[Federal Register Volume 78, Number 170 (Tuesday, September 3, 2013)]
[Notices]
[Pages 54280-54294]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-21247]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0201]
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 August 9, 2013, to August 21, 2013. The last
biweekly notice was published on August 20, 2013 (78 FR 51219).
ADDRESSES: You may submit comment by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0201. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, contact
the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section
of this document.
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: 3WFN, 06-44M, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
[[Page 54281]]
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2013-0201 when contacting the NRC
about the availability of information regarding this document. You may
access publicly-available information related to this action by the
following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0201.
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-2013-0201 in the subject line of your
comment submission, in order to ensure that the NRC is able to make
your comment submission available to the public in this docket.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment submissions into ADAMS.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 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 basis 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
[[Page 54282]]
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 information (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a request or
petition for hearing (even in instances in which the participant, or
its counsel or representative, already holds an NRC-issued digital ID
certificate). Based upon this information, the Secretary will establish
an electronic docket for the hearing in this proceeding if the
Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with the NRC's 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's 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 to
[email protected], or by a toll-free call at 1-866 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: 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
[[Page 54283]]
available to the public at http://ehd1.nrc.gov/ehd/, unless excluded
pursuant to an order of the Commission, or the presiding officer.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or home phone numbers
in their filings, unless an NRC regulation or other law requires
submission of such information. However, a request to intervene will
require including information on local residence in order to
demonstrate a proximity assertion of interest in the proceeding. With
respect to copyrighted works, except for limited excerpts that serve
the purpose of the adjudicatory filings and would constitute a Fair Use
application, participants are requested not to include copyrighted
materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i)-(iii).
For further details with respect to 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].
Exelon Generation Company (EGC), LLC, Docket Nos. 50-373, and 50-374,
LaSalle County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: October 15, 2012, and August 12, 2013.
Description of amendment request: The proposed amendments would
remove License Conditions which are no longer necessary to address an
interim configuration of the LaSalle County Station (LSCS), Unit 2,
spent fuel pool prior to completing installation of NETCO-SNAP-
IN[supreg] inserts. By letter dated August 12, 2013, EGC provided
additional information and expanded the scope of the application as
originally noticed. The August 12, 2013, letter proposed to clarify
language in the LSCS, Units 1 and 2, Technical Specifications (TS)
applicable to the design features for TS 4.3, `Fuel Storage.' The
proposed amendment was initially published in the Federal Register
Biweekly notice on April 2, 2013 (78 FR 19751).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided on
August 12, 2013, its revised 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 removes License Conditions within the LSCS
Unit 2 Operating License related to interim configurations of the
SFP during the installation of the NETCO-SNAP-IN[supreg] inserts and
the required completion date for installation. The proposed change
also revises TS Section 4.3.1 to clarify that for the Unit 2 SFP,
spent fuel shall only be stored in storage rack cells containing a
neutron absorbing rack insert. All changes proposed by EGC in this
license amendment request are administrative in nature because they
remove License Conditions that have either been satisfied or that
are no longer applicable, and the revision to TS Section 4.3.1
ensures spent fuel is stored only in cells that contain inserts.
There are no physical changes to the facilities, nor any changes to
the station operating procedures, limiting conditions for operation,
or limiting safety system settings.
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 removes License Conditions within the LSCS
Unit 2 Operating License related to interim configurations of the
SFP during the installation of the NETCO-SNAP-IN[supreg] inserts and
the required completion date for installation. The proposed change
also revises TS Section 4.3.1 to clarify that for the Unit 2 SFP,
spent fuel shall only be stored in storage rack cells containing a
neutron absorbing rack insert. There are no changes to the SFP
criticality analysis associated with the proposed change. No
physical changes to the plant are proposed, and there are no changes
to the manner in which the plant is operated. Rather, the proposed
change is administrative because it involves removing License
Conditions that have either been satisfied or that are no longer
applicable, and the revision to TS Section 4.3.1 ensures spent fuel
is stored only in cells that contain inserts.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change removes License Conditions within the LSCS
Unit 2 Operating License related to interim configurations of the
SFP during the installation of the NETCO-SNAP-IN[supreg] inserts and
the required completion date for installation. The proposed change
also revises TS Section 4.3.1 to clarify that for the Unit 2 SFP,
spent fuel shall only be stored in storage rack cells containing a
neutron absorbing rack insert. Plant safety margins are established
through limiting conditions for operation, limiting safety system
settings, and safety limits specified in Technical Specifications.
The proposed change does not alter these established safety margins.
The proposed change does not alter the criticality analysis for the
SFP and does not affect the SFP criticality safety margin. The
proposed change is administrative because it involves removing
License Conditions that have either been satisfied or that are no
longer applicable, and the revision to TS Section 4.3.1 ensures
spent fuel is stored only in cells that contain inserts.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Ms. Tamra Domeyer, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
Acting NRC Branch Chief: Jeremy S. Bowen.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: June 10, 2013.
Description of amendment request: The proposed amendment revises
Technical Specification (TS) Surveillance Requirements (SR) 3.8.4.2 and
3.8.4.5. The proposed change would resolve a non-cited violation (NCV)
that was documented in an NRC's Inspection Report. Specifically, the
NRC identified an NCV for the failure to verify that safety-related
batteries would remain operable if all the inter-cell and terminal
connections were at the maximum resistance value allowed by SR 3.8.4.2
and SR 3.8.4.5 (i.e., 150 micro-ohms).
[[Page 54284]]
The proposed change maintains the existing resistance limit for inter-
cell and terminal connections, and adds new acceptance criteria for
total battery connection resistance to ensure that the safety-related
batteries can perform their specified safety function.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below
:1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The revisions of SR 3.8.4.2 and SR 3.8.4.5 to add a battery
connector resistance acceptance criterion will not challenge the
ability of the safety-related batteries to perform their safety
function. The total battery connection resistance is a parameter
that is representative of overall battery performance, and ensures
that the safety-related batteries remain capable of performing their
specified safety function. Appropriate monitoring and maintenance
will continue to be performed on the safety-related batteries. In
addition, the safety-related batteries are within the scope of 10
CFR 50.65, ``Requirements for monitoring the effectiveness of
maintenance at nuclear power plants,'' which will ensure the control
of maintenance activities associated with this equipment.
