[Federal Register Volume 78, Number 151 (Tuesday, August 6, 2013)]
[Notices]
[Pages 47785-47795]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-18851]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0175]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
Background
Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission publish notice of any amendments issued, or proposed to be
issued and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from July 11, 2013, to July 23, 2013. The last
biweekly notice was published on July 23, 2013, (78 FR 44167).
ADDRESSES: You may submit comment by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0175. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422;
[[Page 47786]]
email: [email protected]. For technical questions, contact the
individual(s) listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: 3WFN, 06A44M, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2013-0175 when contacting the NRC
about the availability of information regarding this document. You may
access information related to this document, which the NRC possesses
and is publicly-available, by the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0175.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly-available documents online in the NRC
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in
ADAMS by performing a search on the document date and docket number.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2013-0175 in the subject line of your
comment submission, in order to ensure that the NRC is able to make
your comment submission available to the public in this docket.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment submissions into ADAMS.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Section 50.92 of Title 10 of the Code
of Federal Regulations (10 CFR), this means that operation of the
facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license or
combined license. Requests for a hearing and a petition for leave to
intervene shall be filed in accordance with the Commission's ``Agency
Rules of Practice and Procedure'' in 10 CFR Part 2. Interested
person(s) should consult a current copy of 10 CFR 2.309, which is
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The
NRC regulations are accessible electronically from the NRC Library on
the NRC's Web site at 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
[[Page 47787]]
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's E-Filing rule (72 FR 49139;
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to request (1) a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a request or
petition for hearing (even in instances in which the participant, or
its counsel or representative, already holds an NRC-issued digital ID
certificate). Based upon this information, the Secretary will establish
an electronic docket for the hearing in this proceeding if the
Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with the NRC guidance
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email to
[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:
[[Page 47788]]
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.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. With respect to copyrighted works, except
for limited excerpts that serve the purpose of the adjudicatory filings
and would constitute a Fair Use application, participants are requested
not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i)(iii).
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: May 28, 2013.
Description of amendment request: The amendment would modify safety
limits (SL) in Technical Specification (TS) 2.1.1, ``Reactor Core
SLs,'' to reduce the minimum reactor dome pressure associated with the
critical power correlation from 785 pounds per square inch gauge (psig)
to 685 psig. The RBS has evaluated the critical power correlation for
the General Electric Nuclear Energy advanced fuel designs (i.e., GE14
and GNF2 fuels) used at the facility which will allow for a lower-bound
pressure. The change will provide a greater pressure margin such that
the reactor remains above the proposed low SL of 685 psig in the event
of a Pressure Regulator Maximum Demand Open transient.
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.
Decreasing the reactor dome pressure limit in TS Safety Limits
2.1.1 for reactor Rated Thermal Power range effectively expands the
validity range for the GEXL 14 and GEXL 17 correlations and the
calculation of Minimum Critical Power Ratio Safety Limit (MCPR). The
MCPR rises during the pressure reduction following the scram that
terminates the Pressure Regulator Failure Open (PRFO) transient. Since
the change does not involve a modification of any plant hardware, the
probability and consequence of the PRFO transient are essentially
unchanged. The reduction in the reactor dome pressure safety limit from
785 psig to 685 psig provides greater margin to accommodate the
pressure reduction during the transient within the revised TS limit.
The proposed change will continue to support the validity range for
the GEXL correlations applied at RBS and the calculation of MCPR as
approved. The proposed TS revision involves no significant changes to
the operation of any systems or components in normal, accident or
transient operating conditions.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed reduction in the reactor dome pressure safety limit
from 785 psig to 685 psig is a change based upon previously approved
documents and does not involve changes to the plant hardware or its
operating characteristics. As a result, no new failure modes are being
introduced.
Therefore, the change does not introduce a new or different kind of
accident from those previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the plant
structures, systems, and components, and through the parameters for
safe operation and setpoints for the actuation of equipment relied upon
to respond to transients and design basis accidents. The proposed
change in reactor dome pressure enhances the safety margin, which
protects the fuel cladding integrity during a depressurization
transient, but does not change the requirements governing operation or
availability of safety equipment assumed to operate to preserve the
margin of safety. The change does not alter the behavior of plant
equipment, which remains unchanged. The available pressure range is
expanded by the change, thus offering greater margin for pressure
reduction during the transient.
