Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 54280-54294 [2013-21247]
Download as PDF
54280
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
emcdonald on DSK67QTVN1PROD with NOTICES
associated with exposure to high
concentrations of diesel particulate
matter. The standards contain
information collection requirements for
underground coal mine operators in
72.510(a) & (b), 72.520(a) & (b).
Section 72.510(a) requires
underground coal mine operators to
provide annual training to all miners
who may be exposed to diesel
emissions. The training must include
health risks associated with exposure to
diesel particulate matter; methods used
in the mine to control diesel particulate
concentrations; identification of the
personnel responsible for maintaining
those controls; and actions miners must
take to ensure controls operate as
intended.
Section 72.510(b) requires
underground coal mine operators to
keep a record of the training for one
year.
Section 72.520(a) and (b) requires
underground coal mine operators to
maintain an inventory of diesel powered
equipment units together with a list of
information about any unit’s emission
control or filtration system. The list
must be updated within 7 calendar days
of any change.
II. Desired Focus of Comments
MSHA is particularly interested in
comments that:
• Evaluate whether the collection of
information is necessary for the proper
performance of the functions of the
agency, including whether the
information has practical utility;
• Evaluate the accuracy of the
MSHA’s estimate of the burden of the
collection of information, including the
validity of the methodology and
assumptions used;
• Enhance the quality, utility, and
clarity of the information to be
collected; and
• Minimize the burden of the
collection of information on those who
are to respond, including through the
use of appropriate automated,
electronic, mechanical, or other
technological collection techniques or
other forms of information technology,
e.g., permitting electronic submission of
responses.
This information collection request is
available on MSHA’s Web site listed in
order of OMB number at https://
www.msha.gov/regs/fedreg/
informationcollection/
informationcollection.asp. The
information collection request will be
available on MSHA’s Web site for 60
days after the publication date of this
notice, and on https://
www.regulations.gov. Because
comments will not be edited to remove
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
any identifying or contact information,
MSHA cautions the commenter against
including any information in the
submission that should not be publicly
disclosed.
The public may also examine publicly
available documents at MSHA, Office of
Standards, Regulations, and Variances,
1100 Wilson Boulevard, Room 2350,
Arlington VA 22209–3939 by signing in
at the receptionist’s desk on the 21st
floor.
Questions about the information
collection requirements may be directed
to the person listed in the FOR FURTHER
INFORMATION CONTACT section of this
notice.
III. Current Actions
This request for collection of
information contains notification and
recordkeeping provisions for the
Proposed Information Collection
Request Submitted for Public Comment
and Recommendations; Health
Standards for Diesel Particulate Matter
Exposure (Underground Coal Mines) 30
CFR 72.510 and 72.520. MSHA does not
intend to publish the results from this
information collection and is not
seeking approval to either display or not
display the expiration date for the OMB
approval of this information collection.
There are no certification exceptions
identified and this information
collection and the collection of this
information does not employ statistical
methods.
Type of Review: Extension.
Agency: Mine Safety and Health
Administration.
Title: Diesel Particulate Matter
Exposure (Underground Coal. Mines) 30
CFR 72.510 and 72.520.
OMB Number: 1219–0124.
Affected Public: Business or other forprofit.
Total Number of Respondents: 206.
Frequency: On occasion.
Total Number of Responses: 53,631.
Total Burden Hours: 703 hours.
Total Annual Respondent or
Recordkeeper Cost Burden: $9.
Comments submitted in response to
this notice will be summarized and
included in the request for Office of
Management and Budget approval of the
information collection request; they will
also become a matter of public record.
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).
[FR Doc. 2013–21361 Filed 8–30–13; 8:45 am]
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
https://www.regulations.gov and search
for Docket ID NRC–2013–0201. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–287–3422;
email: Carol.Gallagher@nrc.gov. 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.
BILLING CODE 4510–43–P
SUPPLEMENTARY INFORMATION:
Dated: August 28th, 2013.
George F. Triebsch,
Certifying Officer.
PO 00000
Frm 00049
Fmt 4703
Sfmt 4703
ADDRESSES:
E:\FR\FM\03SEN1.SGM
03SEN1
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
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
https://www.regulations.gov and search
for Docket ID NRC–2013–0201.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may access publiclyavailable documents online in the NRC
Library at https://www.nrc.gov/readingrm/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 pdr.resource@nrc.gov.
