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC Web site 
at https://www.nrc.gov/site-help/e-submittals.html, by email at 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-(866) 672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 
20852, Attention: Rulemaking and Adjudications Staff. Participants 
filing a document in this manner are responsible for serving the 
document on all other participants. Filing is considered complete by 
first-class mail as of the time of deposit in the mail, or by courier, 
express mail, or expedited delivery service upon depositing the 
document with the provider of the service. A presiding officer, having 
granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. With 
respect to copyrighted works, except for limited excerpts that serve 
the purpose of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Non-timely filings 
will not be entertained absent a determination by the presiding officer 
that the petition or request should be granted or the contentions 
should be admitted, based on a balancing of the factors specified in 10 
CFR 2.309(c)(1)(i)-(viii).
    For further details with respect to this license amendment 
application, see the application for amendment which is available for 
public inspection at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20874. Publicly available documents created or received at the NRC are 
accessible electronically through ADAMS in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to 
ADAMS or who encounter problems in accessing the documents located in 
ADAMS, should contact the NRC's PDR Reference staff at 1-(800) 397-
4209, (301) 415-4737, or by email to pdr.resource@nrc.gov.

Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of amendment request: October 18, 2011.
    Description of amendment request: The proposed amendment involves 
administrative changes. The proposed changes include correcting 
typographical errors, removing unwarranted formatting, clarifying 
symbols and pages, reformatting of previously deleted pages, 
incorporating a consistent abbreviation of average reactor coolant 
temperature, deleting notes that are no longer applicable, and 
replacing certain drawing figures with versions that are more clear.
    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, with NRC edits in brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    No physical changes to the facility will occur as a result of 
this proposed amendment. The proposed changes will not alter the 
physical design or operational procedures associated with any plant 
structure, system, or component. The proposed changes are 
administrative in nature and have no affect on plant operation.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are administrative in nature. The proposed 
changes do not alter the physical design, safety limits, or safety 
analysis assumptions associated with the operation of the plant. 
Accordingly, the changes do not introduce any new accident 
initiators, nor do they reduce or adversely affect the capabilities 
of any plant structure, system, or component to perform their safety 
function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes [maintain compliance with the requirements 
contained in 10 CFR 50.36, ``Technical specifications.''] The 
proposed changes are administrative in nature. The proposed changes 
do not alter the physical design, safety limits, or safety analysis 
assumptions associated with the operation of the plant.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, with the NRC edits above, it

[[Page 77568]]

