[Federal Register Volume 78, Number 122 (Tuesday, June 25, 2013)]
[Notices]
[Pages 38078-38087]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-14880]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0134]
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 May 30, 2013 to June 12, 2013. The last
biweekly notice was published on June 11, 2013 (78 FR 35058).
ADDRESSES: You may submit comment by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0134. Address
questions about NRC dockets to Carol
[[Page 38079]]
Gallagher; telephone: 301-492-3668; email: [email protected]. For
technical questions, contact the individual(s) listed in the FOR
FURTHER INFORMATION CONTACT section of this document.
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2013-0134 when contacting the NRC
about the availability of information regarding this document. You may
access information related to this document, which the NRC possesses
and is publicly-available, by the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0134.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly-available documents online in the NRC
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in
ADAMS by performing a search on the document date and docket number.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2013-0134 in the subject line of your
comment submission, in order to ensure that the NRC is able to make
your comment submission available to the public in this docket.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment submissions into ADAMS.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Section 50.92 of Title 10 of the Code
of Federal Regulations (10 CFR), this means that operation of the
facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license or
combined license. Requests for a hearing and a petition for leave to
intervene shall be filed in accordance with the Commission's ``Agency
Rules of Practice and Procedure'' in 10 CFR part 2. Interested
person(s) should consult a current copy of 10 CFR 2.309, which is
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The
NRC regulations are accessible electronically from the NRC Library on
the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is
filed by the above date, the Commission or a presiding officer
designated by the Commission or by the Chief Administrative Judge of
the Atomic Safety and Licensing Board Panel, will rule on the request
and/or petition; and the Secretary or the Chief Administrative Judge of
the Atomic Safety and Licensing Board will issue a notice of a hearing
or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The
[[Page 38080]]
petition must also identify the specific contentions which the
requestor/petitioner seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139;
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to request (1) a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a request or
petition for hearing (even in instances in which the participant, or
its counsel or representative, already holds an NRC-issued digital ID
certificate). Based upon this information, the Secretary will establish
an electronic docket for the hearing in this proceeding if the
Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with the NRC guidance
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC's Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email at
[email protected], or by a toll-free call at 1-866-672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North,
[[Page 38081]]
11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking
and Adjudications Staff. Participants filing a document in this manner
are responsible for serving the document on all other participants.
Filing is considered complete by first-class mail as of the time of
deposit in the mail, or by courier, express mail, or expedited delivery
service upon depositing the document with the provider of the service.
A presiding officer, having granted an exemption request from using E-
Filing, may require a participant or party to use E-Filing if the
presiding officer subsequently determines that the reason for granting
the exemption from use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. With respect to copyrighted works, except
for limited excerpts that serve the purpose of the adjudicatory filings
and would constitute a Fair Use application, participants are requested
not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the following three
factors in 10 CFR 2.309(c)(1): (i) The information upon which the
filing is based was not previously available; (ii) the information upon
which the filing is based is materially different from information
previously available; and (iii) the filing has been submitted in a
timely fashion based on the availability of the subsequent information.
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS, should contact the NRC's PDR Reference staff at 1-800-397-4209,
301-415-4737, or by email to [email protected].
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit 3 (MPS-3), New London County, Connecticut
Date of amendment request: April 25, 2013.
Description of amendment request: The amendments would revise the
peak calculated containment internal pressure (Pa) for the
design basis loss of coolant accident (LOCA) described in Technical
Specification (TS) 6.8.4.f, ``Containment Leakage Rate Testing
Program'' for MPS-3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to Pa does not alter the assumed
initiators to any analyzed event. The probability of an accident
previously evaluated will not be significantly increased by this
proposed change.
The change in Pa will not affect radiological dose
consequence analyses. MPS-3 radiological dose consequence analyses
assume a certain containment atmosphere leak rate based on the
maximum allowable containment leakage rate, which is not affected by
the change in peak calculated containment internal pressure. The
Appendix J containment leakage rate testing program will continue to
ensure that containment leakage remains within the leakage assumed
in the offsite dose consequence analyses. The consequences of an
accident previously evaluated will not be significantly increased by
this proposed change.
Therefore, operation of the facility in accordance with the
proposed change to Pa will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change provides a higher Pa than
currently described in TS 6.8.4.f. This change is a result of an
increase in the M&E [mass and energy] release input for the LOCA
containment response analysis. The [Pa] remains below the
containment design pressure of 45 psig [pounds per square inch
gauge]. This change does not involve any alteration in the plant
configuration (no new or different type of equipment will be
installed) or make changes in the methods governing normal plant
operation. The change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Therefore, operation of the facility in accordance with the
proposed change to TS 6.8.4.f would not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The [Pa] remains below the containment design
pressure of 45 psig. Since the MPS3 radiological consequence
analyses are based on the maximum allowable containment leakage
rate, which is not being revised, the change in the [Pa]
does not represent a significant change in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
Acting NRC Branch Chief: Robert Beall.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: April 16, 2013.