Current TS requirements will not be altered and will continue to
require that the equipment be regularly monitored and tested. Since
the proposed change does not alter the manner in which the batteries
are operated, there is no significant impact on reactor operation.
The proposed change does not involve a physical change to the
batteries, nor does it change the safety function of the batteries.
The DC power system/batteries will retain adequate independency,
redundancy, capacity, and testability to permit the functioning
required of the engineered safety features. The proposed TS revision
involves no significant changes to the operation of any systems or
components in normal or accident operating conditions and no changes
to existing structures, systems, or components.
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 changes revising SR 3.8.4.2 and SR 3.8.4.5 to add
an additional acceptance criterion for battery connector resistance
is an increase in conservatism, without a change in system testing
methods, operation, or control. Safety-related batteries installed
in the plant will be required to meet criteria more restrictive and
conservative than current acceptance criteria and standards. The
proposed change does not affect the manner in which the batteries
are tested and maintained; therefore, there are no new failure
mechanisms for the system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the
plant structures, systems, and components, the parameters within
which the plant is operated, and the setpoints for the actuation of
equipment relied upon to respond to an event. The proposed change
does not modify the safety limits or setpoints at which protective
actions are initiated. The new acceptance criterion is more
restrictive than the existing acceptance criteria for inter-cell and
terminal connection resistance, and the proposed change ensures the
availability and operability of safety-related battery operability
and availability.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
Acting NRC Branch Chief: Jeremy Bowen.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: July 5, 2013.
Description of amendment request: The proposed amendment includes
supporting changes to NMP2 Technical Specification (TS) 3.1.7,
``Standby Liquid Control (SLC) System,'' to increase the isotopic
enrichment of boron-10 in the sodium pentaborate solution utilized in
the SLC System and decrease the SLC System tank volume. The following
are the proposed changes to the NMP2 TS 3.1.7, ``Standby Liquid Control
(SLC) System'':
Revise the acceptance criterion in SR 3.1.7.10 by
increasing the sodium pentaborate boron-10 enrichment requirement from
>= 25 atom percent to >= 92 atom percent, and make a corresponding
change in TS Figure 3.1.7-1, ``Sodium Pentaborate Solution Volume/
Concentration Requirements.''
Revise TS Figure 3.1.7-1 to account for the decrease in
the minimum volume of the SLC system tank. At a sodium pentaborate
concentration of 13.6% the minimum volume changes from 4,558.6 gallons
to 1,600 gallons. At a sodium pentaborate concentration of 14.4%, the
minimum volume changes from 4,288 gallons to 1,530 gallons.
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. Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The SLC System is used to mitigate the consequences of an
Anticipated Transient Without SCRAM (ATWS) special event and is used
to limit the radiological dose during a Loss of Coolant Accident
(LOCA). The proposed changes do not affect the capability of the SLC
System to perform these two functions in accordance with the
assumptions of the associated analyses.
A SLC System failure is not a precursor of any previously
evaluated accident in the NMP2 Updated Safety Analysis Report
(USAR). Consequently there is no change in the probability of an
accident previously evaluated.
The current ATWS analysis is not adversely affected by the
proposed changes because the reactivity insertion rate would
increase by a factor greater than 3 and the amount of injected
boron-10 is not reduced. The ability of the SLC System to mitigate
radiological dose in the event of a LOCA is not affected by these
changes.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
Structures, systems and components (SSCs) previously required
for the mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes do not adversely
affect safety-related SSCs and do not challenge the performance or
integrity of any safety-related SSC. The physical changes to the SLC
System are limited to the increase in the boron-10 enrichment of the
sodium pentaborate solution in the SLC System storage tank, the
corresponding decrease in the net sodium pentaborate solution volume
requirement in the SLC System storage tank, and the associated
instrumentation changes. In addition, the effective SLC System flow
rate utilized in the boron equivalency analysis is reduced. The
proposed changes do not otherwise affect the design or operation of
the SLC System.
This change does not adversely affect any current system
interfaces or create any new interfaces that could result in an
accident or
[[Page 54285]]
malfunction of a different kind than was previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Will the change involve a significant reduction in a margin
of safety?
Response: No.
The SLC System is used to mitigate the consequences of an ATWS
event and is used to limit the radiological dose during a LOCA. The
proposed changes do not affect the capability of the SLC System to
perform these two functions in accordance with the assumptions of
the associated analyses. The current ATWS analysis is not adversely
affected by the proposed changes because the reactivity insertion
rate would increase by a factor greater than 3 and the amount of
injected boron-10 is not reduced. The ability of the SLC System to
mitigate radiological dose in the event of a LOCA by maintaining
suppression pool pH >= 7.0 is not affected by these changes.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Gautam Sen, Senior Counsel, Constellation
Energy Nuclear Group, LLC, 100 Constellation Way, Suite 200C,
Baltimore, MD 21202.
Acting NRC Branch Chief: Robert Beall.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello
Nuclear Generating Plant (MNGP), Wright County, Minnesota
Date of amendment request: April 19, 2013.
Description of amendment request: The licensee proposed to revise
MNGP Technical Specification (TS) 1.1, ``Definitions,'' to modify the
definition of ``Shutdown Margin (SDM)'' to require calculation of the
SDM at a reactor moderator temperature of 68 degrees Fahrenheit
([deg]F), or at a higher temperature that represents the most reactive
state throughout the operating cycle. This change is needed for newer
boiling water reactor fuel designs which may be more reactive at
shutdown temperatures above 68[emsp14][deg]F. The proposed change is
consistent with Technical Specifications Task Force (TSTF) Traveler
TSTF-535, Revision 0, ``Revise Shutdown Margin Definition to Address
Advanced Fuel Designs.'' Notice of availability of TSTF-535 was
published in the Federal Register on February 26, 2013 (78 FR 13100).