appears that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota

    Date of amendment request: August 11, 2011.
    Description of amendment request: The proposed amendments would 
make changes to the diesel fuel oil license bases and amend technical 
specifications (TS) 3.7.8, ``Cooling Water (CL) System'' and 3.8.3, 
``Diesel Fuel Oil.'' The proposed TS changes would revise current 
requirements to reflect the addition of the license bases, resolve non-
conservative emergency diesel generator fuel oil supply volumes, 
incorporate portions of Technical Specification Task Force Traveler 
501, ``Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee 
Control,'' and provide administrative changes to the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This license amendment request proposes addition of a diesel 
fuel oil supply license basis and revision of the associated 
Technical Specifications to require an adequate emergency diesel 
generator and diesel driven cooling water pump fuel oil supply for 
mitigation of a design basis accident with a loss of offsite power. 
This license amendment request also proposes to: adopt provisions of 
Technical Specifications Task Force (TSTF) industry traveler 501 
(TSTF-501) to specify diesel fuel oil supply requirements as 
required days for the supply and relocate the corresponding volume 
to the Technical Specification Bases; and, make minor wording 
changes to improve conformance to the content guidance of NUREG-
1431, ``Standard Technical Specifications, Westinghouse Plants.''
    The emergency diesel generators, diesel driven cooling water 
pumps and their supporting diesel fuel oil storage systems are not 
accident initiators and therefore the proposed diesel fuel oil 
supply license basis addition and proposed Technical Specification 
changes do not involve an increase in the probability of an 
accident.
    The proposed change to the emergency diesel generator fuel oil 
supply license basis and the associated Technical Specification 
changes will assure that the emergency diesel generator's diesel 
driven cooling water pumps perform their required design basis 
accident mitigation safety function with a loss of offsite power. 
Since the emergency diesel generators will provide required 
electrical power as assumed in the accident analyses and the cooling 
water diesel will provide cooling water as assumed in the accident 
analyses, the results of the previous accident analyses are not 
changed and the license basis changes proposed in this license 
amendment request do not involve a significant increase in the 
consequences of an accident.
    Specification of the diesel fuel oil supply requirements as 
required days supply in accordance with TSTF-501 continues to assure 
an adequate quantity of diesel fuel oil is required to be stored; 
the emergency diesel generators and diesel driven cooling water 
pumps will have sufficient diesel fuel oil to mitigate a design 
basis accident with a loss of offsite power, as assumed in the 
accident analyses, until the fuel supply can be replenished; and 
therefore, this change does not involve a significant increase in 
the consequences of an accident.
    The proposed minor Technical Specification wording changes to 
improve alignment with the content guidance of NUREG-1431 are 
administrative and thus do not involve an increase in the 
consequences of an accident.
    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.
    This license amendment request proposes addition of a diesel 
fuel oil supply license basis and revision of the associated 
Technical Specifications to require an adequate emergency diesel 
generator and diesel driven cooling water pump fuel oil supply for 
mitigation of a design basis accident with a loss of offsite power. 
This license amendment request also proposes to: adopt provisions of 
Technical Specifications Task Force (TSTF) industry traveler 501 
(TSTF-501) to specify diesel fuel oil supply requirements as 
required days for the supply and relocate the corresponding volume 
to the Technical Specification Bases; and, make minor wording 
changes to improve conformance to the content guidance of NUREG-
1431, ``Standard Technical Specifications, Westinghouse Plants.''
    The proposed diesel fuel oil supply license basis change and the 
associated Technical Specification changes assure that each 
emergency diesel generator and diesel driven cooling water pump has 
an adequate supply of diesel fuel oil, assuming an active single 
failure, to mitigate a design basis accident with a loss of offsite 
power until the fuel oil supply can be replenished. The proposed 
license basis change and associated Technical Specification changes 
do not create new failure modes or mechanisms and no new accident 
precursors are generated. The proposed specification of the diesel 
fuel oil supply requirements as required days supply in accordance 
with TSTF-501 does not create new failure modes or mechanisms and 
does not generate new accident[s]. These proposed changes do not 
challenge the performance or integrity of any safety-related system. 
Surveillance requirements for the emergency diesel generator and 
diesel driven cooling water pump fuel oil supplies will continue to 
demonstrate that the Limiting Conditions for Operation are met and 
the emergency diesel generators and diesel driven cooling water 
pumps have adequate supplies of diesel fuel oil to perform their 
safety functions.
    The proposed minor Technical Specification wording changes to 
improve alignment with the content guidance of NUREG-1431 are 
administrative and thus do not create the possibility of a new or 
different kind of accident.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This license amendment request proposes addition of a diesel 
fuel oil supply license basis and revision of the associated 
Technical Specifications to require an adequate emergency diesel 
generator and diesel driven cooling water pump fuel oil supply for 
mitigation of a design basis accident with a loss of offsite power. 
This license amendment request also proposes to: adopt provisions of 
Technical Specifications Task Force (TSTF) industry traveler 501 
(TSTF-501) to specify diesel fuel oil supply requirements as 
required days for the supply and relocate the corresponding volume 
to the Technical Specification Bases; and, make minor wording 
changes to improve conformance to the content guidance of NUREG-
1431, ``Standard Technical Specifications, Westinghouse Plants.''
    The proposed diesel fuel oil supply licensing basis addition and 
the associated Technical Specification changes involve the addition 
of a new requirement to assure that each emergency diesel generator 
and diesel driven cooling water pump has an adequate supply of 
diesel fuel oil, assuming an active single failure, to mitigate a 
design basis accident with a loss of offsite power until the fuel 
oil supply can be replenished. The current license basis for 
mitigation of an external flood without a single failure will be 
maintained. Therefore, margins of safety are increased and thus no 
margin of safety is reduced due to these changes.
    Specification of the diesel fuel oil supply requirements as 
required days supply in accordance with TSTF-501 continues to assure 
an adequate quantity of diesel fuel oil is required to be stored and 
thus does not reduce a margin of safety.

