[Federal Register Volume 78, Number 73 (Tuesday, April 16, 2013)]
[Notices]
[Pages 22563-22576]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-08756]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0069]
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 March 21 to April 3, 2013. The last biweekly
notice was published on April 2, 2013 (78 FR 19746).
ADDRESSES: You may access information and comment submissions related
to this document, which the NRC possesses and is publicly-available, by
searching on http://www.regulations.gov under Docket ID NRC-2013-0069.
You may submit comments by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket NRC-2013-0069. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected].
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
Fax comments to: RADB at 301-492-3446.
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-2013-0069 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-2013-0069.
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-2013-0069 in the subject line of your
comment submission, in order to ensure that the NRC is able to make
your comment submission available to the public in this docket.
The NRC cautions you not to include identifying or contact
information in comment submissions that you do not want to be publicly
disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS, and the NRC does not edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information in their comment submissions
that they do not want to be publicly disclosed. Your request should
state that the NRC will not edit comment submissions to remove such
information before making the comment submissions available to the
public or entering the comment submissions into ADAMS.
[[Page 22564]]
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 Sec. 50.92 of Title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated; or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
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 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
[[Page 22565]]
documents and access the E-Submittal server for any proceeding in which
it is participating; and (2) advise the Secretary that the participant
will be submitting a request or petition for hearing (even in instances
in which the participant, or its counsel or representative, already
holds an NRC-issued digital ID certificate). Based upon this
information, the Secretary will establish an electronic docket for the
hearing in this proceeding if the Secretary has not already established
an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with the NRC guidance
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC's Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email at
[email protected], or by a toll-free call at 1-866 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants. Filing is considered
complete by first-class mail as of the time of deposit in the mail, or
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting 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 following three
factors in 10 CFR 2.309(c)(1): (i) The information upon which the
filing is based was not previously available; (ii) the information upon
which the filing is based is materially different from information
previously available; and (iii) the filing has been submitted in a
timely fashion based on the availability of the subsequent information.
For further details with respect to this license amendment
application, see the application for amendment, which is available for
public inspection at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS should contact the NRC's PDR Reference staff at 1-800-397-4209,
301-415-4737, or by email to [email protected].
Detroit Edison, Docket No. 50-341, Fermi 2, Monroe County, Michigan
Date of amendment request: January 11, 2013.
Description of amendment request: The proposed amendment would
revise Fermi 2 Technical Specifications (TS) to
[[Page 22566]]
incorporate the NRC-approved TSTF-423, Revision 1. The proposed
amendment would modify TS to risk-inform requirements regarding
selected Required Action end states by incorporating the boiling water
reactor (BWR) owner's group (BWROG) approved Topical Report NEDC-32988-
A, Revision 2, ``Technical Justification to Support Risk-Informed
Modification to Selected Required Action End States for BWR Plants.''
Additionally, the proposed amendment would modify the TS Required
Actions with a Note prohibiting the use of limiting condition for
operation (LCO) 3.0.4.a when entering the preferred end state (Mode 3)
on startup.
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 allows a change to certain required end
states when the TS Completion Times for remaining in power operation
will be exceeded. Most of the requested technical specification (TS)
changes are to permit an end state of hot shutdown (Mode 3) rather
than an end state of cold shutdown (Mode 4) contained in the current
TS. The request was limited to: (1) Those end states where entry
into the shutdown mode is for a short interval, (2) entry is
initiated by inoperability of a single train of equipment or a
restriction on a plant operational parameter, unless otherwise
stated in the applicable TS, and (3) the primary purpose is to
correct the initiating condition and return to power operation as
soon as is practical. Risk insights from both the qualitative and
quantitative risk assessments were used in specific TS assessments.
Such assessments are documented in Section 6 of topical report NEDC-
32988-A, Revision 2, ``Technical Justification to Support Risk
Informed Modification to Selected Required Action End States for BWR
Plants.'' They provide an integrated discussion of deterministic and
probabilistic issues, focusing on specific TSs, which are used to
support the proposed TS end state and associated restrictions. The
NRC staff finds that the risk insights support the conclusions of
the specific TS assessments. Therefore, the probability of an
accident previously evaluated is not significantly increased, if at
all. The consequences of an accident after adopting TSTF-423 are no
different than the consequences of an accident prior to adopting
TSTF-423. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns.
Therefore, 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 does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
If risk is assessed and managed, allowing a change to certain
required end states when the TS Completion Times for remaining in
power operation are exceeded (i.e., entry into hot shutdown rather
than cold shutdown to repair equipment) will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change and the commitment by the licensee to adhere to the guidance
in TSTF-IG-05-02, ``Implementation Guidance for TSTF-423, Revision
1, `Technical Specifications End States, NEDC-32988-A,'' will
further minimize possible concerns.
Thus, based on the above, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change allows, for some systems, entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed. The BWROG's risk assessment approach is
comprehensive and follows NRC staff guidance as documented in
Regulatory Guides (RG) 1.174 and 1.177. In addition, the analyses
show that the criteria of the three-tiered approach for allowing TS
changes are met. The risk impact of the proposed TS changes was
assessed following the three-tiered approach recommended in RG
1.177. A risk assessment was performed to justify the proposed TS
changes. The net change to the margin of safety is insignificant.
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: Bruce R. Masters, DTE Energy, General
Counsel--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
NRC Branch Chief: Robert D. Carlson.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3 (ONS1, ONS2, and ONS3),
Oconee County, South Carolina
Date of amendment request: October 30, 2012.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TSs) to allow operation of a
reverse osmosis system during normal plant operation to purify the
water in the borated water storage tanks and the spent fuel pools.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change requests NRC's approval of design features
and controls that will be used to ensure that periodic limited
operation of a Reverse Osmosis (RO) System during Unit operation
does not significantly impact the Borated Water Storage Tank (BWST)
or Spent Fuel Pool (SFP) function or other plant equipment. The
proposed change also requests NRC to approve proposed Technical
Specification (TS) requirements that will impose operating
restrictions and isolation requirements on the RO System. Duke
Energy evaluated the effect of potential failures, identified
precautionary measures that must be taken before and during RO
System operation, and identified specific required operator actions
to protect affected structures, systems, and components (SSCs)
important to safety.