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 (NSHC), which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the definition of SDM. SDM is not an
initiator to any accident previously evaluated. Accordingly, the
proposed change to the definition of ADM has no effect on the
probability of any accident previously evaluated. ADM is an
assumption in the analysis of some previously evaluated accidents
and inadequate SDM could lead to an increase in consequences for
those accidents. However, the proposed change revised the SDM
definition to ensure that the correct SDM is determined for all fuel
types at all times during the fuel cycle. As a result, the proposed
change does not adversely affect the consequences of any accident
previously evaluated.
Therefore, it is concluded that these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the definition of SDM. The change
does not involve a physical alteration of the plant (i.e., no new of
different type of equipment will be installed) or a change in
methods governing normal plant operations. The change does not alter
assumptions made in the safety analysis regarding SDM.
Therefore, it is concluded that these changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revised the definition of SDM. The proposed
change does not alter the manner in which safety limits, limiting
safety system settings or limiting conditions for operation are
determined. The proposed change ensures that the SDM assumed in
determining safety limits, limiting safety system settings or
limiting conditions for operation is correct for all BWR fuel types
at all times during the fuel cycle.
Therefore, it is concluded that these changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for the licensee: Peter M. Glass, Assistant General
Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN
55401.
NRC Branch Chief: Robert D. Carlson.
Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power
Plant (HBPP), Unit 3 Humboldt County, California
Date of amendment request: May 3, 2013.
Description of amendment request: The proposed amendment would add
License Condition 2.C.5 that approves the License Termination Plan
(LTP) and adds a license condition that establishes the criteria for
determining when changes to the LTP require prior NRC approval.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The change allows for the approval of the LTP and provides the
criteria for when changes to the LTP require prior NRC approval.
This change does not affect possible initiating events for the
decommissioning accidents previously evaluated in the Humboldt Bay
Power Plant (HBPP) defueled safety analysis report (DSAR), as
updated, appendix A, ``Implications of Decommissioning Accidents
with Potential for Radiological Impacts to the Environment,' or
alter the configuration or operation of the facility. Safety limits,
limiting safety system settings, and limiting control systems are no
longer applicable to HBPP in the permanently defueled mode, and are
therefore not relevant.
The proposed change does not affect the boundaries used to
evaluate compliance with liquid or gaseous effluent limits, and has
no impact on plant operations.
Therefore, the proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The safety analysis for the facility remains accurate as
described in the HBPP DSAR, as updated, appendix A. There are
sections of the LTP that refer to the decommissioning activities
still remaining (e.g. removal of large components, decontamination,
etc.). However, these activities are performed in accordance with
approved HBPP work packages/steps and undergo 10 CFR 50.59 screening
prior to initiation. The proposed amendment merely makes mention of
these processes and does not bring about physical changes to the
facility. Therefore, the facility
[[Page 54286]]
conditions for which the postulated accidents have been evaluated
are still valid and no new accident scenarios, failure mechanisms,
or single failures are introduced by this amendment. The system
operating procedures are not affected.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
There are no changes to the design or operation of the facility
resulting from this amendment. The proposed change does not affect
the boundaries used to evaluate compliance with liquid or gaseous
effluent limits, and has no impact on plant shutdown operations.
Accordingly, neither the postulated accident assumptions in the
DSAR, as updated, appendix A, nor the Technical Specifications are
affected.
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: Ms. Jennifer K. Post, Pacific Gas and
Electric Company, 77 Beale Street, B30A, San Francisco, CA.
NRC Branch Chief: Bruce Watson.
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: July 17, 2013.
Description of amendment request: The proposed amendment would
depart from VCSNS Units 2 and 3 plant-specific Design Control Document
(DCD) Tier 2 and Tier 2* material contained within the Updated Final
Safety Analysis Report (UFSAR) to acknowledge various obstructions and
interferences (other than wall openings and penetrations) that may
cause a change to the design spacing of shear studs and the design and
spacing of wall module trusses in a local area, and to acknowledge
appropriate weld types.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design function of the containment structural modules is to
support the reactor coolant system components and related piping
systems and equipment. The design functions of the affected
structural modules in the auxiliary building are to provide support
and protection for new and spent fuel and the equipment needed to
support fuel handling, cooling, and storage in the spent fuel racks,
and to provide support, protection, and separation for the seismic
Category I mechanical and electrical equipment located outside the
containment building. The design function of the shear studs is to
enable the concrete and steel faceplates to act in a composite
manner and transfer loads into the concrete of the structural
modules. The structural modules are seismic Category I structures
and are designed for dead, live, thermal, pressure, safe shutdown
earthquake loads, and loads due to postulated pipe breaks. The loads
and load combinations applicable to the structural modules in the
auxiliary building are the same as for the containment internal
structures except that there are no design basis accident loadings
due to the automatic depressurization system or pressure loads due
to pipe breaks. The proposed changes to the UFSAR are to include
types of interferences other than wall openings and penetrations
that may cause a change in the design spacing of shear studs and the
design and spacing of wall module trusses in a local area. The
proposed changes clarify that the stud spacing is specified as a
design value and add the tolerance for stud spacing. The revised
spacing including the tolerance continues to be in conformance with
the design and analysis requirements identified in the UFSAR. The
proposed changes also include clarification of a requirement for a
complete joint penetration weld. The thickness, geometry, and
strength of the structures are not adversely altered. The material
of the steel plates is not altered. The properties of the concrete
included in the structural modules are not altered. As a result, the
design function of the containment structural modules is not
adversely affected by the proposed change. There is no change to
plant systems or the response of systems to postulated accident
conditions. There is no change to the predicted radioactive releases
due to postulated accident conditions. The plant response to
previously evaluated accidents or external events is not adversely
affected, nor does the change described create any new accident
precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to the UFSAR acknowledge types of
interferences (other than wall openings and penetrations) that may
cause a change in the typical design spacing of shear studs and the
design and spacing of wall module trusses in a local area. The
proposed changes clarify that the stud spacing is specified as a
design value and provide the tolerance for stud spacing. The revised
spacing, including the tolerance, continues to be in conformance
with the design and analysis requirements identified in the UFSAR.