[[Page 77569]]

    The proposed minor Technical Specification wording changes to 
improve alignment with the content guidance of NUREG-1431 are 
administrative and thus do not involve a significant reduction in a 
margin of safety.
    The proposed Technical Specification changes do not adversely 
affect the availability, operability, or performance of safety-
related systems and components: the emergency diesel generators 
[and] diesel driven cooling water pumps will continue to perform 
their safety functions. The ability of operable structures, systems, 
and components to perform their designated safety functions are 
unaffected by these proposed changes. The operability requirements 
of the proposed Technical Specifications are consistent with the 
initial condition assumptions of the safety analyses, and the 
Surveillance requirements for the emergency diesel generator and 
diesel driven cooling water pump fuel oil supplies will assure that 
the Limiting Conditions for Operation are met and the emergency 
diesel generator's diesel driven cooling water pumps have adequate 
supplies of diesel fuel oil to perform their safety functions.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Acting Branch Chief: Terry A. Beltz.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Unit 1 and 2, San Luis Obispo County, 
California

    Date of amendment request: June 1, 2011.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW) 
System,'' TS 3.6.6, ``Containment Spray and Cooling Systems,'' TS 
3.8.1, ``AC [Alternating Current] Sources--Operating,'' TS 3.8.9, 
``Distribution Systems--Operating,'' and TS 1.3, ``Completion Times,'' 
Example 1.3-3. These changes are consistent with Technical 
Specification Task Force (TSTF) Change Travelers TSTF-245, Revision 1, 
``AFW Train Operable when in Service,'' TSTF-340, Revision 3, ``Allow 7 
day Completion Time for a Turbine-driven AFW Pump Inoperable,'' TSTF-
412, Revision 3, ``Provide Actions for One Steam Supply to Turbine 
Driven AFW/EFW [Emergency Feedwater] Pump Inoperable,'' and TSTF-439, 
Revision 2, ``Eliminate Second Completion Times Limiting Time From 
Discovery of Failure to Meet an LCO [Limiting Condition for 
Operation].''
    Specifically, the changes consistent with TSTF-245, Revision 1, and 
TSTF-340, Revision 3, would revise TS 3.7.5 to clarify the operability 
of an AFW train during alternate alignments and provide added 
flexibility in Mode 3 to repair and test the turbine-driven AFW (TDAFW) 
pump following a refueling outage. The changes consistent with TSTF-
412, Revision 3, would revise TS 3.7.5 to establish conditions, 
required actions, and completion times for the condition where one 
steam supply to the TDAFW is inoperable concurrent with an inoperable 
motor-driven AFW (MDAFW) train. The TSTF-412, Revision 3, Notice of 
Availability was published in the Federal Register on July 17, 2007 (72 
FR 39089), using the consolidated line item improvement process 
(CLIIP). The changes consistent with TSTF-439, Revision 2, would remove 
second completion times from TS Example 1.3-3; TS 3.6.6 Required 
Actions A.1, A.2, and C.1; TS 3.7.5 Required Actions A.1 and B.1; TS 
3.8.1 Required Actions A.2 and B.4; and TS 3.8.9 Required Actions A.1, 
B.1, and C.1. In addition, the amendment would add a new Condition B, 
required actions, and completion times to TS 3.7.5 to provide specific 
actions to be taken when automatic control of the MDAFW level control 
valves is not functional.
    Basis for proposed no significant hazards consideration 
determination: For the proposed changes related to TSTF-245, Revision 
1, TSTF-340, Revision 3, and new TS 3.7.5 Condition B, as required by 
10 CFR 50.91(a), the licensee has provided its analysis of the issue of 
no significant hazards consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the requirements in Technical 
Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to 
clarify the OPERABILITY of an AFW train during alternate alignments, 
to provide added flexibility in MODE 3 to repair and test the 
turbine driven AFW pump following a refueling outage, and to clarify 
the OPERABILITY of the turbine driven AFW train with one steam 
supply inoperable. 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 affects only the actions taken when portions of the 
AFW System are unavailable and does not affect the design of the AFW 
System. The change to TS 3.7.5 adding actions for inoperable 
automatic control of level control valves does not change any of the 
assumptions in accidents previously evaluated and would not have an 
impact on accident consequences. 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 represent 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 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. The changes to the 
Conditions and Required Actions do not change any existing accident 
scenarios, nor create any new or different accident scenarios.
    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. The 
proposed change 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, it is concluded that the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in plant operation in a configuration outside the design basis.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.