The new high energy piping and non-seismic piping being
installed for the RO System is non-QA1 and is postulated to fail and
cause an Auxiliary Building flood. Duke Energy determined that
adequate time is available to isolate the flood source (BWST or SFP)
prior to affecting SSCs important to safety.
The existing Auxiliary Building Flood evaluation postulates a
single break in the non-seismic piping occurring in a seismic event.
The addition of the RO System will not increase the probability of a
seismic event. The existing postulated source of the pipe break in
the Auxiliary Building is due to the piping not being seismically
designed. The new RO System piping is considered a potential source
of a single pipe break for the same reason. The new non-seismic RO
System piping is of similar quality as the existing non-seismic
piping and is no more likely to fail than the existing piping. As
such, the addition of new non-seismic piping does not significantly
increase the probability of occurrence of an Auxiliary Building
flood due to a single pipe break. An Auxiliary Building flood due to
a non-seismic RO
[[Page 22567]]
System pipe break does not increase the consequences of the flood
since the new non-seismic pipe break is bounded by the Auxiliary
Building flood caused by existing non-seismic pipe breaks.
Procedural controls will ensure that the boron concentration
does not go below the TS limit as a result of water returned from
the RO System with lower boron concentration. Thus, no adverse
effects from decreased boron concentration will occur.
The RO System takes suction from the top of the SFP to protect
SFP inventory. Plant procedures will prohibit the use of the RO
System for the Units 1 & 2 SFP during the time period directly after
an outage that requires the Units 1 & 2 SFP level to be maintained
higher than the TS Limiting Condition for Operation (LCO) 3.7.11
level requirement. The higher level is required to support TS LCO
3.10.1 requirements for Standby Shutdown Facility (SSF) Reactor
Coolant (RC) Makeup System operability (due to the additional decay
heat from the recently offloaded spent fuel). Plant procedures will
also specify the siphon be broken during this time period so the SFP
water above the RO suction point cannot be siphoned off if the RO
piping breaks. The proposed change does not impact the fuel
assemblies, the movement of fuel, or the movement of fuel shipping
casks. The SFP boron concentration, level, and temperature limits
will not be outside of required parameters due to restrictions/
requirements on the system's operation. In addition, the proposed
new TS will require the siphon be broken during movement of
irradiated fuel assemblies in the SFP or movement of cask over the
SFP. Therefore, RO System operation cannot occur during these
activities, effectively eliminating a Fuel Handling Accidents (FHA)
from occurring while the RO System is in operation.
The BWST is used for mitigation of Steam Generator Tube Rupture
(SGTR), Main Steam Line Break (MSLB), and Loss of Coolant Accidents
(LOCAs). The SGTR and MSLB are bounded by the small break (SBLOCA)
analyses with respect to the performance requirements for the High
Pressure Injection (HPI) System. In the normal mode of Unit
operation, the BWST is not an accident initiator. The SFP is
evaluated to maintain acceptable criticality margin for all abnormal
and accident conditions including FHAs and cask drop accidents. Both
the BWST and SFP are specified by TS requirements to have minimum
levels/volumes and boron concentrations. The BWST also has TS
requirements for temperature. Prior to RO System operation,
procedures will require the minimum required initial boron
concentration and initial level/volume to be adjusted. Additionally,
they will require the RO System to be operated for a specified
maximum time period before readjusting volume and boron
concentration prior to another RO session. This ensures that the TS
specified boron concentration and level/volume limits for both the
SFP and the BWST are not exceeded during RO System operation. Thus,
the design functions of the BWST and the SFP will continue to be met
during RO System operation.
Since the BWST and SFP will still have TS boron concentration
and level/volume requirements and the RO System will be isolated
prior to increasing radiation levels preventing access to the
isolation valve, the mitigation of a LOCA or FHA does not result in
an increase in dose consequence. Since the design basis LOCA
analysis for Oconee assumes 5 gpm back-leakage from the Reactor
Building sump to the BWST, the Emergency Operating Procedure will
require the RO System to be isolated from the BWST prior to switch
over to the recirculation phase. The proposed TS will require the RO
system to be isolated (by breaking the siphon) from the SFPs during
fuel handling activities and will require the seismic boundary valve
between the BWST and RO System to be OPERABLE in MODES 1, 2, 3, and
4.
The additional controls imposed by the proposed Technical
Specifications (TSs) will provide additional assurance that
isolation valves and operating restrictions credited to eliminate
the need to analyze new release pathways introduced by the RO system
will be in place.
Therefore, installation and operation of the RO System during
Unit operation and the proposed TS imposing operating restrictions
do not significantly increase 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 RO System adds non-seismic piping in the Auxiliary Building.
However, the break of a single non-seismic pipe in the Auxiliary
Building has already been postulated as an event in the licensing
basis. The RO System also does not create the possibility of a
seismic event concurrent with a LOCA since a seismic event is a
natural phenomena event. The RO System does not adversely affect the
Reactor Coolant System pressure boundary. The suction to the RO
System, when using the system for BWST purification, contains a
normally closed manual seismic boundary valve so the seismic design
criteria is met for separation of seismic/non-seismic piping
boundaries.
Duke Energy also evaluated potential releases of radioactive
liquid to the environment due to RO System piping failures. Design
features, controls imposed by the proposed TS, and procedural
controls will preclude release of radioactive materials outside the
Auxiliary Building by ensuring the RO System will be isolated when
required.
The SFP suction line is designed such that the SFP water level
will not go below TS required levels, thus the fuel assemblies will
have the TS required water level over them. Procedural controls will
restrict the use of the RO System and require breaking vacuum on the
Units 1 & 2 SFP suction line when the SSF conditions require the SFP
level be raised to support SSF RC Makeup System operability. Thus,
the SFP water level will not be reduced below required water levels
for these conditions. RO System operating restrictions will prevent
reducing the SFP boron concentration below TS limits.