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.
emcdonald on DSK67QTVN1PROD with NOTICES
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 https://
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.
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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
§ 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 60day 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
PO 00000
Frm 00050
Fmt 4703
Sfmt 4703
54281
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 https://
www.nrc.gov/reading-rm/doccollections/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
E:\FR\FM\03SEN1.SGM
03SEN1
emcdonald on DSK67QTVN1PROD with NOTICES
54282
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
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 EFiling process requires participants to
submit and serve all adjudicatory
documents over the internet, or in some
cases to mail copies on electronic
storage media. Participants may not
submit paper copies of their filings
unless they seek an exemption in
accordance with the procedures
described below.
To comply with the procedural
requirements of E-Filing, at least 10
days prior to the filing deadline, the
participant should contact the Office of
the Secretary by email at
hearing.docket@nrc.gov, or by telephone
at 301–415–1677, to request (1) a digital
information (ID) certificate, which
allows the participant (or its counsel or
representative) to digitally sign
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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 NRCissued digital ID certificate). Based upon
this information, the Secretary will
establish an electronic docket for the
hearing in this proceeding if the
Secretary has not already established an
electronic docket.
Information about applying for a
digital ID certificate is available on the
NRC’s public Web site at https://
www.nrc.gov/site-help/e-submittals/
apply-certificates.html. System
requirements for accessing the ESubmittal server are detailed in the
NRC’s ‘‘Guidance for Electronic
Submission,’’ which is available on the
agency’s public Web site at https://
www.nrc.gov/site-help/esubmittals.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 Webbased submission form, including the
installation of the Web browser plug-in,
is available on the NRC’s public Web
site at https://www.nrc.gov/site-help/esubmittals.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 https://www.nrc.gov/sitehelp/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 EFiling system time-stamps the document
and sends the submitter an email notice
confirming receipt of the document. The
PO 00000
Frm 00051
Fmt 4703
Sfmt 4703
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
https://www.nrc.gov/site-help/esubmittals.html, by email to
MSHD.Resource@nrc.gov, or by a tollfree 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 firstclass 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
E:\FR\FM\03SEN1.SGM
03SEN1
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
emcdonald on DSK67QTVN1PROD with NOTICES
available to the public at https://
ehd1.nrc.gov/ehd/, unless excluded
pursuant to an order of the Commission,
or the presiding officer. Participants are
requested not to include personal
privacy information, such as social
security numbers, home addresses, or
home phone numbers in their filings,
unless an NRC regulation or other law
requires submission of such
information. 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 https://
www.nrc.gov/reading-rm/adams.html.
Persons who do not have access to
ADAMS or who encounter problems in
accessing the documents located in
ADAMS, should contact the NRC’s PDR
Reference staff at 1–800–397–4209, 301–
415–4737, or by email to pdr.resource@
nrc.gov.
Exelon Generation Company (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®
inserts. By letter dated August 12, 2013,
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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® 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® 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
PO 00000
Frm 00052
Fmt 4703
Sfmt 4703
54283
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® 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).
E:\FR\FM\03SEN1.SGM
03SEN1
54284
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
emcdonald on DSK67QTVN1PROD with NOTICES
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 safetyrelated batteries remain capable of
performing their specified safety function.
Appropriate monitoring and maintenance
will continue to be performed on the safetyrelated batteries. In addition, the safetyrelated 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
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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.’’
PO 00000
Frm 00053
Fmt 4703
Sfmt 4703
• 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
E:\FR\FM\03SEN1.SGM
03SEN1
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
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.
emcdonald on DSK67QTVN1PROD with NOTICES
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 (°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 °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).
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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.
PO 00000
Frm 00054
Fmt 4703
Sfmt 4703
54285
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
E:\FR\FM\03SEN1.SGM
03SEN1
54286
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
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.
emcdonald on DSK67QTVN1PROD with NOTICES
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 plantspecific 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
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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
PO 00000
Frm 00055
Fmt 4703
Sfmt 4703
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.
E:\FR\FM\03SEN1.SGM
03SEN1
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
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 NonClass 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 Backup tie to the Non-Class 1E EDS Battery.
emcdonald on DSK67QTVN1PROD with NOTICES
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.
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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 safetyrelated 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
PO 00000
Frm 00056
Fmt 4703
Sfmt 4703
54287
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
E:\FR\FM\03SEN1.SGM
03SEN1
54288
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
emcdonald on DSK67QTVN1PROD with NOTICES
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.
