[Federal Register Volume 76, Number 239 (Tuesday, December 13, 2011)]
[Notices]
[Pages 77565-77574]
From the Federal Register Online via the Government Publishing 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 http://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 http://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
[email protected].
Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and
Directives Branch (RADB), Office of Administration, Mail Stop: TWB-05-
B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
Fax comments to: RADB at (301) 492-3446.
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 http://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
[email protected].
Federal Rulemaking Web Site: Public comments and supporting
materials related to this notice can be found at http://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 http://www.nrc.gov/reading-rm/doc-collections/cfr/. If
a request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule (72 FR 49139,
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at (301) 415-1677, to request (1) A digital identification
(ID) certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a request or
petition for hearing (even in instances in which the participant, or
its counsel or representative, already holds an NRC-issued digital ID
certificate). Based upon this information, the Secretary will establish
an electronic docket for the hearing in this proceeding if the
Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC
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
[[Page 77567]]
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC 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 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 http://www.nrc.gov/site-help/e-submittals.html, by email at
[email protected], or by a toll-free call at 1-(866) 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information. With
respect to copyrighted works, except for limited excerpts that serve
the purpose of the adjudicatory filings and would constitute a Fair Use
application, participants are requested not to include copyrighted
materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Non-timely filings
will not be entertained absent a determination by the presiding officer
that the petition or request should be granted or the contentions
should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20874. Publicly available documents created or received at the NRC are
accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS, should contact the NRC's PDR Reference staff at 1-(800) 397-
4209, (301) 415-4737, or by email to [email protected].
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 requirements
to assure safety related structures, systems, and components (SSCs)
are verified to be consistent with the safety analysis and licensing
basis. In this specific case, a response time verification
requirement will be added to the PFRT function.
The proposed changes will not modify any system interface. The
proposed changes will not affect the probability of any event
initiators. There will be no degradation in the performance of, or
an increase in the number of challenges imposed on, safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance. The proposed change will not alter any
assumptions nor change any mitigation actions in the radiological
consequences evaluations in the UFSAR.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility or the manner in which
the plant is operated and maintained. The proposed changes do not
alter nor prevent the ability of SSCs from performing their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits. The proposed change is consistent
with the safety analyses assumptions and resultant consequences.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
There are no hardware changes nor are there any changes in the
method by which any safety related plant system performs its safety
function. This change will not affect the normal method of plant
operation nor change any operating parameters. No performance
requirements will be affected; however, the proposed change does
impose additional surveillance requirements. The additional
surveillance requirements are consistent with assumptions made in
the safety analyses and licensing basis.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this change. There will be no adverse effect or challenges
imposed on any safety-related system as a result of this change.
Therefore, the proposed change does not create the possibility
of a new or different 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 does not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Limits. There
will be no effect on the manner in which Safety Limits or Limiting
Conditions of Operations are determined nor will there be any effect
on those plant systems necessary to assure the accomplishment of
protection functions.
The safety analyses limits assumed in the accident analysis are
unchanged. The imposition of additional surveillance requirements
increases the margin of safety by assuring that the affected safety
analyses assumptions on equipment response time are verified on a
periodic frequency.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Branch Chief: Gloria J. Kulesa.
Notice of Issuance of Amendments to Facility Operating 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.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the NRC's 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 the Agencywide Documents Access and Management
System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the NRC's PDR
Reference staff at 1 (800) 397-4209, (301) 415-4737 or by email to
[email protected].
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: April 11, 2011.
Brief description of amendment: The amendment modified Technical
Specification (TS) 3.4.7, ``RCS [Reactor Coolant System] Leakage
Detection
[[Page 77574]]
Instrumentation,'' to define a new time limit for restoring inoperable
reactor coolant system (RCS) leakage detection instrumentation to
operable status; establish alternate methods of monitoring RCS leakage
when one or more required monitors are inoperable; and make TS Bases
changes which reflect the proposed changes and more accurately reflect
the contents of the facility design basis related to operability of the
RCS leakage detection instrumentation. These changes are consistent
with NRC-approved Revision 3 to Technical Specification Task Force
(TSTF) Change Traveler TSTF-514, ``Revise BWR [Boiling-Water Reactor]
Operability Requirements and Actions for RCS Leakage Instrumentation,''
as part of the consolidated line item improvement process.
Date of issuance: November 21, 2011.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 172.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: June 28, 2011 (76 FR
37847).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 21, 2011.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: November 10, 2010, as
supplemented by letter dated August 26, 2011.
Brief description of amendments: The change revised the PPL
Susquehanna, LLC (PPL) Unit 1 and Unit 2 Technical Specifications (TSs)
Surveillance Requirements (SRs) 3.4.3.1 ``Safety/Relief Valves (S/
RVs)'' to the lower tolerances from -3% to -5%. These changes would be
limited to the lower tolerances and does not affect the upper
tolerances. These changes only apply to the lower as-found tolerances
and not to the as-left tolerances, which will remain unchanged at
1% of the safety lift setpoint. The as-found tolerances are
used for determining past operability and to increase sample sizes for
S/RV testing should the upper tolerances be exceeded. There will be no
revision to the actual setpoints of the valves installed in the plant
due to this change.
Date of issuance: November 17, 2011.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment Nos.: 257 for Unit 1 and 237 for Unit 2.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Licenses and Technical Specifications.
Date of initial notice in Federal Register: February 22, 2011 (76
FR 9828).
The supplement dated August 26, 2011, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 17, 2011.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: December 21, 2010.
Brief description of amendments: The amendments revised Technical
Specification (TS) 5.3.1, ``FUEL ASSEMBLIES,'' by adding Optimized
ZIRLOTM fuel rods to the fuel matrix in addition to Zircaloy
or ZIRLOTM fuel rods that are currently in use. The
amendments also added a reference to an NRC-approved Westinghouse
Electric Company, LLC topical report regarding Optimized
ZIRLOTM to Section 6.9.1.6, ``Core Operating Limits Report
(COLR).''
Date of issuance: November 17, 2011.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: Unit 1--198; Unit 2--186.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: April 5, 2011 (76 FR
18804).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 17, 2011.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of application for amendment: August 10, 2011.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.8.1 ``AC [Alternating Current] Sources--
Operating.'' The change modified Surveillance Requirement (SR) Notes
associated with SR 3.8.1, SR 3.8.1.9, SR 3.8.1.10, SR 3.8.1.11, SR
3.8.1.13, SR 3.8.1.16, SR 3.8.1.18, and SR 3.8.1.19. The amendment
changed the WBN Unit 1 TS 3.8.1 to permit performance of the WBN Unit 2
integrated safeguards test without requiring WBN Unit 1 be shut down.
Date of issuance: November 22, 2011.
Effective date: As of the date of issuance and shall be implemented
no later than 30 days from date of issuance.
Amendment No.: 89.
Facility Operating License No. NPF-90: Amendment revised the
License and TSs.
Date of initial notice in Federal Register: September 20, 2011 (76
FR 58306).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 22, 2011.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 2nd day of December 2011.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2011-31901 Filed 12-12-11; 8:45 am]
BILLING CODE 7590-01-P