Since the BWST and SFP will still have TS boron concentration
and level/volume requirements and the RO System will be isolated
prior to increasing radiation levels preventing access to the
isolation valve, the mitigation of a LOCA or FHA does not result in
an increase in dose consequence. Since the design basis LOCA
analysis for Oconee assumes 5 gpm back-leakage from the Reactor
Building sump to the BWST, the Emergency Operating Procedure will
require the RO System to be isolated from the BWST prior to switch
over to the recirculation phase. The proposed TS will require the RO
system to be isolated (by breaking the siphon) from the SFPs prior
to movement of irradiated fuel assemblies in the SFP or movement of
cask over the SFP and will require the seismic boundary valve
between the BWST and RO System to be OPERABLE in MODES 1, 2, 3, and
4.
The additional controls imposed by the proposed TSs will provide
additional assurance that isolation valves and operating
restrictions credited to eliminate the need to analyze new release
pathways introduced by the RO system will be in place.
Therefore, operation of the RO System during Unit operation will
not create the possibility of a new or different kind of accident
from any kind of accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The RO System adds non-seismic piping in the Auxiliary Building.
Duke Energy evaluated the impact of RO System operation on SSCs
important to safety and determined that the proposed TS controls and
procedural controls will ensure that TS limits for SFP and BWST
volume, temperature, and boron concentration will continue to be met
during RO operation. For the BWST, these controls will ensure the TS
minimum BWST boron concentration and level are available to mitigate
the consequences of a small break LOCA or a large break LOCA. For
the SFP, these controls ensure the assumptions of the fuel handling
and cask drop accident analyses are preserved. Additionally, the
failure of non-seismic RO System piping will not significantly
impact SSCs important to safety. Oconee's licensing basis does not
assume a design basis event occurs simultaneously with a seismic
event. The proposed change does not significantly impact the
condition or performance of SSCs relied upon for accident
mitigation. This change does not alter the existing TS allowable
values or analytical limits. The existing operating margin between
Unit conditions and actual Unit setpoints is not significantly
reduced due to these changes. The assumptions and results in any
safety analyses are not impacted. Therefore, operation of the RO
System during Unit operation 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.
[[Page 22568]]
Attorney for licensee: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street-EC07H, Charlotte, NC
28202-1802.
NRC Branch Chief: Robert J. Pascarelli.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of amendment request: February 22, 2013.
Description of amendment request: The proposed amendments would
revise the Technical Specification curves for pressure and temperature
limits on the reactor coolant system, and limits on heatup and cooldown
rates.
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 amendment replaces the current Oconee Nuclear
Station (ONS) Units 1, 2, and 3 pressure/temperature (P-T) limit
curves applicable to 33 effective full power years (EFPY) in
Technical Specification (TS) 3.4.3 with new P-T limit curves
applicable to 54 EFPY. The proposed changes also revise the Reactor
Coolant System heatup and cooldown rates and allowable reactor
coolant pump combinations of TS Tables 3.4.3-1 and 3.4.3-2. The
pressure-temperature (P-T) limit curves in the TSs were
conservatively generated in accordance with fracture toughness
requirements of ASME Code Section XI, Appendix G, and the minimum
pressure and temperature requirements as listed in Table 1 of 10 CFR
Part 50, Appendix G. The proposed changes do not impact the
capability of the reactor coolant pressure boundary (i.e., no change
in operating pressure, materials, seismic loading, etc.).
Therefore, the proposed changes do not increase the potential
for the occurrence of a loss of coolant accident (LOCA). The changes
do not modify the reactor coolant system pressure boundary, nor make
any physical changes to the facility design, material, or
construction standards. The probability of any design basis accident
(DBA) is not affected by this change, nor are the consequences of
any DBA affected by this change. The proposed P-T limits, heatup and
cooldown rates and allowable operating reactor coolant pump
combinations are not considered to be an initiator or contributor to
any accident analysis addressed in the ONS Updated Final Safety
Analyses Report (UFSAR).
The proposed changes will not impact assumptions and conditions
previously used in the radiological consequence evaluations nor
affect the mitigation of these consequences due to an accident
described in the UFSAR. Also, the proposed changes will not impact a
plant system such that previously analyzed SSCs might be more likely
to fail. The initiating conditions and assumptions for accidents
described in the UFSAR remain as analyzed.
Therefore, the probability or consequences of an accident
previously evaluated is not significantly increased.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The requirements for P-T limit curves have been in place since
the beginning of plant operation. The revised curves are based on a
later edition to Section XI of the ASME Code that incorporates
current industry standards for P-T curves. The revised curves are
based on reactor vessel irradiation damage predictions using
Regulatory Guide 1.99 methodology. No new failure modes are
identified nor are any SSCs required to be operated outside the
design bases.
Therefore, the possibility of a new or different kind of
accident from any kind of accident previously evaluated is not
created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed P-T curves continue to maintain the safety margins
of 10 CFR Part 50, Appendix G, by defining the limits of operation
which prevent non-ductile failure of the reactor pressure vessel.
Analyses have demonstrated that the fracture toughness requirements
are satisfied and that conservative operating restrictions are
maintained for the purpose of low temperature overpressure
protection. The P-T limit curves provide assurance that the RCS
pressure boundary will behave in a ductile manner and that the
probability of a rapidly propagating fracture is minimized.
Therefore, this request does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Deputy General Counsel,
Duke Energy Corporation, 526 South Church Street-EC07H, Charlotte, NC
28202-1802.
NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and 2,
Ogle County, Illinois
Date of amendment request: December 21, 2012.
Description of amendment request: The proposed amendment would
Revise Technical Specifications (TS) 3.3.6, ``Containment Ventilation
Isolation Instrumentation.'' Specifically, this amendment request
proposes to revise Footnote (b) of TS Table 3.3.6-1, ``Containment
Ventilation Isolation Instrumentation,'' which specifies the
``Containment Radiation--High'' trip setpoint for two containment area
radiation monitors (i.e., 1(2) RE-AR011 and 1(2) RE-AR012). The
proposed changes would revise the ``Containment Radiation--High'' trip
setpoint from the current, overly conservative value (i.e., a
submersion dose rate of less than or equal to 10 mRhr in the
containment building), to less than or equal to 2 times the containment
building background radiation reading at rated thermal power, which is
consistent with NUREG-1431, ``Standard Technical Specifications,
Westinghouse Plants.'' Upon reaching the ``Containment Radiation--
High'' setpoint, these area radiation monitors provide an isolation
signal to the containment normal purge, mini purge and post-LOCA (Loss
of Coolant Accident) systems' containment isolation valves.