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: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: April 5, 2013.
Description of amendment request: The proposed amendment would
revise the Pilgrim Technical Specifications (TSs) to reduce the reactor
steam dome pressure from 785 pounds per square inch, gauge (psig) to
685 psig specified in TS Reactor Core Safety Limits 2.1.1 and 2.1.2.
The proposed amendment is intended to address the potential to exceed
the low pressure TS safety limit associated with a pressure regulator
failure open (PRFO)--maximum
[[Page 47789]]
demand abnormal operation occurrence, as identified by General Electric
Nuclear Energy in its report, ``10 CFR 21 Reportable Condition
Notification: Potential to Exceed Low Pressure Technical Specification
Safety Limit,'' MFN 05-021, dated March 29, 2005.
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, along with the NRC's edits in
square brackets:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Decreasing the reactor dome pressure in Technical Specification
Safety Limits 2.1.1 and 2.1.2 for reactor Rated Thermal Power ranges
effectively expands the validity range for GEXL [GE critical quality-
boiling length correlation] and the calculation of Minimum Critical
Power Ratio Safety Limit (MCPR). MCPR rises during the pressure
reduction following the scram that terminates the PRFO transient. Since
the change does not involve a modification of any plant hardware, the
probability and consequence of the PRFO transient are essentially
unchanged. The reduction in the reactor dome pressure value in the
safety limit from 785 psig to 685 psig provides adequate margin to
accommodate the pressure reduction during the transient within the
revised TS limit.
The expanded GEXL correlation range supports Pilgrim's revised low
pressure safety limit of 685 psig. The proposed TS revision involves no
significant changes to the operation of any systems or components in
normal or accident or transient operating conditions.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed reduction in the reactor dome pressure value in the
safety limit from 785 psig to 685 psig reflects a wider range of
applicability for the GEXL correlation which is approved by the NRC for
fuels in use at Pilgrim and does not involve changes to the plant
hardware or its operating characteristics. As a result, no new failure
modes are being introduced.
Therefore, the [proposed] change does not [create the possibility
of] a new or different kind of accident from any [accident] previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the plant
structures, systems, and components, and through the parameters for
safe operation and setpoints for the actuation of equipment relied upon
to respond to transients and design basis accidents. The proposed
change in reactor dome pressure restores the safety margin, which
protects the fuel cladding integrity during a depressurization
transient, but does not change the requirements governing operation or
availability of safety equipment assumed to operate to preserve the
margin of safety. The change does not alter the behavior of plant
equipment, which remains unchanged. The reduction in the reactor dome
pressure value in the safety limit from 785 psig to 685 psig provides
adequate margin to accommodate the pressure reduction during the
transient within the revised TS limit.
Therefore, the proposed change does not involve a significant
reduction in [a] margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Acting Branch Chief: Robert Beall.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: May 14, 2013.
Description of amendment request: The proposed amendment would
revise the Vermont Yankee Technical Specifications (TSs) to reduce
reactor pressure associated with the fuel cladding integrity safety
limits (SLs) from 800 pounds per square inch, absolute (psia) to 700
psia in SLs 1.1.A and 1.1.B. The proposed change is intended to address
the potential to exceed the low pressure TS SL associated with a
pressure regulator failure-maximum demand open (PRFO) transient as
reported by General Electric Nuclear Energy in its Part 21
Communication, ``Potential to Exceed Low Pressure Technical
Specification Safety Limit,'' SC05-03, dated March 29, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the reactor pressure in Fuel Cladding
Integrity Safety Limits 1.1.A and 1.1.B does not alter the use of the
analytical methods used to determine the safety limits that have been
previously reviewed and approved by the NRC. The proposed change is in
accordance with NRC approved critical power correlation methodologies
and as such maintains required safety margins. The proposed change does
not adversely affect accident initiators or precursors nor does it
alter the design assumptions, conditions, or configuration of the
facility or the manner in which the plant is operated and maintained.
The proposed change does not alter or prevent the ability of
structures, systems, and components (SSCs) from performing their
intended function to mitigate the consequences of an initiating event
within the assumed acceptance limits. The proposed change does not
require any physical change to any plant SSCs nor does it require any
change in systems or plant operations. The proposed change is
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 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 which any plant systems perform a safety function. No new
accident scenarios, failure mechanisms, or limiting single failures are
introduced as a result of the proposed change.