Stud spacing and sizing are evaluated to demonstrate that stud
loadings and shear transfer capability are within acceptable limits
and that the structural module acts in a composite manner. An
additional proposed change is to clarify a requirement for a
complete joint penetration weld. The thickness, geometry, and
strength of the structures are not adversely altered. The materials
of the steel plates are not altered. The properties of the concrete
included in the structural modules are not altered. The changes to
the internal design of the structural modules do not create any new
accident precursors. As a result, the design function of the modules
is not adversely affected by the proposed changes.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The criteria and requirements of American Concrete Institute
(ACI) 349 and American Institute of Steel Construction (AISC) N690
provide a margin of safety to structural failure. The design of the
shear studs and wall trusses for the structural wall modules
conforms to applicable criteria and requirements in ACI 349 and AISC
N690 and, therefore, maintain the margin of safety. The proposed
changes to the UFSAR acknowledge types of interferences (other than
wall openings and penetrations) that may cause a change in the
typical design spacing of shear studs and the design and spacing of
wall module trusses in a local area. The proposed changes clarify
that the stud spacing is specified as a design value and add the
tolerance for stud spacing. The revised spacing including the
tolerance continues to be in conformance with the design and
analysis requirements identified in the UFSAR. An additional
proposed change is to clarify a requirement for a complete joint
penetration weld. There is no change to the capacity of the weld or
to the design requirements of the modules. There is no change to the
method of evaluation from that used in the design basis
calculations.
Therefore, the proposed amendment does not result in a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
[[Page 54287]]
NRC Branch Chief: Lawrence Burkhart.
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 15, 2013, and revised on July 10,
2013, and supplemented on August 16, 2013.
Description of amendment request: The proposed change would amend
Combined License 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 (DCD) Tier 1 (and corresponding Combined License
Appendix C information) and Tier 2 material by making changes to the
Non-Class 1E dc and Uninterruptible Power Supply System (EDS) and
Uninterruptible Power Supply System (IDS) and making changes to the
corresponding Tier 1 information in Appendix C to the Combined License.
The proposed changes would:
(1) Increase EDS total equipment capacity, component ratings,
and protective device sizing to support increased load demand,
(2) Relocate equipment and moving Turbine Building (TB) first
bay EDS Battery Room and Charger Room. The floor elevation increases
from elevation 148'-0'' to elevation 148'-10'' to accommodate
associated equipment cabling with this activity, and
(3) Remove the Class 1E IDS Battery Back-up tie to the Non-Class
1E EDS Battery.
Because this proposed change requires a departure from Tier 1
information in the Westinghouse Advanced Passive 1000 design control
document (DCD), the licensee also requested an exemption from the
requirements of the Generic DCD Tier 1 in accordance with 52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design function of the Turbine Building (TB) is to provide
weather protection for the laydown and maintenance of major turbine/
generator components. The TB first bay is a seismic Category II
structure designed to prevent the collapse under a safe shutdown
earthquake (SSE) to protect the adjacent auxiliary building. The
electrical system and air-handling units are designed to provide
electrical power to plant loads and maintain acceptable temperatures
for electrical equipment rooms and work areas. The electrical
equipment continues to be in accordance with the same codes and
standards stated in the Updated Final Safety Analysis Report
(UFSAR). The proposed relocation of equipment, including the
increase in floor elevation by 10 inches to accommodate overhead
equipment cabling, does not impact the TB design function. The TB
first bay continues to meet seismic Category II requirements. Based
on this, the proposed changes would not increase the probability of
an accident previously evaluated.
The proposed changes do not involve any accident initiating
event, thus the probabilities of the accidents previously evaluated
are not affected. The relocation of equipment does not involve any
safety-related structures, systems, or components; the affected
rooms do not represent a radioactive material barrier; and this
activity does not affect the containment of radioactive material.
The radioactive material source terms and release paths used in the
safety analyses are unchanged, thus the radiological releases in the
accident analyses are not affected. Therefore, the consequences of
an accident previously evaluated are not affected.
Therefore the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes would use the same type of electrical
equipment with higher ratings and capacity, change the source of a
battery back-up, and relocate equipment. The electrical equipment
will continue to perform its design functions because the same
electrical codes and standards as stated in the UFSAR continue to be
met. Therefore the proposed changes do not affect equipment failure
probabilities or alter any accident initiator or initiating sequence
of events. The proposed changes in location of equipment and
elevation of the TB first bay floor do not affect the design
function of the TB first bay to protect the adjacent auxiliary
building by meeting seismic Category II structure requirements, or
affect the operation of the relocated equipment, or the ability of
the relocated equipment to meet its design functions. Because the
SSCs and equipment affected by the proposed changes continue to meet
their design functions, the structural codes and standards as stated
in the UFSAR, the proposed changes do not introduce a different type
of accident than those previously considered.
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 current seismic requirements applicable to the seismic
Category II TB first bay structure, including the seismic modeling
and analysis methods, will continue to apply to the TB first bay
floor elevation increase. The proposed changes to relocate equipment
and the increase in the floor elevation will continue to meet the
fire rating requirements and will be in accordance with the same
codes and standards currently identified in the UFSAR. The proposed
changes to the electrical equipment will continue to meet existing
electrical equipment industry standard recommendations identified in
the UFSAR. Because no safety analysis or design basis acceptance
limit/criterion is challenged or exceeded by these proposed changes,
no margin of safety is reduced.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence Burkhart.
Southern Nuclear Operating Company, Docket Nos. 52-025, and 52-026,
Vogtle Electric Generating Station (VEGP) Units 3 and 4, Burke County,
Georgia
Date of amendment request: July 2, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-91, and NPF-92 for VEGP Units 3 and 4,
respectively, by revising Tier 2* and associated Tier 2 information
related to the design details of connections in several locations
between the steel plate composite construction (SC) used for the shield
building and the standard reinforced concrete (RC) walls, floors, and
roofs of the auxiliary building and lower walls of the shield building.
These connections are also referred to as ``RC to SC connections.''
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
[[Page 54288]]
seismic Category I requirements as defined in Regulatory Guide 1.29.