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 containment ventilation isolation radiation monitors serve
two primary functions, they:
a. act as backup to the SI [safety injection] signal to ensure
closing of the purge valves; and
b. are the primary means for automatically isolating containment
in the event of a fuel handling accident in containment.
Upon sensing a high radiation condition in containment, these
area radiation monitors provide an isolation signal to the
containment normal purge, mini purge and post- LOCA systems
containment isolation valves (i.e., a containment ventilation
isolation signal).
The accidents that could potentially be impacted by the proposed
change were evaluated; specifically the Loss of Coolant Accident
(LOCA), Control Rod Ejection Accident (CREA) and Fuel Handling
Accident (FHA) in Containment. The proposed change has no impact on
probability of these accidents occurring as the subject containment
radiation area monitors detect a high radiation condition resulting
from these accidents. The radiation monitors do not initiate any
accidents or transients. Changing the ``Containment
[[Page 22569]]
Radiation--High'' trip setpoint from ``<=10 mR/hr in the containment
building,'' to ``<=2 times the containment building background
radiation reading at rated thermal power'' only affects the point
(i.e., the radiation level in containment) at which a containment
ventilation isolation signal would be generated. The requested
change does not involve any physical plant modifications or
operational changes that could adversely affect system reliability
or performance of the radiation monitors, or that could affect the
probability of operator error.
The requested change does not affect any postulated accident
precursors and therefore, will not affect the probability of an
accident previously evaluated.
The proposed change was evaluated to determine the impact on the
dose consequences of the LOCA, CREA, or FHA in containment. The
evaluation assumed a containment purge was in progress at the onset
of the subject accidents and showed that the proposed change in the
containment radiation monitors' setpoint had no effect on the purge
valve isolation time. With regard to the LOCA and CREA, the safety
analysis assumes a prompt purge valve isolation time (i.e.,
approximately 60 seconds) that significantly bounds the radiation
monitor sensing and response time, and actual valve design closure
time (i.e., a total of approximately 7 seconds). The radiation
monitor setpoint is not considered in the safety analysis and any
dose contribution associated with the containment purge, due to the
proposed change in setpoint, was shown to be unaffected; therefore,
the proposed change has no impact on the already insignificant dose
contribution attributed to a containment purge during an accident of
less than one mrem.
The dose consequences associated with the FHA in containment are
also not impacted by the proposed change in containment radiation
monitor setpoint. The existing dose consequences resulting from a
FHA with moving non-RECENTLY IRRADIATED FUEL (i.e., fuel moved more
than 48 hours after reactor shutdown) conservatively assume the
containment purge valves remain open throughout the event;
therefore, a change in the isolation setpoint does not impact the
results of this analysis. With regard to movement of RECENTLY
IRRADIATED FUEL (i.e., fuel moved less then 48 hours after reactor
shutdown), EGC's [Exelon Generation Company] proposal deletes TS LCO
[limiting condition for operation] 3.9.4.c.2 which allowed the
containment purge valves to be open provided the containment
radiation isolation system is OPERABLE. Deletion of TS LCO 3.9.4.c.2
ensures that the containment purge valves are in the closed position
when moving RECENTLY IRRADIATED FUEL, thus removing dependence on
the containment radiation isolation system and associated radiation
monitor setpoint from the FHA dose consequences.
The four other additional TS changes associated with the
deletion of LCO 3.9.4, Item c.2, proposed for consistency (i.e.,
deleting a NOTE regarding MODE applicability, deleting a CONDITION
related only to LCO 3.9.4.c.2, deleting a footnote regarding MODE
applicability; and deleting two surveillances related to LCO
3.9.4.c.2), also have no affect on either the probability or
consequences of an accident previously evaluated.
Based on the above discussion, 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 changes do not result in a change to the design of
the Containment Ventilation Isolation System or the manner in which
the system operates or provides plant protection. The containment
radiation monitors will sense radiation levels in the same way and
will respond in the same manner when the setpoint is exceeded. The
change in the ``Containment Radiation--High'' setpoint does not
create a new failure mode for the associated containment radiation
monitors or for any other plant equipment. The deletion of LCO
3.9.4, Item c.2, in support of the setpoint change during refueling
operations, is more conservative than the current allowances and
actually eliminates a potential failure mode for the assumed open
containment ventilation isolation valves as the proposed deletion of
LCO 3.9.4, Item c.2 would require the valves to be closed prior to
moving RECENTLY IRRADIATED FUEL.
The changes do not result in the creation of any new accident
precursors, the creation of any changes to the existing accident
scenarios, nor do they create any new or different accident
scenarios. Subsequently, the accidents defined in the UFSAR [updated
final safety analysis report] continue to represent the credible
spectrum of events to be analyzed which determine safe plant
operation.
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 analysis methodologies used in the subject safety analyses
are not modified as a result of the proposed TS changes to the
``Containment Radiation--High'' trip setpoint or the deletion of LCO
3.9.4, Item c.2, or any of the other four associated TS changes.
Although the ``Containment Radiation--High'' trip setpoint is being
increased, the increase in response time to a high radiation
condition in containment, when compared to the current setpoint, is
negligible due to the projected prompt rise in containment radiation
level upon initiation of a LOCA. The dose consequences and resultant
margin of safety to the regulatory acceptance limits, due to
revising the ``Containment Radiation--High'' setpoint to <= 2 times
the containment building background radiation reading at rated
thermal power, was shown to be unaffected for normal at-power
containment releases; have a negligible impact on the associated
LOCA and CREA accident dose consequences; and have no impact on the
FHA when moving RECENTLY IRRADIATED FUEL. Therefore, the proposed
changes do not impact any analysis margins.