The proposed change does not introduce any new accident precursors,
nor does it involve any physical plant alterations or changes in the
methods governing normal plant operation. Also, the change does not
impose any new or
[[Page 47790]]
different requirements or eliminate any existing requirements. The
change does not alter assumptions made in the safety analysis.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is related to confidence in the ability of the
fission product barriers (fuel cladding, reactor coolant system, and
primary containment) to perform their design functions during and
following postulated accidents. Evaluation of the 10 CFR Part 21 issue
by General Electric determined that the PRFO transient provides
additional margin to the Minimum Critical Power Ratio Safety Limit and
is not a threat to fuel cladding integrity.
The proposed change to Fuel Integrity Cladding Safety Limits 1.1.A
and 1.1.B is consistent with, and within the capabilities of the
applicable NRC approved critical power correlations, and thus continues
to ensure that valid critical power calculations are performed. No
setpoints at which protective actions are initiated are altered by the
proposed change. The proposed change does not alter the manner in which
the safety limits are determined. This change is consistent with plant
design and does not change the TS operability requirements; thus,
previously evaluated accidents are not affected by this proposed
change.
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: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Acting Branch Chief: Robert Beall.
Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie
Plant, Unit 1, St. Lucie County, Florida
Date of amendment request: May 10, 2013.
Description of amendment request: The amendment will revise the
Technical Specifications (TSs) to allow the use of M5[supreg] fuel rod
cladding material at St. Lucie Plant, Unit 1. The current acceptable
fuel rod cladding material is identified in TS 5.3.1, Reactor Core,
Fuel Assemblies. The proposed change would revise TS 5.3.1 to add
M5[supreg] to the approved fuel rod cladding materials and TS 6.9.1.11
to add Framatome (AREVA) topical report BAW-10240(P)(A), Revision 0,
``Incorporation of M5[supreg] Properties in Framatome ANP Approved
Methods,'' to the analytical methods used to determine the core
operating limits previously reviewed and approved by the NRC.
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 would allow the use of M5[supreg] fuel rod
cladding in the St. Lucie Unit 1 reactor. The topical report BAW-
10240(P)--A prepared by Framatome, currently known as AREVA, has been
approved by the NRC for use with M5[supreg] fuel cladding. The fuel
cladding itself is not an accident initiator and does not affect
accident probability. Use of M5[supreg] fuel cladding, which has
essentially the same properties as currently licensed Zircaloy, has
been shown to meet all 10 CFR 50.46 acceptance criteria and, therefore,
will not increase the consequences of an accident.
The proposed change to Technical Specification 6.9.1.11 (Core
Operating Limits Report (COLR)) enables the use of the appropriate
methodology to analyze accidents for cores containing fuel with
M5[supreg] cladding to ensure that the plant continues to meet
applicable design criteria and safety analysis acceptance criteria. The
proposed change to the list of NRC-approved methodologies listed in
Technical Specification 6.9.1.11 has no impact on plant operation and
configuration. The list of methodologies in Technical Specification
6.9.1.11 does not impact either the initiation of an accident or the
mitigation of its 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 kind of accident from any accident previously evaluated?
Response: No.
Use of M5[supreg] clad fuel will not result in changes in the
operation or configuration of the facility. The material properties of
M5[supreg] are similar to those of Zircaloy. Therefore, M5[supreg] fuel
rod cladding will perform similarly to those fabricated from Zircaloy,
thus precluding the possibility of the fuel becoming an accident
initiator and causing a new or different type of accident. The proposed
change to Technical Specification 5.3.1, to add M5[supreg] as a fuel
clad material, does not create any new accident initiators.
The proposed change to the list of NRC-approved methodologies
listed in Technical Specification 6.9.1.11, to add BAW-10240(P)--A, has
no impact on any plant configuration or system performance. There is no
change to the parameters within which the plant is normally operated,
and thus the possibility of a new or different type of accident is not
created.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not involve a significant reduction in the
margin of safety because it has been demonstrated that the material
properties of the M5[supreg] are not significantly different from those
of Zircaloy. The M5[supreg] is expected to perform similarly to
Zircaloy for all normal operating and accident scenarios, including
both loss of coolant accident (LOCA) and non-LOCA scenarios. For LOCA
scenarios, plant-specific LOCA analyses using M5[supreg] properties
demonstrate that the acceptance criteria of 10 CFR 50.46 have been
satisfied.