The change to the detail design of connections between the RC and SC
structures do 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 changes to the detail design do 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 do the changes describe create any
new accident precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes are to the detail design of connections
between the RC and SC structures. The changes to the detail design
of connections do not change the criteria and requirements for the
design and analysis of the nuclear island structures. The changes to
the detail design of connections do not change the design function,
support, design, or operation of mechanical and fluid systems. The
changes to the detail design of connections do not change the
methods used to connect the RC to the SC. The changes to the detail
design of the connections do 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 are
not adversely affected by the proposed changes.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
No safety analysis or design basis acceptance limit/criterion is
involved by the requested changes, thus, no margin of safety is
reduced.
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. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmington, AL 35203-2015.
NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Docket Nos. 52-025, and 52-026,
Vogtle Electric Generating Station (VEGP) Units 3 and 4, Burke County,
Georgia
Date of amendment request: July 15, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-91, and NPF-92 for VEGP Units 3 and 4,
respectively, by revising Tier 2* information related to the
construction of Module CA03. Some of these changes include the removal
of specifically mentioned materials, increasing anchoring supports and
allowing the use of anchor bars with hooks.
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 to the design details for the in-containment refueling
water storage tank (IRWST) west wall 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, nor does it change the seismic
Category I classification. The change to the design details for the
IRWST west wall 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 revise design details for the IRWST
west wall. The change of the design details for the IRWST west wall
does not change the design requirements of the nuclear island
structures, nor the seismic Category I classification. The change of
the design details for the IRWST west wall does not change the
design function, support, design, or operation of mechanical and
fluid systems. The change of the design details for the IRWST west
wall does not result in a new failure mechanism for the nuclear
island structures or introduce any new accident precursors. As a
result, the design function of the nuclear island structures is 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
involved by the requested changes, thus, no margin of safety is
reduced.
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. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence J. Burkhart.
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: August 6, 2013.
Description of amendment request: The proposed change would amend
Combined License 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 (DCD) Tier 1(and corresponding Combined License
Appendix C information) and Tier 2 material by revising the safety
function and classification of Liquid Radwaste System (WLS) drain hubs
in the Chemical and Volume Control System and Passive Core Cooling
System (PXS) compartments. In addition, the proposed changes would
modify the PXS compartment drain piping connection; WLS valve types,
and depiction of components in the WLS figures.
Because, this proposed change requires a departure from Tier 1
information in the Westinghouse Advanced Passive 1000 DCD, the licensee
also requested an exemption from the requirements of the Generic
[[Page 54289]]
DCD Tier 1 in accordance with 52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design function of the WLS is containment isolation and the
prevention of backflow in the drain lines from the CVS compartment
and the PXS compartment to the containment sump which prevents cross
flooding of these compartments. The proposed changes to the WLS
drainage function; the CVS and PXS compartment drain hubs; and the
WLS valve types do not affect these design functions or any other
system design function. Revising the drain hub safety
classification, the PXS drains connection type, and the WLS valve
types do not involve any accident initiating event or component
failure. The changes to how components (valves, filters) are
depicted in the figure provide consistency with the figure legend
and do not alter any system functions. The system will utilize the
same codes and standards previously used for the system. Since there
are no impacts on accident initiating events or component failures,
the probability of an accident previously evaluated is not affected.
The radioactive material source terms and release paths used in the
safety analyses are unchanged, thus the radiological releases in the
Updated Final Safety Analysis Report (UFSAR) accident analyses are
not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to the WLS system do not adversely affect
the design or quality of any structure, system or component.
Revising the WLS safety functions and re-classifying the drain hubs
as nonsafety-related does not create a new fault or sequence of
events that could result in a radioactive material release nor do
the changes to the WLS piping connections, valve types and the
depiction of components on the figure have any impact on any
accident previously evaluated.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to the WLS system drain hubs, piping
connection, valve type, and Tier 1 figure depiction would not affect
any radioactive material barrier. No safety analysis or design basis
acceptance limit/criterion is challenged or exceeded by the proposed
change, thus no margin of safety is reduced.
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, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348, and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: December 21, 2012, as supplemented on
May 21, 2013.
Description of amendment request: The proposed amendments would
revise the Joseph M. Farley Nuclear Plant (FNP) Facility Operating
Licenses (FOL), Appendix C, to require Southern Nuclear Operating
Company (SNC) to fully implement and maintain in effect the Degraded
Voltage Protection modification schedule.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the FNP FOL that incorporates the
Degraded Voltage Protection modification implementation schedule is
administrative in nature. This proposed change does not alter
accident analysis assumptions, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested or inspected.
Therefore, this proposed change does not involve a significant
increase in the Probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change to the FNP FOL that incorporates the
Degraded Voltage Protection modification implementation schedule is
administrative in nature. This proposed change does not alter
accident analysis assumptions, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested or inspected.
Therefore, this proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Plant safety margins are established through limiting conditions
for operation, limiting safety system settings, and safety limits
specified in the technical specifications. The proposed change to
the FNP FOL is administrative in nature. Because there is no change
to these established safety margins as a result of this change, the
proposed change does not involve a significant reduction in a margin
of safety.
Therefore, the proposed change does not involve a significant
reduction in margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Leigh D. Perry, SVP & General Counsel,
Southern Nuclear Operating Company, 40 Inverness Center Parkway,
Birmingham, AL 35242.
NRC Branch Chief: Robert J. Pascarelli.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321, and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: December 21, 2012, as supplemented June
21, 2013.
Description of amendment request: The proposed License Amendment
Request (LAR) would revise the Edwin I. Hatch Nuclear Plant (HNP)
Facility Operating Licenses to require Southern Nuclear Operating
Company (SNC) to implement modifications that will eliminate the need
for administrative controls with regard to protection of the plant from
degraded grid voltage conditions for HNP.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 54290]]
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the HNP FOL that incorporates the
Degraded Voltage Protection modification implementation schedule is
administrative in nature. This proposed change does not alter
accident analysis assumptions, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested or inspected.
Therefore, this proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change to the HNP FOL that incorporates the
Degraded Voltage Protection modification implementation schedule is
administrative in nature. This proposed change does not alter
accident analysis assumptions, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested or inspected.