The proposed changes do not alter the manner in which the safety
limits, limiting safety system setpoints, or limiting conditions for
operation are determined. The current safety analyses remain
bounding since their conclusions are not affected by the proposed
changes. The safety systems credited in the safety analyses will
continue to be available to perform their mitigation functions. All
protection signals credited as the primary or secondary accident
mitigating functions, and all operator actions credited in the
accident analyses remain the same. The proposed changes will not
result in plant operation in a configuration outside the design
basis.
Based on the above information, the proposed change does not
result in a significant reduction in the margin of safety.
Based on the above evaluation, EGC concludes that the proposed
amendments do not involve a significant hazards consideration under
the standards set forth in 10 CFR 50.92, paragraph (c), and,
accordingly, a finding of no significant hazards consideration is
justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Acting Branch Chief: Jeremy S. Bowen.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: January 29, 2013.
Description of amendment request: The license amendment request
proposes to remove completed and satisfied license conditions and to
correct inadvertent errors and incorrect references.
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 amendments do not change or modify the fuel, fuel
handling processes, fuel storage racks, number of fuel assemblies
[[Page 22570]]
that may be stored in the spent fuel pool (SFP), decay heat
generation rate, or the spent fuel pool cooling and cleanup system.
The proposed amendments only limit crediting of burnable absorbers
in the spent fuel pool to Integrated Fuel Burnable Absorber (IFBA)
rods that were specifically addressed in the currently approved
criticality analysis ([Westinghouse Commercial Atomic Power report]
WCAP-1 7094-P, Revision 3). The removal of the phrase ``or an
equivalent amount of another burnable absorber'' eliminates the
possibility of crediting a burnable absorber other than IFBA for
storage of spent fuel assemblies in the spent fuel pool without
prior NRC's approval. The deletion of the license condition
associated with the Boraflex Remedy is editorial as it is no longer
applicable. The proposed amendments do not affect the ability of the
BAST [boric acid storage tank] to perform its function or the
ability of the CREVS [control room emergency ventilation system] to
perform its function. These latter proposed TS [technical
specification] changes correct inadvertent errors and are consistent
with the stated intent of original license submittals or delete
license conditions that are no longer applicable or that have been
fully satisfied.
The proposed amendments do not cause any physical change to the
existing spent fuel storage configuration, fuel makeup, RCS [reactor
coolant system] pressure boundary, reactor containment, or plant
systems. The proposed amendments do not affect any precursors to any
accident previously evaluated or do not affect any known mitigation
equipment or strategies.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendments do not change or modify the fuel, fuel
handling processes, fuel racks, number of fuel assemblies that may
be stored in the pool, decay heat generation rate, or the spent fuel
pool cooling and cleanup system. The proposed amendments do not
result in any changes to spent fuel or to fuel storage
configurations. The removal of the phrase ``or an equivalent amount
of another burnable absorber'' eliminates the possibility of
crediting a burnable absorber other than IFBA for storage of spent
fuel assemblies in the spent fuel pool without prior NRC approval.
The proposed amendments do not affect the ability of the BAST to
perform its function or the ability of the CREVS to perform its
function. These latter proposed TS changes correct inadvertent
errors and are consistent with the stated intent of the original
license submittals, delete license conditions that are no longer
applicable or have been fully satisfied.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Response: No.
The proposed amendments do not change or modify the fuel, fuel
handling processes, fuel racks, number of fuel assemblies that may
be stored in the pool, decay heat generation rate, or the spent fuel
pool cooling and cleanup system. Therefore, the proposed amendments
have no impact to the existing margin of safety for subcriticality
required by 10 CFR 50.68(b)(4). The other proposed OL [operating
license] & TS changes correct inadvertent errors and are consistent
with the stated intent of the original license submittals or delete
license conditions that are no longer applicable or have been fully
satisfied.
Therefore, the proposed amendments do 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 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.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station,
Nemaha County, Nebraska
Date of amendment request: June 25, 2012.
Description of amendment request: The amendment would revise the
description of the Fuel Handling Accident (FHA) in Section XIV-6.4 of
the Cooper Nuclear Station (CNS) Updated Safety Analysis Report (USAR).
The revised USAR FHA description is based on changes to the Design
Basis Accident FHA dose calculation, to reflect a 24-month fuel cycle
source term using a Global Nuclear Fuels (GNF) 10 x 10 fuel array,
reduce the bounding Radial Peaking Factor, and revise the total
effective dose equivalent (TEDE) contribution to consider the shine
contribution from external sources.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The analyses changes described by this proposed change to the
USAR are not initiators to events, and, therefore, do not involve
the probability of an accident. The changes to the FHA calculation
for radiological dose following a FHA incorporate the following:
--accounts for the increase to the source term owing to the use of
Global Nuclear Fuels (GNF) 10 x 10 fuel exposed over a 24-month fuel
cycle,
--reduces the Radial Peaking Factor from 2.0 to 1.95, and
--uses a calculated Control Room shine contribution that is added to
the FHA dose consequences.
The NRC computer code RADTRAD Version 3.03 is used for the
offsite and Control Room dose calculation. The RADTRAD code was
approved for use with the CNS FHA alternative source term (AST) dose
calculation in License Amendment 222.
Because the analysis affected by the changes are not considered
to be an initiator to any previously analyzed accident, these
changes cannot increase the probability of any previously evaluated
accident. Therefore, these changes do not increase the probability
of occurrence of an accident evaluated previously in the USAR.
The changes in FHA dose consequences to the Control Room
occupant resulting from the 24-month cycle/GNF 10 x 10 source term
(without crediting the offset by a reduced Radial Peaking Factor),
results in more than a minimal increase in the consequences of an
accident previously evaluated in the USAR, as stated in 10 CFR
50.59(c)(2)(iii). However, the resultant dose remains well within
the regulatory limits of 10 CFR 50.67. When the reduced Radial
Peaking Factor is applied, the dose consequences are minor.
Therefore, this change does not significantly increase the
consequences of an accident previously evaluated in the USAR.