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: James Petro, Managing Attorney--Nuclear,
Florida Power & Light, P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Jessie F. Quichocho.
[[Page 47791]]
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County,
California
Date of amendment request: June 6, 2013.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.7.10, ``Control Room Ventilation
System (CRVS),'' and TS 5.6.5, ``Core Operating Limits Report (COLR),''
to incorporate editorial changes. Specifically, the proposed amendments
delete footnote (1), which expired on December 10, 2012, and is no
longer applicable, from TS 3.7.10 Condition A Completion Time, and
corrects inconsistent wording between TS 5.6.5a.4 and TS 3.2.1, between
TS 5.6.5a.5, and TS 3.2.2, and between TS 5.6.5a.9 and TS 3.4.1.
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 editorial changes do not involve any physical changes
to structures, systems or components. The proposed editorial change to
TS 3.7.10 deletes a footnote that is no longer applicable. The proposed
editorial changes to TS 5.6.5 correct administrative discrepancies in
the TS to provide consistency with the existing TS Sections 3.2.1,
3.2.2 and 3.4.1.
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 proposed editorial changes to TS 3.7.10 and TS 5.6.5 do not
involve an accident.
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 editorial changes to TS 3.7.10 and TS 5.6.5 do not
impact accident analyses, fission product barriers, or margin of
safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve 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, Unit 1,
Fairfield County, South Carolina
Date of amendment request: April 3, 2013.
Description of amendment request: The proposed amendment would add
an exception to Technical Specification 3.0.4 in Technical
Specification 3/4.7.6, Control Room Emergency Filtration System
(CREFS).
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 adds an exception to the provisions of
Specification 3.0.4 in Technical Specification 3/4.7.6, ``Control Room
Emergency Filtration System (CREFS)'' that was previously included in
this Technical Specification prior to Amendment 180. The proposed
change would allow entry into the applicable Modes of Technical
Specification 3/4.7.6 Actions b.1 and b.2 (Modes 5 and 6) while relying
on the actions. The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions, conditions,
or configuration of the facility. The proposed change does not alter or
prevent the ability of structures, systems, and components (SSCs) to
perform their intended function to mitigate the consequences of an
initiating event within the assumed acceptance limits. The proposed
change does not alter the Technical Specification Limiting Condition
for Operation, Applicability, or remedial Actions that provide for the
safe operation of the plant when the Limiting Condition for Operation
is not met. The Actions in Technical Specification 3/4.7.6 Action
statement b. continue to ensure the safe operation of the plant in the
same manner as before. In addition, the proposed change does not affect
the Surveillance Requirements of Technical Specification 3/4.7.6. As
such, the Surveillance Requirements continue to provide the same level
of assurance as before that the CREFS and control room boundary will
perform their required safety functions to mitigate the consequences of
events within the assumed acceptance limits.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change adds an exception to the provisions of
Specification 3.0.4 in Technical Specification 3/4.7.6, ``Control Room
Emergency Filtration System (CREFS)'' that was previously included in
this Technical Specification prior to Amendment 180. The proposed
change would allow entry into the applicable Modes of Technical
Specification 3/4.7.6 Actions b.1 and b.2 (Modes 5 and 6) while relying
on the actions. The proposed change does not alter the operability
requirements or remedial Actions of Technical Specification 3/4.7.6,
nor does the change affect the CREFS or control room boundary function
during accident conditions. The change does not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or a significant change in the methods governing
normal plant operation. The change does not alter assumptions made in
the applicable safety analyses. As such, the proposed change does not
impact the safety analyses assumptions and is consistent with current
plant operating practices.
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 adds an exception to the provisions of
Specification 3.0.4 in Technical Specification 3/4.7.6, ``Control Room
Emergency Filtration System (CREFS)''
[[Page 47792]]
that was previously included in this Technical Specification prior to
Amendment 180. The proposed change would allow entry into the
applicable Modes of Technical Specification 3/4.7.6 Actions b.1 and b.2
(Modes 5 and 6) while relying on the actions. 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 affected by the change. The
proposed change will not result in plant operation in a configuration
outside the design basis for an unacceptable period of time without
compensatory measures. The proposed change does not adversely affect
systems that respond to safely shutdown the plant and to maintain the
plant in a safe shutdown condition. As such, the CREFS and control room
boundary will continue to provide the same level of safety as before.