    For the proposed changes related to TSTF-412, Revision 3, in its 
application dated June 1, 2011, the licensee has affirmed the 
applicability of the model no significant hazards consideration 
published in the Federal Register as part of the CLIIP (72 FR 39093; 
July 17, 2007). As required by 10 CFR 50.91(a), an analysis of the 
issue of no significant

[[Page 77570]]

hazards consideration, from the model application, 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 Auxiliary/Emergency Feedwater (AFW/EFW) System is not an 
initiator of any design basis accident or event, and therefore the 
proposed changes do not increase the probability of any accident 
previously evaluated. The proposed changes to address the condition 
of one or two motor driven AFW/EFW trains inoperable and the turbine 
driven AFW/EFW train inoperable due to one steam supply inoperable 
do not change the response of the plant to any accidents.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not adversely 
affect the ability of structures, systems, and components (SSCs) to 
perform their intended safety function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not result in a change in the manner in 
which the AFW/EFW System provides plant protection. The AFW/EFW 
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. The changes to 
the Conditions and Required Actions do not change any existing 
accident scenarios, nor create any new or different accident 
scenarios.
    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 or 
eliminate any existing 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 changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by these changes. The proposed changes will not 
result in plant operation in a configuration outside the design 
basis.
    Therefore, it is concluded that the proposed change does not 
involve a significant reduction in a margin of safety.

    For the proposed changes related to TSTF-439, Revision 2, as 
required by 10 CFR 50.91(a), the licensee has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes eliminate certain Completion Times from the 
Technical Specifications. Completion Times are not an initiator to 
any accident previously evaluated. As a result, the probability of 
an accident previously evaluated is not affected. The consequences 
of an accident during the revised Completion Time are no different 
than the consequences of the same accident during the existing 
Completion Times. As a result, the consequences of an accident 
previously evaluated are not affected by this change. The proposed 
changes do not alter or prevent the ability of structures, systems, 
and components from performing their intended function to mitigate 
the consequences of an initiating event within the assumed 
acceptance limits. The proposed changes do not affect the source 
term, containment isolation, or radiological release assumptions 
used in evaluating the radiological consequences of an accident 
previously evaluated. Further, the proposed changes do not increase 
the types or amounts of radioactive effluent that may be released 
offsite, nor significantly increase individual or cumulative 
occupational/public radiation exposures. The proposed changes are 
consistent with the safety analysis assumptions and resultant 
consequences.
    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 accident from any accident previously evaluated?
    Response: No.
    The changes do not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The changes 
do not alter any assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to delete the second Completion Time does 
not alter the manner in which safety limits, limiting safety system 
settings or limiting conditions for operation are determined. The 
safety analysis acceptance criteria are not affected by this change. 
The proposed changes will not result in plant operation in a 
configuration outside of the design basis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    Based on the above, the NRC staff has reviewed the licensee's 
analyses 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: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: Michael T. Markley.