In summary, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not involve initiators to any events in the
USAR, nor does the activity create the possibility for any new
accidents. Rather, this change is a result of the evaluation of the
most limiting FHA, which can occur at CNS. The changes to the FHA
calculation for radiological dose following a FHA incorporate the
following:
--accounts for the increase to the source term owing to the use of
GNF 10 x 10 fuel exposed over a 24-month fuel cycle,
--reduces the Radial Peaking Factor from 2.0 to 1.95, and
--uses a calculated Control Room shine contribution that is added to
the FHA dose consequences.
The RADTRAD code accommodates the use of GNF 10 x 10 fuel
exposed over a 24-month fuel cycle in calculating the FHA dose
consequences. The reduction in Radial Peaking Factor remains
bounding over the
[[Page 22571]]
CNS core design. The calculated Control Room shine contribution
replaces the previously approved qualitative assessment. The
proposed change does not introduce any new modes of plant operation
and does not involve physical modifications to the plant. 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 previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The dose consequences are calculated in accordance with the
regulatory guidance found in RG 1.183. The RADTRAD code was used, as
approved for application at CNS with License Amendment 222. With the
reduced Radial Peaking Factor applied to the GNF 10 x 10 fuel that
has been exposed over a 24-month fuel cycle, the dose consequences
are minor. The calculated shine contribution being added to the
total Control Room occupant FHA dose results are less than the
previous qualitative assessment results that are being replaced.
Accordingly, the safety margins to the regulatory dose limits are
preserved.
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. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Michael T. Markley.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: November 13, 2012.
Description of amendment request: The proposed amendment would
revise Renewed Facility Operating License (RFOL) Condition C.12 to
clarify that the programs and activities, identified in Appendix A of
NUREG-1955 and the Updated Final Safety Analysis Report (UFSAR)
supplement are to be completed before the period of extended operation.
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 amendment does not significantly increase the probability of
an accident since it does not involve a change to any plant
equipment that initiates a plant accident. The change clarifies
RFOLC [RFOL Condition] C.12. The license conditions deal with
administrative controls over information contained in the Updated
Final Safety Analysis Repo[r]t (UFSAR) supplement. The proposed
changes are administrative and the license conditions are not an
initiator or mitigator of any previously evaluated accidents.
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 amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated
since it does not involve any physical alteration of plant equipment
and does not change the method by which any safety-related system
performs its function. The license conditions deal with
administrative controls over information contained in the UFSAR
supplement. No new or different types of equipment will be installed
and the basic operation of installed equipment is unchanged.
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 proposed amendment does not affect design codes or design
margins. The change that clarifies RFOLC C.12 is administrative in
nature and does not have the ability to affect analyzed safety
margins.
Therefore, operation of DAEC in accordance with the proposed
amendment will not involve a significant reduction in the margin to
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. James Petro, P. O. Box 14000, Juno
Beach, FL 33408-0420.
NRC Branch Chief: Robert D. Carlson.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: December 21, 2012.
Description of amendment request: The proposed amendment would
modify the current DAEC Technical Specifications (TS) requirement to
operate the Standby Gas Treatment System for 10 hours on a frequency
specified in the Surveillance Frequency Control Program in accordance
with TSTF-522, Revision 0, ``Revise Ventilation System Surveillance
Requirements to Operate for 10 hours per Month.
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 replaces an existing Surveillance
Requirement to operate the SGT System equipped with electric heaters
for a continuous 10 hour period with a requirement to operate the
SGT System for 15 continuous minutes without the heaters operating.
In addition, the electrical heater output test in the VFTP
(Specification 5.5.7.e) is proposed to be deleted and a
corresponding change in the charcoal filter testing (Specification
5.5.7.c) be made to require the testing be conducted at a humidity
of at least 95% RH, which is more stringent than the current testing
requirement of 70% RH.
These systems are not accident initiators and therefore, these
changes do not involve a significant increase in the probability of
an accident. The proposed system and filter testing changes are
consistent with current regulatory guidance for these systems and
will continue to assure that these systems perform their design
function which may include mitigating accidents. Thus the change
does not involve a significant increase in the consequences of an
accident.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change replaces an existing Surveillance
Requirement to operate the SGT System equipped with electric heaters
for a continuous 10 hour period with a requirement to operate the
systems for 15 continuous minutes without the heaters operating. In
addition, the electrical heater output test in the VFTP
(Specification 5.5.7.e) is proposed to be deleted and a
corresponding change in the charcoal filter testing (Specification
5.5.7.c) be made to require the testing be conducted at a humidity
of at least 95% RH, which is more stringent than the current testing
requirement of 70% RH.
The change proposed for this ventilation system does not change
any system
[[Page 22572]]
operations or maintenance activities. Testing requirements will be
revised and will continue to demonstrate that the Limiting
Conditions for Operation are met and the system components are
capable of performing their intended safety functions. The change
does not create new failure modes or mechanisms and no new accident
precursors are generated.
Therefore, it is concluded that this 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 replaces an existing Surveillance
Requirement to operate the SGT System equipped with electric heaters
for a continuous 10 hour period with a requirement to operate the
systems for 15 continuous minutes without heaters operating. In
addition, the electrical heater output test in the VFTP is proposed
to be deleted and a corresponding change in the charcoal filter
testing be made to require the testing be conducted at a humidity of
at least 95% RH, which is more stringent than the current testing
requirement of 70% RH.
The proposed increase to 95% RH in the required testing of the
charcoal filters compensates for the function of the heaters, which
was to reduce the humidity of the incoming air to below the
currently-specified value of 70% RH for the charcoal. The proposed
change is consistent with regulatory guidance and continues to
ensure that the performance of the charcoal filters is acceptable.
Therefore, it is concluded that this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. James Petro, P.O. Box 14000, Juno Beach,
FL 33408-0420.
NRC Branch Chief: Robert D. Carlson.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: April 20, 2012.