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: Robert J. Pascarelli.
Southern Nuclear Operating Company, Inc. Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
Date of amendment request: June 19, 2013.
Description of amendment request: The proposed changes would amend
Combined License numbers NPF-91 and NPF-92 for Vogtle Electric
Generating Plant Units 3 and 4 by departing from the plant-specific
design control document Tier 2 and Tier 2* material contained within
the updated final safety analysis report (UFSAR) related to the design
of structural wall modules used to construct containment internal
structures and portions of the auxiliary building. The proposed changes
would revise requirements for design spacing of shear studs and the
design of structural elements in order to address interferences and
obstructions other than wall openings.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No
The design function of the containment structural modules is to
support the reactor coolant system components and related piping
systems and equipment. The design functions of the affected structural
modules in the auxiliary building are to provide support and protection
for new and spent fuel and the equipment needed to support fuel
handling, cooling, and storage in the spent fuel racks, and to provide
support, protection, and separation for the seismic Category I
mechanical and electrical equipment located outside the containment
building.
The design function of the shear studs is to enable the concrete
and steel faceplates to act in a composite manner and transfer loads
into the concrete of the structural modules. The structural modules are
seismic Category I structures and are designed for dead, live, thermal,
pressure, safe shutdown earthquake loads, and loads due to postulated
pipe breaks. The loads and load combinations applicable to the
structural modules in the auxiliary building are the same as for the
containment internal structures except that there are no design basis
accident loadings due to the automatic depressurization system or
pressure loads due to pipe breaks. The proposed changes to the UFSAR
are to include types of interferences other than wall openings and
penetrations that may cause a change in the design spacing of shear
studs and the design and spacing of wall module trusses in a local
area. The proposed changes clarify that the stud spacing is specified
as a design value and add the tolerance for stud spacing. The revised
spacing including the tolerance continues to be in conformance with the
design and analysis requirements identified in the UFSAR. The proposed
changes also include clarification of a requirement for a complete
joint penetration weld. The thickness, geometry, and strength of the
structures are not adversely altered. The material of the steel plates
is not altered. The properties of the concrete included in the
structural modules are not altered. As a result, the design function of
the containment structural modules is not adversely affected by the
proposed change. There is no change to plant systems or the response of
systems to postulated accident conditions. There is no change to the
predicted radioactive releases due to postulated accident conditions.
The plant response to previously evaluated accidents or external events
is not adversely affected, nor does the change described create any new
accident precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the UFSAR acknowledge types of
interferences (other than wall openings and penetrations) that may
cause a change in the typical design spacing of shear studs and the
design and spacing of wall module trusses in a local area. The proposed
changes clarify that the stud spacing is specified as a design value
and provide the tolerance for stud spacing. The revised spacing,
including the tolerance, continues to be in conformance with the design
and analysis requirements identified in the UFSAR. Stud spacing and
sizing are evaluated to demonstrate that stud loadings and shear
transfer capability are within acceptable limits and that the
structural module acts in a composite manner. An additional proposed
change is to clarify a requirement for a complete joint penetration
weld. The thickness, geometry, and strength of the structures are not
adversely altered. The materials of the steel plates are not altered.
The properties of the concrete included in the structural modules are
not altered. The changes to the internal design of the structural
modules do not create any new accident precursors. As a result, the
design function of the modules is not adversely affected by the
proposed changes.
Therefore, the proposed amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
The criteria and requirements of American Concrete Institute (ACI)
349 and American Institute of Steel Construction (AISC) N690 provide a
margin of safety to structural failure. The design of the shear studs
and wall trusses for the structural wall modules conforms to applicable
criteria and requirements in ACI 349 and AISC N690 and, therefore,
maintain the margin of
[[Page 47793]]
safety. The proposed changes to the UFSAR acknowledge types of
interferences (other than wall openings and penetrations) that may
cause a change in the typical design spacing of shear studs and the
design and spacing of wall module trusses in a local area. The proposed
changes clarify that the stud spacing is specified as a design value
and add the tolerance for stud spacing. The revised spacing including
the tolerance continues to be in conformance with the design and
analysis requirements identified in the UFSAR. An additional proposed
change is to clarify a requirement for a complete joint penetration
weld. There is no change to the capacity of the weld or to the design
requirements of the modules. There is no change to the method of
evaluation from that used in the design basis calculations.