Therefore, this proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Plant safety margins are established through limiting conditions
for operation, limiting safety system settings, and safety limits
specified in the technical specifications. The proposed change to
the HNP FOL is administrative in nature. Because there is no change
to these established safety margins as a result of this change, the
proposed change does not involve a significant reduction in a margin
of safety.
Therefore, the proposed change does not involve a significant
reduction in margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Leigh D. Perry, SVP & General Counsel,
Southern Nuclear Operating Company, 40 Inverness Center Parkway,
Birmingham, AL 35242.
NRC Branch Chief: Robert J. Pascarelli.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321, and 50-366, Edwin I. Hatch
Nuclear Plant (HNP), Units 1 and 2, Appling County, Georgia
Date of amendment request: July 23, 2013.
Description of amendment request: The proposed amendments would
modify Technical Specification (TS) requirements related to control
room envelope (CRE) habitability in accordance with the Nuclear
Regulatory Commission (NRC)-approved Revision 3 of Technical
Specification Task Force (TSTF) Standard Technical Specifications (STS)
Change Traveler TSTF-448, ``Control Room Habitability.''
The NRC staff published a notice of opportunity for comment in the
Federal Register on October 17, 2006 (71 FR 61075), on possible license
amendments adopting TSTF-448 using the NRC's consolidated line-item
improvement process (CLIIP) for amending licensees' TSs, which included
a model safety evaluation (SE) and model no significant hazards
consideration (NSHC) determination. The NRC staff subsequently issued a
notice of availability of the models for referencing in license
amendment applications in the Federal Register on January 17, 2007 (72
FR 2022), which included the resolution of public comments on the model
SE and model NSHC determination. The licensee affirmed the
applicability of the following NSHC determination in its application
dated July 23, 2013.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC is presented below:
Criterion 1
The Proposed Change Does Not Involve a Significant Increase in
the Probability or Consequences of an Accident Previously Evaluated.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2
The Proposed Change Does Not Create the Possibility of a New or
Different Kind of Accident from any Accident Previously Evaluated.
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Criterion 3
The Proposed Change Does Not Involve a Significant Reduction in
the Margin of Safety.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation as determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
[[Page 54291]]
Attorney for licensee: Leigh D. Perry, SVP & General Counsel,
Southern Nuclear Operating Company, 40 Inverness Center Parkway,
Birmingham, AL 35242.
NRC Branch Chief: Robert Pascarelli.
Tennessee Valley Authority, Docket Nos. 50-327, and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: July 3, 2013 (SQN-TS-12-04).
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TSs) 3/4.6.5, ``Ice Condenser.''
The proposed changes would revise TS Limiting Condition for Operation
3.6.5.1.d and TS Surveillance Requirement 4.6.5.1.d.2 to raise the
overall ice condenser ice weight from 2,225,880 pounds (lbs) to
2,540,808 lbs and to raise the minimum TS ice basket weight from 1145
lbs to 1307 lbs, respectively. These changes are necessary to address
the issues raised in Nuclear Safety Advisory Letter (NSAL) 11-5,
``Westinghouse LOCA [Loss-of-Coolant Accident] Mass and Energy Release
Calculation Issues.'' The issues identified in NSAL-11-5 affected
plant-specific LOCA mass and energy release calculation results that
are used as input to the containment integrity response analyses. The
basis for the proposed changes is provided in WCAP-12455, Revision 1,
Supplement 2R, ``Tennessee Valley Authority [TVA] Sequoyah Nuclear
Plant [SQN] Units 1 and 2 Containment Integrity Reanalyses Engineering
Report.''
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 consequence of an accident previously evaluated?
Response: No.
The analyzed accidents of consideration in regards to changes
affecting the ice condenser are a loss of coolant accident (LOCA)
and a main steam line break (MSLB) inside containment. The ice
condenser is a passive system and is not postulated as being the
initiator of any LOCA or MSLB and is designed to remain functional
following a design basis earthquake. In addition, the ice condenser
does not interconnect or interact with any systems that have an
interface with the reactor coolant or main steam systems.
For SQN, the LOCA is the more severe accident in terms of
containment pressure and ice bed melt out, and is therefore the more
limiting accident. The revised SQN LOCA containment integrity
analysis determined that the post-LOCA peak containment pressure is
below the containment design pressure and that the margin to ice
meltout is maintained. The analysis assumes an ice weight that
ensures sufficient heat removal capability is available from the ice
condenser to limit the accident peak pressure inside containment.
TVA has evaluated the effects of the increased ice condenser ice
weight and determined that the increase in ice weight does not
invalidate the ice condenser seismic qualification, does not
adversely affect the capacity of the ice bed to absorb iodine during
a LOCA, and does not diminish the boron concentration of the
recirculated primary coolant during a LOCA.
TVA has also evaluated differences between the as-built plant
and the assumptions of the revised analysis and determined that the
results of the revised analysis remain valid for Model 57AG steam
generators and for AREVA Advanced W17 High Thermal Performance (HTP)
fuel.
The proposed changes reflect the ice weight assumed in the
containment integrity analysis including conservative allowances for
sublimation and weighing instrument systematic error. Accordingly,
the proposed changes ensure that ice weight values maintain margin
between the calculated peak containment accident pressure and the
containment design pressure. The results of the analysis and the
margins are maintained; therefore, the consequences of a previously
evaluated accident are not adversely affected by the proposed
changes.
Because (1) the ice condenser is not an accident initiator, (2)
the results of the revised analysis remain valid for Model 57AG
steam generators and for AREVA Advanced W17 High Thermal Performance
(HTP) fuel, and (3) the proposed changes to the TSs are limited to
revision of the ice weight values to reflect the revised containment
integrity analysis, there is no change in the probability of an
accident previously evaluated in the SQN Updated Final Safety
Analysis Report (UFSAR).