South Carolina Electric and Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station (VCSNS), 
Unit 1, Fairfield County, South Carolina

    Date of amendment request: October 12, 2011.
    Description of amendment request: The amendment requests 
authorization to update the facility's Final Safety Analysis Report to 
exempt five Unit 1 high-head safety injection system (HHSI) containment 
isolation valves (CIVs) from the VCSNS, Unit No. 1 Local Leak Rate 
Testing (LLRT) Program requirements.
    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 with changes in brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident that has previously 
been evaluated?
    Response: No.
    The amendment request is to remove five Containment Isolation 
Valves (XVG08801A, XVG08801B, XVG08884, XVG08885, and XVG08886) from 
the Local Leak Rate Test (LLRT) program. These valves were 
originally included in the LLRT under 10 CFR [part] 50, Appendix J, 
in what is now Option A. VCSNS has been approved for 10 CFR [Part] 
50, Appendix J, Option B under License Amendment No. 135. Under 
Option B, valves

[[Page 77571]]

may be excluded from LLRT Type C testing if they are not a potential 
containment atmosphere leakage path. Based on the design and 
operation of the Safety Injection System, the valves do not 
constitute a containment atmospheric leakage path as covered in the 
Safety Evaluation. Since the valves are not a leakage path, there is 
no impact on the consequence of an accident. Moreover, the valves 
are not a part of the Reactor Coolant Pressure Boundary and are 
normally closed during plant operation, thus they do not affect the 
probability of an accident in any way. [The change does not affect 
plant equipment or operating practices and therefore does not 
significantly increase the probability or consequences of an 
accident previously evaluated.]
    2. Does the proposed change create the possibility of a new or 
different kind of accident of malfunction that has not previously 
been evaluated?
    Response: No.
    The system design and operation are not changing. This test [* * 
*] [change] does not change the way the valves are used as a part of 
the Safety Injection System. A detailed Failure Modes and Effects 
Analysis were completed to confirm the system operation would meet 
the containment isolation design function. [The change does not add 
new or change existing plant equipment or affect the operating 
practices of the facility. Therefore, the change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.]
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The test [* * *] [change] is within existing regulatory 
requirements. The application of a closed loop outside of 
containment is appropriate and consistent with regulatory positions. 
The closed loop is applied to cold leg recirculation alignment of 
less than 8 hours when a run failure of a charging pump or RHR 
[residual heat removal] pump occurs. The probability of an 
HHSI\Charging Pump failure to run is 7.025E-06 per hour and for a 
LHSI [low-head safety injection]\RHR Pump is 7.689E-06 per hour. 
With containment integrity maintained within the allowable 
regulatory framework, there is no reduction in the margin of safety. 
[The change does not affect plant equipment or operating practices 
and therefore does not involve a significant reduction in 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. Hagood Hamilton, Jr., South Carolina 
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 
29218.
    NRC Branch Chief: Gloria Kulesa.

South Carolina Electric and Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1 
(VCSNS), Fairfield County, South Carolina

    Date of amendment request: October 12, 2011.
    Description of amendment request: The amendment request proposes 
changes to allow for a one time extension to the 10-year frequency of 
the VCSNS containment leakage rate test (e.g., integrated leak rate 
test (ILRT) or ``Type A test'') required by Technical Specification 
(TS) 6.8.4(g). The proposed change would permit the existing ILRT 
frequency to be extended from 10 years to approximately 10.9 years.
    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 with changes in brackets.