Therefore, the proposed amendment does not result in a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Blach & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence Burkhart.
ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power
Station (ZNPS), Units 1 and 2, Lake County, Illinois
Date of amendment request: June 18, 2012, and supplemented June 5,
2013.
Description of amendment request: The proposed amendments would
revise the Physical Security Plan associated with the transfer and
storage of spent fuel at the Independent Spent Fuel Storage
Installation (ISFSI).
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment, which incorporates ISFSI security
functions, does not reduce the ability of the Security organization to
prevent attempts of radiological sabotage and, therefore, does not
increase the probability or consequences of a radiological release
previously evaluated. The proposed ZNPS ISFSI Physical Security Plan
will not affect any important-to-safety systems or components, their
mode of operation or operating strategies. The changes have no effect
on accident initiators or mitigation.
Therefore, the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment incorporating ISFSI security functions does
not affect the operation of systems that are important-to-safety. The
ZNPS ISFSI Physical Security Plan amendment does not affect any of the
parameters or conditions that could contribute to the initiation of any
accident. No new accident scenarios are created as a result of the ZNPS
ISFSI Physical Security Plan. In addition, the design functions of
equipment important to safety are not altered as a result of the
proposed ZNPS ISFSI Physical Security Plan.
Therefore, the proposed ISFSI Security Plan will not create the
possibility of a new or different accident from any previously
evaluated.
3. Does the change involve a significant reduction in a margin of
safety?
Response: No.
Implementation of the proposed amendment incorporating ISFSI
security functions will not reduce a margin of safety as detailed in
the Technical Specifications, as there are no Technical Specification
requirements associated with the physical security system.
Specifically, the proposed ZNPS ISFSI Physical Security Plan does not
represent a change in initial conditions, system response time, or any
other parameter affecting the course of an accident analysis supporting
the Bases of any Technical Specification. The proposed amendment does
not reduce the effectiveness of any security/safeguards measures
currently in place at the ZNPS.
Therefore, the proposed ZNPS ISFSI Physical Security Plan will 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: Russ Workman, Deputy General Counsel,
EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT
84101.
NRC Branch Chief: Bruce Watson.
Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. Publicly available documents created or received at the
NRC are accessible electronically through the Agencywide Documents
Access and Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are
[[Page 47794]]
problems in accessing the documents located in ADAMS, contact the PDR's
Reference staff at 1-800-397-4209, 301-415-4737 or by email to
[email protected].
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: December 21, 2012.
Brief description of amendment:
The amendment revises Fermi 2 operating license to change its name
on the license to ``DTE Electric Company.'' This name change is purely
administrative in nature. Detroit Edison is a wholly owned subsidiary
of DTE Energy Company, and this name change is part of a set of name
changes of DTE Energy subsidiaries to conform their names to the
``DTE'' brand name. No other changes are contained within this
amendment. This change does not involve a transfer of control over or
of an interest in the license for Fermi 2.
Date of issuance: July 12, 2013.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 193.
Facility Operating License No. NPF-43: Amendment revised the
operating license.
Date of initial notice in Federal Register: March 4, 2013 (78 FR
14131).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 12, 2013.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina;
and Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and
2, Mecklenburg County, North Carolina
Date of amendment request: January 21, 2013.
Description of amendment request: The amendments revised the
divider barrier seal test coupons' tensile strength in Technical
Specification Surveillance Requirement 3.6.14.4 from ``> 39.7 psi'' to
``> 39.7 lbs.'' This change is an administrative change to correct an
error where the wrong units were used when Catawba and McGuire
converted to Standard Technical Specifications in 1998 using NUREG-
1431, Revision 1.
Date of issuance: July 16, 2013.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 270, 266, 270 and 250.
Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9 and
NPF-17: Amendments revised the licenses and the technical
specifications.
Date of initial notice in Federal Register: May 14, 2013 (78 FR
28251).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 16, 2013.
No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County,
Texas
Date of amendment request: July 12, 2012, as supplemented by letter
dated October 23, 2012.