Based on the above discussions, the proposed changes do not
involve an 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 ice condenser serves to limit the peak pressure inside
containment following a LOCA or MSLB. The proposed changes are
limited to the revision of the minimum ice weights specified in the
TSs. The revised containment pressure analysis determined that
sufficient ice would be present to maintain the peak containment
pressure below the containment design pressure. No new modes of
operation, accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of this proposed change.
TVA has evaluated the effects of the increased ice condenser ice
weight and determined that the increase in ice weight does not
invalidate the ice condenser seismic qualification, does not
adversely affect the capacity of the ice bed to absorb iodine during
a LOCA, and does not diminish the boron concentration of the
recirculated primary coolant during a LOCA. TVA has also evaluated
differences between the as-built plant and the assumptions of the
revised analysis and determined that the results of the revised
analysis remain valid for Model 57AG steam generators and for AREVA
Advanced W17 High Thermal Performance (HTP) fuel. Because sufficient
ice weight is available to maintain the peak containment pressure
below the containment design pressure, the results of the revised
analysis remain valid for Model 57AG steam generators and for AREVA
Advanced W17 High Thermal Performance (HTP) fuel, and the increase
in ice weight does not invalidate the ice condenser seismic
qualification, the increased ice weight does not create the
possibility of an accident that is different than any already
evaluated in the SQN UFSAR.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The operability of the ice bed ensures that the required ice
inventory will (1) be distributed evenly through the containment
bays, (2) contain sufficient boron to preclude dilution of the
containment sump following the LOCA and (3) contain sufficient heat
removal capability to condense the reactor system volume released
during a LOCA. These conditions are consistent with the assumptions
used in the accident analyses.
The revised analysis demonstrates that the ice condensers will
continue to preclude over-pressurizing the lower containment and
continue to absorb sufficient heat energy to assist in precluding
containment vessel failure. TVA has evaluated the effects of the
increased ice condenser ice weight and determined that the increase
in ice weight does not invalidate the ice condenser seismic
qualification, does not adversely affect the capacity of the ice bed
to absorb iodine during a LOCA, and does not diminish the boron
concentration of the recirculated primary coolant during a LOCA.
The proposed changes are required to resolve non-conservative
TSs currently addressed by administrative controls established in
accordance with Nuclear Regulatory Commission (NRC) Administrative
Letter 98-10. The revised containment integrity response analysis
requires an increase in the required ice weight to ensure that the
post-LOCA peak containment pressure remains within the design
limits. As a result, the proposed changes restore margin between the
accident peak pressure and the containment design pressure and
resolve non-conservative TSs ice weight values currently under
administrative controls. Accordingly, the proposed changes do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review; it appears that the three
[[Page 54292]]
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: General Counsel, Tennessee Valley
Authority, 400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee
37902. Acting NRC Branch Chief: Douglas A. Broaddus.
Virginia Electric and Power Company, Docket Nos. 50-338, and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Virginia Electric and Power Company, Docket No. 50-280, and 50-281,
Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of amendment request: June 26, 2013.
Description of amendment request: The proposed license amendment
(Agencywide Documents Access and Management System (ADAMS) Accession
No. ML13179A014) requests the approval of (1) generic application of
Appendix D, ``Qualification of the ABB-NV and WLOP Critical Heat Flux
(CHF) Correlations in the Dominion VIPRE-D Computer Code,'' to Fleet
Report DOM-NAF-2-A, ``Reactor Core Thermal-Hydraulics Using the VIPRE-D
Computer Code,'' (2) the plant-specific application of Appendix D to
DOM-NAF-2-A to North Anna and Surry Power Stations (in accordance with
Section 2.1 of DOM-NAF-2-A), and (3) an increase in the Surry Power
Station Technical Specification Minimum Temperature for Criticality.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The first and second proposed changes would allow Dominion to
use the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs to
perform licensing calculations for North Anna and Surry, using the
DDLs documented in Appendix D of Fleet Report DOM-NAF-2. Neither
code/correlation pair methodology makes any contribution to the
potential accident initiators and thus cannot increase the
probability of any accident. Further, since the DDLs for ABB-NV and
WLOP meet the required design basis of avoiding departure from
nucleate boiling (DNB) with 95% probability at a 95% confidence
level, the use of the new code/correlations does not increase the
potential consequences of any accident. The pertinent evaluations
that need to be performed as part of the cycle specific reload
safety analysis to confirm that the existing safety analyses remain
applicable have been performed and determined to be acceptable. The
use of a different code/correlation pair will not increase the
probability of an accident because plant systems will not be
operated in a different manner, and system interfaces will not
change. The use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/
correlation pairs to perform licensing calculations for North Anna
and Surry will not result in a measurable impact on normal operating
plant releases and will not increase the predicted radiological
consequences of accidents postulated in the Updated Final Safety
Analysis Report (UFSAR). Therefore, neither the probability of
occurrence nor the consequences of any accident previously evaluated
is significantly increased.
The third proposed change, an increase of the Surry Minimum
Temperature for Criticality limit from 522 [deg]F to 538 [deg]F,
would provide Dominion with increased flexibility during loading
pattern development as well as improved design margins when coupled
with the second proposed change. The Minimum Temperature for
Criticality is used within the reload verification process to ensure
the assumptions made in the safety analysis remain bounding for the
given cycle design. With implementation of the proposed change, the
reload design and licensing requirements will remain in place and
continue to be met at the increased Minimum Temperature for
Criticality limit.
The increase in the Surry Minimum Temperature for Criticality
limit will not increase the probability of an accident because plant
systems will not be operated in a different manner, and system
interfaces will not change. Should the reactor coolant system (RCS)
temperature fall below the proposed limit, the unit would be in an
abnormal condition requiring operator action. The operator actions
are not changing as a result of the increased Minimum Temperature
for Criticality limit. The increase in the Surry Minimum Temperature
for Criticality will not result in a measurable impact on normal
operating plant releases and will not increase the predicted
radiological consequences of accidents postulated in the UFSAR.
Therefore, neither the probability of occurrence nor the
consequences of any accident previously evaluated is significantly
increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed).
The use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation
pairs and the applicable fuel design limits for DNB ratio (DNBR)
does not impact any of the applicable design criteria and the
pertinent licensing basis criteria will continue to be met.