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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
PO 00000
Frm 00057
Fmt 4703
Sfmt 4703
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
E:\FR\FM\03SEN1.SGM
03SEN1
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
emcdonald on DSK67QTVN1PROD with NOTICES
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 reclassifying the drain hubs as nonsafetyrelated 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/
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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
PO 00000
Frm 00058
Fmt 4703
Sfmt 4703
54289
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:
E:\FR\FM\03SEN1.SGM
03SEN1
54290
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
emcdonald on DSK67QTVN1PROD with NOTICES
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.
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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
PO 00000
Frm 00059
Fmt 4703
Sfmt 4703
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.
E:\FR\FM\03SEN1.SGM
03SEN1
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
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
emcdonald on DSK67QTVN1PROD with NOTICES
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-ofCoolant Accident] Mass and Energy
Release Calculation Issues.’’ The issues
identified in NSAL–11–5 affected plantspecific 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
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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
PO 00000
Frm 00060
Fmt 4703
Sfmt 4703
54291
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 asbuilt 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
E:\FR\FM\03SEN1.SGM
03SEN1
54292
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
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
emcdonald on DSK67QTVN1PROD with NOTICES
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
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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 °F to 538 °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/
PO 00000
Frm 00061
Fmt 4703
Sfmt 4703
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.
E:\FR\FM\03SEN1.SGM
03SEN1
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
NRC Branch Chief: Robert Pascarelli.
emcdonald on DSK67QTVN1PROD with NOTICES
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
https://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 pdr.resource@nrc.gov.
VerDate Mar<15>2010
17:57 Aug 30, 2013
Jkt 229001
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,
PO 00000
Frm 00062
Fmt 4703
Sfmt 4703
54293
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
E:\FR\FM\03SEN1.SGM
03SEN1
54294
Federal Register / Vol. 78, No. 170 / Tuesday, September 3, 2013 / Notices
greater than 212 °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
NUCLEAR REGULATORY
COMMISSION
[NRC–2013–0001]
Sunshine Act Notice
AGENCY HOLDING THE MEETINGS: Nuclear
Regulatory Commission.
DATES: Weeks of September 2, 9, 16, 23,
30, October 7, 2013.
PLACE: Commissioners’ Conference
Room, 11555 Rockville Pike, Rockville,
Maryland.
STATUS: Public and Closed.
emcdonald on DSK67QTVN1PROD with NOTICES
There are no meetings scheduled for
the week of September 2, 2013.
Week of September 9, 2013—Tentative
There are no meetings scheduled for
the week of September 9, 2013.
Week of September 16, 2013—Tentative
There are no meetings scheduled for
the week of September 16, 2013.
17:57 Aug 30, 2013
Week of September 30, 2013—Tentative
There are no meetings scheduled for
the week of September 30, 2013.
Week of October 7, 2013—Tentative
There are no meetings scheduled for
the week of October 7, 2013.
*
*
*
*
*
* The schedule for Commission
meetings is subject to change on short
notice. To verify the status of meetings,
call (recording)—301–415–1292.
Contact person for more information:
Rochelle Bavol, 301–415–1651.
*
*
*
*
*
The NRC Commission Meeting
Schedule can be found on the Internet
at: https://www.nrc.gov/public-involve/
public-meetings/schedule.html.
*
*
*
*
*
The NRC provides reasonable
accommodation to individuals with
disabilities where appropriate. If you
need a reasonable accommodation to
participate in these public meetings, or
need this meeting notice or the
transcript or other information from the
public meetings in another format (e.g.
braille, large print), please notify
Kimberly Meyer, NRC Disability
Program Manager, at 301–287–0727, or
by email at kimberly.meyerchambers@nrc.gov. Determinations on
requests for reasonable accommodation
will be made on a case-by-case basis.
*
*
*
*
*
This notice is distributed
electronically to subscribers. If you no
longer wish to receive it, or would like
to be added to the distribution, please
contact the Office of the Secretary,
Washington, DC 20555 (301–415–1969),
or send an email to
darlene.wright@nrc.gov.
Dated: August 28, 2013.
Rochelle C. Bavol,
Policy Coordinator, Office of the Secretary.
Meeting OPEN to the Public
from 2 p.m. to 2:15 p.m. Closed portion
will commence at 2:15 p.m. (approx.).