Brief description of amendments: The amendments revised Technical
Specification (TS) 5.7.1, ``High Radiation Areas with Dose Rates not
Exceeding 1.0 rem [roentgen equivalent man]/hour at 30 Centimeters from
the Radiation Source or from any Surface Penetrated by the Radiation,''
and 5.7.2, ``High Radiation Areas with Dose Rates Greater than 1.0 rem/
hour at 30 Centimeters from the Radiation Source or from any Surface
Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter
from the Radiation Source or from any Surface Penetrated by the
Radiation,'' to allow entry into high radiation areas by personnel
continuously escorted by individuals qualified in radiation protection
procedures and to require a pre-job briefing prior to entry into such
areas. In addition, the amendment incorporates an editorial change to
TS Table 3.3.3-1, ``Post Accident Monitoring Instrumentation.'' The
typographical error in the title of TS Table 3.3.1-1 column ``CONDITION
REFERENCED FROM REQUIRED ACTION E.1,'' is corrected to read,
``CONDITION REFERENCED FROM REQUIRED ACTION D.1,'' to reflect that the
Required Actions for Condition D of TS 3.3.3, ``Post Accident
Monitoring (PAM) Instrumentation'' are listed in the table.
Date of issuance: July 11, 2013.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: Unit 1--159; Unit 2--159.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: November 13, 2012 (77
FR 67683).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 11, 2013.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of application for amendment: September 18, 2012.
Brief description of amendment: The amendment revises the MNGP
Technical Specifications (TS) Sections 3.1.6, ``Rod Pattern Control,''
and 3.3.2.1, ``Control Rod Block Instrumentation,'' to allow MNGP to
reference an optional Banked Position Withdrawal Sequence (BPWS)
shutdown sequence in the TS Bases. In addition, a footnote is revised
in TS Table 3.3.2.1-1, ``Control Rod Block Instrumentation,'' to allow
operators to bypass the rod worth minimizer if conditions for the
optional BPWS shutdown process are satisfied. The changes are
consistent with NRC-approved Technical Specifications Task Force (TSTF)
Improved Standard Technical Specifications Change Traveler, TSTF-476,
Revision 1, ``Improved BPWS Control Rod Insertion Process (NEDO-
33091).''
Date of issuance: July 15, 2013.
Effective date: This license amendment is effective as of the date
of its date of issuance and shall be implemented within 180 days after
start-up from the 2013 Refueling Outage.
Amendment No.: 173.
Renewed Facility Operating License No. DPR-22: Amendment revises
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 11, 2012.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 15, 2013.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Units 1 and 2, Salem County, New Jersey
Date of application for amendments: July 17, 2012, as supplemented
on January 28, 2013, and March 22, 2013.
Brief description of amendments: The amendment revised Salem
Nuclear Generating Station Technical Specification 3.7.6.1 (Unit 1) and
3.7.6 (Unit 2), ``Control Room Emergency Air Conditioning System,'' to
eliminate the separate action statements for securing an inoperable
Control Area Air Conditioning System and Control Room
[[Page 47795]]
Emergency Air Conditioning System isolation damper in the closed
position and entering the actions for an inoperable control room
envelope boundary.
Date of issuance: July 17, 2013.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 304 and 286.
Renewed Facility Operating License Nos. DPR-70 and DPR-75: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 2, 2013 (78 FR
19754).
The supplemental letter dated March 22, 2013, provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 17, 2013.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: August 14, 2012, as supplemented by
letters dated February 28, April 19, and June 24, 2013.
Brief description of amendment request: The amendments revised
Technical Specification (TS) 5.6.5, ``Core Operating Limits Report
(COLR),'' to reference and allow use of Westinghouse WCAP-16045-P-A,
Addendum 1-A, ``Qualification of the NEXUS Nuclear Data Methodology,''
(Reference 1 of Enclosure 1) to determine core operating limits. The
non-proprietary version is WCAP-16045-NP-A, Addendum 1-A (Reference 2
of Enclosure 1).
Date of issuance: July 17, 2013.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos: 191 and 187.
Facility Operating License Nos. NPF-2 and NPF-8: The amendment
revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: October 9, 2012 (77 FR
61440).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 17, 2013.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 29th day of July, 2013.
For The Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2013-18851 Filed 8-5-13; 8:45 am]
BILLING CODE 7590-01-P