Demonstrated adherence to these standards and criteria precludes new
challenges to components and systems that could introduce a new type
of accident. Setpoint safety analysis evaluations have demonstrated
that the use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation
pairs is acceptable. Design and performance criteria will continue
to be met, and no new single failure mechanisms will be created. The
use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs
does not involve any alteration to plant equipment or procedures
that would introduce any new or unique operational modes or accident
precursors.
The increase in the Surry Minimum Temperature for Criticality
does not result in any plant design changes. In addition, the
minimum temperature at which the reactor is taken critical is not an
accident initiator. The nominal average reactor coolant system
temperature during an approach to criticality is several degrees
higher than the limit proposed for the Minimum Temperature for
Criticality.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously analyzed.
3. Does this change involve a significant reduction in a margin
of safety?
The first two proposed changes would allow Dominion to use the
VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs to perform
licensing calculations for North Anna and Surry using the DDLs
documented in Appendix D of Fleet Report DOM-NAF-2. North Anna TS
2.1, ``Safety Limits,'' states that, ``the departure from nucleate
boiling ratio (DNBR) shall be maintained greater than or equal to
the 95/95 DNBR criterion for the DNB correlations and methodologies
specified in Section 5.6.5 [COLR].'' The DNBR limits meet the design
basis of avoiding DNB with 95% probability at a 95% confidence
level. Surry TS 2.1, ``Safety Limits, Reactor Core,'' specifies that
``for transients analyzed using the deterministic methodology, the
DNBR shall be maintained greater than or equal to the applicable DNB
correlation limit.'' The required DNBR margin of safety for North
Anna and Surry, which in this case is the margin between the 95/95
DNBR limit and clad failure, is therefore not reduced. Therefore,
the proposed TS changes do not involve a significant reduction in a
margin of safety.
The increased Minimum Temperature for Criticality in conjunction
with the appropriate core designs will ensure the current TS limits
for the most positive moderator temperature coefficient will
continue to be satisfied. The current analyses are bounding and
remain applicable with the increased Minimum Temperature for
Criticality. 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
[[Page 54293]]
NRC Branch Chief: Robert Pascarelli.
Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the 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].
Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power
Plant, Unit 1, (Fermi 1) Monroe County, Michigan.
Date of amendment request: December 21, 2012 (ML13002A037).
Brief description of amendment: This amendment revised the Fermi 1
license to change the licensee's name on the license to ``DTE Electric
Company.'' This name change is purely administrative in nature. Detroit
Edison is a wholly owned subsidiary of DTE Energy Company, and this
name change is part of a set of name changes of DTE Energy subsidiaries
to conform their names to the ``DTE'' brand name. No other changes are
contained within this amendment. This change does not involve a
transfer of control over or of an interest in the license for Fermi 1.
Date of issuance: August 8, 2013.
Effective date: On the date of issuance of this amendment and must
be fully implemented no later than 60-calendar days from the date of
issuance.
Amendment No.: 21.
Facility Operating License No. DPR-9: Amendment revised the License
by replacing ``the Detroit Edison'' with ``DTE Electric'' on pages 1,
2, 4, and 5.
Date of initial notice in Federal Register: March 19, 2013 (78 FR
16876).
The NRC's related evaluation of the amendment is contained in a
Safety Evaluation dated August 8, 2013.
No significant hazards consideration comments: None received.
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413, and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: November 22, 2011, as
supplemented by letters dated July 9, 2012, November 12, 2012, January
28, 2013, and May 15, 2013.
Brief description of amendments: The amendments revised the
Technical Specifications to allow single discharge header operation of
the nuclear service water system for a time period of 14 days.
Date of issuance: August 9, 2013.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 271 and 267.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the licenses and the Technical Specifications.
Date of initial notice in Federal Register: May 15, 2012 (77 FR
28630). The supplements dated July 9, 2012, November 12, 2012, January
28, 2013, and May 15, 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 amendments is contained
in a Safety Evaluation dated August 9, 2013.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station (VYNPS),
Vernon, Vermont
Date of amendment request: December 21, 2012, as supplemented on
March 19, April 29, May 7, May 14, and June 26, 2013.
Brief description of amendment: The amendment revised the VYNPS
licensing basis relative to how the station satisfies the requirements
of 10 CFR 50.63, ``Loss of all alternating current power,'' by
replacing the Vernon Hydroelectric Station with an onsite diesel
generator as the alternate alternating current power source that would
provide acceptable capability to withstand a station blackout under 10
CFR 50.63(c)(2). The change involves revisions to the VYNPS facility
and procedures described in the Updated Final Safety Analysis Report.
Date of Issuance: August 15, 2013.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 258.
Facility Operating License No. DPR-28: The amendment revised the
License.
Date of initial notice in Federal Register: March 19, 2013 (78 FR
16881).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated August 15, 2013.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota (NSPM), Docket No. 50-263,
Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota
Date of application for amendment: December 6, 2012.
Brief description of amendment: The amendment revises the MNGP
Technical Specifications (TS) Section 3.10.1. Specifically, the
amendment revises Limiting Condition for Operation 3.10.1 and the
associated TS Bases to expand its scope to include provisions for
temperature excursions
[[Page 54294]]
greater than 212[emsp14][deg]F as a consequence of inservice leak and
hydrostatic testing, and as a consequence of scram time testing
initiated in conjunction with an inservice or hydrostatic test, while
considering operation conditions to be in Mode 4. The changes are
consistent with NRC-approved Technical Specifications Task Force (TSTF)
Improved Standard Technical Specifications Change Traveler, TSTF-484,
Revision 0, ``Use of TS 3.10.1 for Scram Time Testing Activities.''
Date of issuance: August 9, 2013.
Effective date: This license amendment is effective as of the date
of its date of issuance and will be implemented within 120 days of
issuance.
Amendment No.: 174.
Renewed Facility Operating License No. DPR-22: Amendment revises
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 4, 2013.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 9, 2013.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 23rd day of August 2013.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2013-21247 Filed 8-30-13; 8:45 am]
BILLING CODE 7590-01-P