Description of amendment request: The proposed amendment would
revise the TS 3.1.7 to approve the use of an alternative method, other
than the current method of the use of movable incore detectors system,
to monitor the position of control rod or shutdown rod, in the event of
a malfunction of the microprocessor rod position indication (MRPI)
system. The use of this alternative method would reduce the required
frequency of flux mapping using the movable incore detector system to
determine the position of the control or shutdown rod position that is
not being indicated. This will reduce the wear on the movable incore
detector system that is also used to complete other required TS
surveillances.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change provides an alternative method for verifying
rod position of one rod. The proposed change meets the intent of the
current specification in that it ensures verification of position of
the rod once every 8 hours. The proposed change provides only an
alternative method of monitoring rod position and does not change
the assumptions or results of any previously evaluated accident.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change provides only an alternative method of
determining the position of one rod. No new accident initiators are
introduced by the proposed alternative manner of performing rod
position verification. The proposed change does not affect the
reactor protection system. Hence, no new failure modes are created
that would cause a new or different kind of accidents from any
accident previously evaluated.
Therefore, operation of the facility in accordance with the
proposed amendments would not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The basis of TS 3.1.7 states that the operability of the rod
position indicators is required to determine control rod positions
and thereby ensure compliance with the control rod alignment and
insertion limits. The proposed change does not alter the requirement
to determine rod position but provides an alternative method for
determining the position of the affected rod. As a result, the
initial conditions of the accident analysis are preserved and the
consequences of previously analyzed accidents are unaffected.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.
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: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor,
Baltimore, MD 21202.
NRC Acting Branch Chief: Sean Meighan.
South Carolina Electric and Gas Docket Nos.: 52-027 and 52-028, Virgil
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: March 26, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear
Station (VCSNS) Units 2 and 3 by departing from the plant-specific
design control document Tier 2* material contained within the Updated
Safety Analysis Report (UFSAR) by revising the structural criteria code
for anchoring of reinforcement bar within the nuclear island walls.
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 functions of the nuclear island structures are to
provide support, protection, and separation for the seismic Category
I mechanical and electrical equipment located in the nuclear island.
The nuclear island structures are structurally designed to meet
seismic Category I requirements as defined in Regulatory Guide 1.29.
The change of the requirements for anchoring headed
reinforcement does not have an adverse impact on the response of the
nuclear island structures to safe shutdown earthquake ground motions
or loads due to anticipated transients or postulated accident
conditions. The change of the requirements for anchoring headed
reinforcement does not impact the support, design, or operation of
mechanical and fluid systems. 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
[[Page 22573]]
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 change is to provide the requirements for anchoring
nuclear island headed reinforcement. The proposed change does not
change the design of the nuclear island structures except to a
limited extent to redistribute the shear reinforcement in the walls
of the nuclear island. The proposed change does not impact the
support, design, or operation of mechanical or fluid systems. The
proposed change does not result in a new failure mechanism for the
nuclear island structures or new accident precursors. As a result,
the design functions of the nuclear island structures and the
seismic Category I mechanical and electrical equipment located in
the nuclear island are not adversely affected by the proposed
change.
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.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed change, thus, no margin of
safety is reduced. The limited application of alternative criteria
for headed reinforcement does not reduce the margin of safety.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart, Acting.
Southern Nuclear Operating Company Docket Nos.: 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County,
Georgia
Date of amendment request: March 20, 2013.
Description of amendment request: The proposed change would amend
Combined Licenses Nos.: NPF-91 and NPF-92 for Vogtle Electric
Generating Plant (VEGP) Units 3 and 4 by departing from the plant-
specific design control document Tier 2* material contained within the
Updated Safety Analysis Report (UFSAR) by revising the structural
criteria code for anchoring of reinforcement bar within the nuclear
island walls.
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 functions of the nuclear island structures are to
provide support, protection, and separation for the seismic Category
I mechanical and electrical equipment located in the nuclear island.
The nuclear island structures are structurally designed to meet
seismic Category I requirements as defined in Regulatory Guide 1.29.
The change of the requirements for anchoring headed
reinforcement does not have an adverse impact on the response of the
nuclear island structures to safe shutdown earthquake ground motions
or loads due to anticipated transients or postulated accident
conditions. The change of the requirements for anchoring headed
reinforcement does not impact the support, design, or operation of
mechanical and fluid systems. 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 change is to provide the requirements for anchoring
nuclear island headed reinforcement. The proposed change does not
change the design of the nuclear island structures except to a
limited extent to redistribute the shear reinforcement in the walls
of the nuclear island. The proposed change does not impact the
support, design, or operation of mechanical or fluid systems. The
proposed change does not result in a new failure mechanism for the
nuclear island structures or new accident precursors. As a result,
the design functions of the nuclear island structures and the
seismic Category I mechanical and electrical equipment located in
the nuclear island are not adversely affected by the proposed
change.
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.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed change, thus, no margin of
safety is reduced. The limited application of alternative criteria
for headed reinforcement does not reduce the margin of safety.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Blach & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Acting Branch Chief: Lawrence Burkhart.
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.
[[Page 22574]]
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the NRC's Public Document Room (PDR), located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through the Agencywide Documents Access and
Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to
[email protected].
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit 2, New London County, Connecticut
Date of amendment request: December 17, 2012, as supplemented by
January 31, 2013.
Description of amendment request: The amendment revised the
Millstone Power Station, Unit 2 (MPS2) Technical Specification (TS)
Surveillance Requirement 4.4.3.2 to remove the requirement to perform
the quarterly surveillance for a pressurizer power-operated relief
valve (PORV) block valve that is being maintained closed in accordance
with TS 3.4.3 Action a. The proposed change is consistent with the
requirements of the Standard Technical Specification--Combustion
Engineering Plants (NUREG-1432, Revision 4).
Date of issuance: March 26, 2013.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 314.
Renewed Facility Operating License No. DPR-65: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: January 22, 2012 (78 FR
4472). The supplemental letter dated January 31, 2013, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 2013.
No significant hazards consideration comments received: No.
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: December 8, 2011, as supplemented by
letters dated April 11, May 2, and September 5, 2012, and January 9 and
March 8, 2013.
Brief description of amendment: The amendment revised Surveillance
Requirement (SR) 3.3.8.1.3 (calibration of loss of power
instrumentation) to extend the frequency of the SR from 18 to 24
months, and revised certain Allowable Values in TS 3.3.8.1, ``Loss of
Power Instrumentation.''
Date of issuance: March 29, 2013.