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident that has previously 
been evaluated?
    Response: No.
    The proposed [* * *] [change] involves a one-time extension to 
the current interval for Type A containment testing. The current 
test interval of 120 months (10 years) would be extended on a one-
time basis to no longer than approximately 130 months from the last 
Type A test. The proposed extension does not involve a physical 
change to the plant or a change in the manner in which the plant is 
operated or controlled. The containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the reactor containment itself and the testing requirements invoked 
to periodically demonstrate the integrity of the reactor containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident, and do not involve the prevention or identification of 
any precursors of an accident.
    Therefore, this proposed extension does not involve a 
significant increase in the probability of an accident previously 
evaluated nor does it create the possibility of a new or different 
kind of accident.
    The integrity of the reactor containment is subject to two types 
of failure mechanisms which can be categorized as (1) Activity based 
and (2) time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment itself combined with the containment inspections 
performed in accordance with the [American Society of Mechanical 
Engineers (ASME), Section Xl, Boiler and Pressure Vessel Code,] the 
Maintenance Rule, and Licensing commitments serve to provide a high 
degree of assurance that the containment will not degrade in a 
manner that is detectable only by a Type A test. Based on the above, 
the proposed extension does not involve a significant increase in 
the consequences of an accident previously evaluated.
    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 revision to the TS involves a one-time extension to 
the current interval for Type A containment testing. The reactor 
containment and the testing requirements invoked to periodically 
demonstrate the integrity of the reactor containment exist to ensure 
the plant's ability to mitigate the consequences of an accident and 
do not involve the prevention or identification of any precursors of 
an accident. The proposed TS change does not involve a physical 
change to the plant or the manner in which the plant is operated or 
controlled.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to the TS involves a one-time extension to 
the current interval for Type A containment testing. The proposed TS 
change does not involve a physical change to the plant or a change 
in the manner in which the plant is operated or controlled. The 
specific requirements and conditions of the Primary Containment Leak 
Rate Testing Program, as defined in the TS, exist to ensure that the 
degree of reactor containment structural integrity and leak-
tightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by TS 
is maintained. The proposed change involves only the extension of 
the interval between Type A containment leak rate tests. The 
proposed surveillance interval extension is bounded by the 15 month 
extension currently authorized within [Nuclear Energy Institute] NEI 
94-01, Revision 0. Type B and C containment leak rate tests will 
continue to be performed at the frequency currently required by TS. 
Industry experience supports the conclusion that Type B and C 
testing detects a large percentage of containment leakage paths and 
that the percentage of containment leakage paths that are detected 
only by Type A testing is small. The containment inspections 
performed in accordance with ASME, Section Xl and the Maintenance 
Rule serve to provide a high degree of assurance that the 
containment will not degrade in a manner that is detectable only by 
Type A testing. The combination of these factors ensures that the 
margin of safety that is in plant safety analysis is maintained. The 
design, operation, testing methods and

[[Page 77572]]

acceptance criteria for Type A, B, and C containment leakage tests 
specified in applicable codes and standards will continue to be met, 
with the acceptance of this proposed change, since these are not 
affected by changes to the Type A test interval.
    Therefore, the proposed TS change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. Hagood Hamilton, Jr., South Carolina 
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina 
29218.
    NRC Branch Chief: Gloria Kulesa.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment request: September 2, 2011.
    Description of amendment request: The amendments would revise a 
number of Technical Specification (TS) requirements, to impose similar 
restrictions on the movement of non-irradiated fuel assemblies to those 
currently in place for movement of irradiated fuel assemblies. The 
additional restrictions will limit the movement of all fuel assemblies 
over irradiated fuel assemblies in containment or in the fuel storage 
pool.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises Technical Specifications 
applicability wording regarding the movement of fuel assemblies in 
containment and the fuel storage pool at the San Onofre Nuclear 
Generating Station (SONGS) Units 2 and 3 to include the movement of 
both irradiated and non-irradiated fuel assemblies. The proposed 
applicability is more comprehensive than the current applicability.
    Expanding the applicability of the relevant Technical 
Specifications is necessary to account for updated fuel drop 
analyses which demonstrate that impacted spent fuel assemblies may 
be damaged. Consequently, movement of nonirradiated fuel assemblies 
could result in a Fuel Handling Accident that has radiological 
consequences. Changing the applicability of the relevant Technical 
Specifications does not affect the probability of a Fuel Handling 
Accident. The expanded applicability provides assurance that 
equipment designed to mitigate a Fuel Handling Accident is capable 
of performing its specified safety function.
    The dose consequences due to failure of two assemblies remain 
within the Regulatory Guide 1.183 and 10 CFR 50.67 acceptance 
criteria limits. The Exclusion Area Boundary (EAB), Low Population 
Zone (LPZ) and Control Room dose results and associated limits are 
presented below:

----------------------------------------------------------------------------------------------------------------
                                                                        Regulatory guide
       FHA inside fuel handing building         New analysis  FHA-     1.183 limit  (rem     10 CFR 50.67 limit
                                                  FHB  (rem TEDE)            TEDE)               (rem TEDE)
----------------------------------------------------------------------------------------------------------------
EAB..........................................                   1.7                  <=6.3                    25
LPZ..........................................                  <0.1                    6.3                  <=25
Control Room.................................                   0.6                  <=5                     <=5
----------------------------------------------------------------------------------------------------------------


----------------------------------------------------------------------------------------------------------------
                                                                        Regulatory guide
            FHA inside containment              New analysis FHA-IC     1.183 limit (rem     10 CFR 50.67 Limit
                                                    (rem TEDE)               TEDE)               (rem TEDE)
----------------------------------------------------------------------------------------------------------------
EAB..........................................                   1.7                  <=6.3                  <=25
LPZ..........................................                  <0.1                  <=6.3                  <=25
Control Room.................................                   0.6                  <=5                     <=5
----------------------------------------------------------------------------------------------------------------

    Consequently, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The updated fuel assembly drop analysis demonstrates that 
impacted fuel assemblies may be damaged as the result of a dropped 
fuel assembly. The existing SONGS Technical Specifications regarding 
movement of fuel assemblies are not applicable for movement of non-
irradiated fuel assemblies. A drop of a non-irradiated fuel assembly 
that has radiological consequences could occur during periods when 
equipment that would be required to mitigate those consequences is 
not required to be OPERABLE in accordance with the existing 
Technical Specifications.
    The proposed change to the Technical Specifications 
applicability language regarding the movement of fuel assemblies in 
containment and the fuel storage pool at SONGS Units 2 and 3 ensure 
that Limiting Conditions for Operation and appropriate Required 
Actions for required equipment are in effect during fuel movement. 
This provides assurance that any Fuel Handling Accident that may 
occur will remain within the initial assumptions of accident 
analyses.
    Consequently, there is no possibility of a new or different kind 
of accident due to the proposed change.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not affect protection criterion for 
plant equipment and will not reduce the margin of safety. By 
extending the Technical Specification applicability to the movement 
of non-irradiated fuel assemblies, the current margin of safety is 
maintained.
    Consequently, there is no significant reduction in a margin of 
safety due to the proposed change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Michael T. Markley.

Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Houston 
County, Alabama

    Date of amendment request: September 9, 2011.

[[Page 77573]]

    Description of amendment request: The proposed change would add 
Surveillance Requirement (SR) 3.3.1.14 to FNP TS Table 3.3.1-1, 
``Reactor Trip System [RTS] Instrumentation,'' Function 3, ``Power 
Range Neutron Flux High Positive Rate'' to the Technical 
Specifications. SR 3.3.1.14 requires verification that the RTS Response 
Time is within limits every 18 months on a Staggered Test Basis. 
Function 3 is the Power Range Neutron Flux High Positive Rate Trip 
(PFRT) function.
    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 to Farley Nuclear Plant (FNP) Technical 
Specification (TS) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' Table 3.3.1-1, ``Reactor Trip System 
Instrumentation,'' does not significantly increase the probability 
or consequences of an accident previously evaluated in the Update[d] 
Final Safety Analysis Report (UFSAR). The overall protection system 
performance will remain within the bounds of the accident analysis 
since there are no hardware changes. The design of the Reactor Trip 
System (RTS) instrumentation, specifically the power range neutron 
flux high positive rate trip (PFRT) function, will be unaffected. 
The reactor protection system will continue to function in a manner 
consistent with the plant design basis. All design, material, and 
construction standards, that were applicable prior to the request, 
are maintained.
    The proposed change imposes additional surveillance req