MATTERS TO BE CONSIDERED: 1.
President’s Report.
2. Tribute—Ambassador Demetrios J.
Marantis.
3. Tribute—Robert D. Hormats.
4. Confirmation—Michael S. Whalen
as Vice President, Structured Finance.
5. Minutes of the Open Session of the
June 13, 2013 Board of Directors
Meeting.
FURTHER MATTERS TO BE CONSIDERED:
(Closed to the Public 2:15 p.m.):
1. Finance Project—Kenya and
Tanzania.
2. Finance Project—Pakistan.
3. Finance Project—Chile.
4. Finance Project—Brazil.
5. Finance Project—Turkey.
6. Finance Project—Chile.
7. Minutes of the Closed Session of
the June 13, 2013 Board of Directors
Meeting.
8. Minutes of the August 14, 2013
Special Meeting of the Board of
Directors.
9. Minutes of the August 19, 2013
Special Meeting of the Board of
Directors.
10. Reports.
11. Pending Major Projects.
Written summaries of the projects to
be presented will be posted on OPIC’s
Web site on or about August 29, 2013.
CONTACT PERSON FOR INFORMATION:
Information on the meeting may be
obtained from Connie M. Downs at (202)
336–8438.
STATUS:
Dated: August 29, 2013.
Connie M. Downs,
Corporate Secretary, Overseas Private
Investment Corporation.
[FR Doc. 2013–21450 Filed 8–29–13; 4:15 pm]
BILLING CODE 3210–01–P
SECURITIES AND EXCHANGE
COMMISSION
Investment Company Act Release No.
30679; File No. 812–14167
[FR Doc. 2013–21465 Filed 8–29–13; 4:15 pm]
Jkt 229001
BILLING CODE 7590–01–P
Franklin Templeton International Trust,
et al.; Notice of Application
OVERSEAS PRIVATE INVESTMENT
CORPORATION
Week of September 2, 2013
VerDate Mar<15>2010
Week of September 23, 2013—Tentative
There are no meetings scheduled for
the week of September 23, 2013.
August 27, 2013.
Sunshine Act Meeting; Board of
Directors Meeting
Thursday, September 19,
2013, 2 p.m. (OPEN Portion) 2:15 p.m.
(CLOSED Portion).
PLACE: Offices of the Corporation,
Twelfth Floor Board Room, 1100 New
York Avenue NW., Washington, DC.
TIME AND DATE:
PO 00000
Frm 00063
Fmt 4703
Sfmt 4703
Securities and Exchange
Commission (‘‘Commission’’).
ACTION: Notice of an application under
section 6(c) of the Investment Company
Act of 1940 (‘‘Act’’) for an exemption
from section 15(a) of the Act and rule
18f–2 under the Act, as well as from
certain disclosure requirements.
AGENCY:
Applicants
request an order that would permit them
SUMMARY OF APPLICATION:
E:\FR\FM\03SEN1.SGM
03SEN1
Agencies
[Federal Register Volume 78, Number 170 (Tuesday, September 3, 2013)]
[Notices]
[Pages 54280-54294]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-21247]
=======================================================================
-----------------------------------------------------------------------
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 https://www.regulations.gov and search for Docket ID NRC-2013-0201. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov. 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 https://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 https://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 pdr.resource@nrc.gov. 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 https://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 https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is
filed by the above date, the Commission or a presiding officer
designated by the Commission or by the Chief Administrative Judge of
the Atomic Safety and Licensing Board Panel, will rule on the request
and/or petition; and the Secretary or the Chief Administrative Judge of
the Atomic Safety and Licensing Board will issue a notice of a hearing
or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the 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 hearing.docket@nrc.gov, 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 https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at https://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC'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 https://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with the NRC's guidance
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC'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 https://www.nrc.gov/site-help/e-submittals.html, by email to
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants. Filing is considered
complete by first-class mail as of the time of deposit in the mail, or
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is
[[Page 54283]]
available to the public at https://ehd1.nrc.gov/ehd/, unless excluded
pursuant to an order of the Commission, or the presiding officer.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or home phone numbers
in their filings, unless an NRC regulation or other law requires
submission of such information. 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 https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209,
301-415-4737, or by email to pdr.resource@nrc.gov.
Exelon Generation Company (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 https://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
pdr.resource@nrc.gov.
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