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 179.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 17, 2012 (77 FR
22811). The supplemental letters dated April 11, May 2, and September
5, 2012, and January 9 and March 8, 2013, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 29, 2013.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of application for amendment: September 6, 2012.
Brief description of amendment: The amendment revised the technical
specifications (TS) by adding a new Limiting Condition for Operation
(LCO) 3.0.8 associated with the impact of inoperable snubbers. This LCO
establishes conditions under which TS systems would remain operable
when required snubbers are not capable of providing the related support
function. The proposed amendment is consistent with NRC's approved
Technical Specification Task Force (TSTF) Improved Standard Technical
Specifications Change Traveler, TSTF-372, Revision 4, ``Addition of LCO
3.0.8, Inoperability of Snubbers.''
Date of issuance: March 29, 2013.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 251.
Facility Operating License No. DPR-20: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 2012 (77
FR 70841).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 29, 2013.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and
2, Ogle County, Illinois
Date of application for amendment: March 22, 2012, as supplemented
by letter dated December 3, 2012.
Brief description of amendment: The proposed amendment would modify
technical specification (TS) requirements regarding steam generator
tube inspections and reporting as described in Technical Specifications
Task Force (TSTF)-510, ``Revision to Steam Generator Program Inspection
Frequencies and Tube Sample Selection,'' with proposed variations and
deviations.
Date of Issuance:. March 25, 2013.
Effective Date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 172 and 170.
Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66:
The amendments revised the TS and license.
Date of initial notice in Federal Register: (77 FR 31660; May 29,
2012). The December 3, 2012, supplement did not increase the scope of
the application and did not change the NRC staff's initial proposed
finding of no significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 25, 2013.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: April 27, 2012, as supplemented
on October 15, 2012.
Brief description of amendments: The amendments: (1) Adopted a new
methodology for preparation of the
[[Page 22575]]
reactor coolant system pressure-temperature (P-T) limits, (2) relocated
the P-T limits in the Technical Specifications (TSs) to a new licensee-
controlled document, the Pressure and Temperature Limits Report (PTLR),
and (3) modified the TSs to add references to the PTLR.
Date of issuance: April 1, 2013.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendments Nos.: 286 and 289.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the License and TSs.
Date of initial notice in Federal Register: July 3, 2012 (77 FR
39525). The letter dated October 15, 2012, provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the application beyond
the scope of the original Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 1, 2013.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: September 6, 2012, as
supplemented by letter dated January 11, 2013.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3/4.6.2.3, ``Recirculation pH Control System and
NaTB Basket Minimum Loading Requirement,'' to reduce the minimum
loading requirement of sodium tetraborate.
Date of issuance: April 2, 2013.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 257 and 253.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the TSs.
Date of initial notice in Federal Register: January 25, 2013 (78 FR
5505). The supplement dated January 11, 2013, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 2, 2013.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan
Date of application for amendments: September 12, 2012
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to adopt NRC-approved TS Task Force
(TSTF) Traveler TSTF-510, Revision 2, ``Revision to Steam Generator
Program Inspection Frequencies and Tube Sample Selection,'' using the
consolidated line item improvement process. Specifically, the
amendments revise TS 3.4.17, ``Steam Generator (SG) Tube Integrity,''
TS 5.5.7, ``Steam Generator (SG) Program,'' and TS 5.6.7, ``Steam
Generator Tube Inspection Report,'' and include TS Bases changes that
summarize and clarify the purpose of the TS.
Date of issuance: March 22, 2013.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment Nos.: 320 and 304.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revise the Operating Licenses and the Technical Specifications.
Date of initial notice in Federal Register: December 26, 2012 (77
FR 76080).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 22, 2013.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit 1, Washington County, Nebraska
Date of amendment request: March 9, 2012, as supplemented by letter
dated October 31, 2012.
Brief description of amendment: The amendment relocated the Fort
Calhoun Station (FCS) Technical Specification (TS) Limiting Condition
of Operation (LCO) 2.17, Miscellaneous Radioactive Material Sources,
and the associated Surveillance Requirement (SR) 3.13, Radioactive
Material Sources Surveillance, from the FCS TSs. NUREG-1432, Revision
3, ``Standard Technical Specifications, Combustion Engineering
Plants,'' does not contain a TS or SR for radioactive source
surveillance. The operability and surveillance requirements for leak
checking of miscellaneous radioactive material sources will be
incorporated into the FCS Updated Safety Analysis Report and associated
plant procedures.
Date of issuance: March 29, 2013.
Effective date: As of its date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment No.: 271.
Renewed Facility Operating License No. DPR-40: The amendment
revised the facility operating license and the Technical
Specifications.
Date of initial notice in Federal Register: November 13, 2012 (77
FR 67684). The supplemental letter dated October 31, 2012, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated March 29, 2013.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-272, Salem Nuclear Generating Station,
Unit 1, Salem County, New Jersey
Date of application for amendment: May 8, 2012.
Brief description of amendment: The amendment revised Salem Unit 1
Technical Specification (TS) 6.8.4.i, ``Steam Generator (SG) Program,''
to permanently exclude portions of the tube below the top of the steam
generator tubesheet from periodic steam generator tube inspections. In
addition, this amendment also revises TS 6.9.1.10, ``Steam Generator
Tube Inspection Report,'' to provide permanent reporting requirements
that have been previously established on an interim basis. The
amendment was submitted pursuant to 10 CFR 50.90, ``Application for
amendment of license, construction permit, or early site permit.''
Date of issuance: March 28, 2013.
Effective date: The license amendment is effective as of the date
of issuance and shall be implemented within 60 days.
Amendment No.: 303.
Renewed Facility Operating License No. DPR-70: The amendment
revised the facility operating license and the Technical
Specifications.
Date of initial notice in Federal Register: January 22, 2013 (78 FR
4474).
The Commission' related evaluation of the amendments is contained
in a Safety Evaluation dated March 28, 2013.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 5th day of April 2013.
[[Page 22576]]
For the Nuclear Regulatory Commission.
John D. Monninger,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2013-08756 Filed 4-15-13; 8:45 am]
BILLING CODE 7590-01-P