Description of amendment request: The proposed amendments would
remove superseded Technical Specification (TS) requirements McGuire
Nuclear Station (MNS), Units 1 and 2. By letter dated May 28, 2010,
Duke Energy submitted a license amendment request (LAR) to modify TS to
allow the manual operation of the Containment Spray System in lieu of
automatic actuation, and revise the minimum volume and low level
setpoint on the Refueling Water Storage Tank. Because the associated
modifications were implemented on a staggered basis for each MNS Unit
during refueling outages, the deletion or modification of these TS
requirements was accomplished via the use of temporary footnotes. This
allowed the
[[Page 38082]]
requirements to be either applicable or non-applicable, depending upon
whether the modifications had not been implemented or implemented,
respectively. The LAR contained a commitment for MNS to submit a
follow-up administrative license amendment request to delete the
superseded temporary TS requirements within 180 days of the
installation of the associated modifications for the final MNS Unit. By
letter dated September 12, 2011, the NRC issued amendments regarding
the TS changes requested in the May 28, 2010 LAR. Installation of the
associated modifications on the final MNS Unit was completed on October
18, 2012. This LAR satisfies the MNS commitment to delete the
superseded temporary TS requirements described in the May 28, 2010 LAR.
In addition, this LAR makes an administrative non-technical editorial
correction by relocating NOTE 1 on TS page 3.3.2-15 to TS page 3.3.2-
14. Relocating NOTE 1 back to TS page 3.3.2-14 is consistent with the
reference to this NOTE in TS Table 3.3.2-1, Engineered Safety Feature
Actuation System (ESFAS) Instrumentation, Function 9, Containment
Pressure Control System.
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:
Criterion 1:
Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This LAR proposes administrative non-technical changes only.
These proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configurations of the facility. The proposed changes do 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.
Given the above discussion, it is concluded the proposed
amendment does not significantly increase the probability or
consequences of an accident previously evaluated.
Criterion 2:
Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This LAR proposes administrative non-technical changes only. The
proposed changes will not alter the design requirements of any SSC
or its function during accident conditions. No new or different
accidents result from the changes proposed. The changes do not
involve a physical alteration of the plant or any changes in methods
governing normal plant operation. The changes do not alter
assumptions made in the safety analysis.
Given the above discussion, it is concluded the proposed
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
Criterion 3:
Does the proposed amendment involve a significant reduction in
the margin of safety?
Response: No.
This LAR proposes administrative non-technical changes only. The
proposed changes do not alter the manner in which safety limits,
limiting safety system settings or limiting conditions for operation
are determined. The safety analysis acceptance criteria are not
affected by these changes. The proposed changes will not result in
plant operation in a configuration outside the design basis. The
proposed changes do not adversely affect systems that respond to
safely shutdown the plant and to maintain the plant in a safe
shutdown condition.
Given the above discussion, it is concluded the proposed
amendment does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC
28202.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit 2, Westchester County, New York
Date of amendment request: April 15, 2013.
Description of amendment request: The proposed change would revise
Technical Specification 3.5.4, ``Refueling Water Storage Tank (RWST)''
such that the non-seismically qualified piping of the temporary Boric
Acid Recovery System (BARS) may be connected to the seismic piping of
the RWST. Operation of the BARS from the RWST will be under
administrative controls for a limited period of time (i.e., 30 days for
RWST filtration prior to each fuel cycle). This change will only be
applicable until Refueling Outage R22 ends (Spring 2016).
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 use of the non seismic Boric Acid Recovery System (BARS) to
recirculate and filter the Refueling Water Storage Tank (RWST) water
does not involve any changes or create any new interfaces with the
reactor coolant system or main steam system piping. Therefore, the
connection of the BARS Purification Loop to the RWST would not
affect the probability of these accidents occurring. The BARS is not
credited for safe shutdown of the plant or accident mitigation.
Administrative controls ensure that the BARS can be isolated as
necessary and in sufficient time to assure that the RWST volume will
be adequate to perform the safety function as designed. Since the
RWST will continue to perform its safety function and overall system
performance is not affected, the consequences of the accident are
not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The design of the RWST and the SFP [spent fuel pool]
Purification Loop has been revised to allow recirculation and
purification using the BARS for a short period of time (not to
exceed 30 days per fuel cycle) for the next two fuel cycles. The
added BARS takes RWST water in and processes it out without
additional connections that could affect other systems and without
an impact from its installation. Procedures for the operation of the
plant, including BARs, will not create the possibility of a new or
different type of accident. Contingent upon manual operator action,
a BARS line break will not result in a loss of the RWST safety
function. Similarly, an active or passive failure in the BARS will
not affect safety related structures, systems or components.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The SFP Purification Loop and recirculation and purification of
the RWST water using the BARS is not credited for safe shutdown of
the plant or accident mitigation. RWST volume will be maximized
prior to purification and timely operator action can be taken to
isolate the non seismic system from the RWST to assure it can
perform its function. This will result in no significant reduction
in the margin of safety.
Therefore the proposed change does not significantly reduce the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 38083]]
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
Acting NRC Branch Chief: Sean Meighan.
National Institute of Standards and Technology (NIST), Docket No. 50-
184, Center for Neutron Research (NBSR), Montgomery County, Maryland
Date of amendment request: July 12, 2012, as supplemented on May
14, 2013.
Description of amendment request: The proposed amendments would
revise NIST NBSR's Technical specifications, Sections 3.7, 4.7, and
6.8, pertaining to the environmental monitoring requirements and
records retention which clarifies environmental sampling procedure and
record retention processes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment corrects a deficiency in the license
issued in 2009 that created a disagreement in the periodicity of
environmental sampling within the license Technical Specifications.
Additionally, the proposed amendment aligns the record retention
requirement (section 6.8) of the license technical specifications
with the consensus standard ANSI/ANS 15.1. This standard has been
endorsed by the NRC under Regulatory Guide 2.2. Neither of these
proposed changes will have any influence or impact on reactor
operations or previously analyzed accidents. There are no physical
changes to the facility as a result of these administrative changes.
Therefore, the proposed 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.
No accident of any kind would be created by the proposed
administrative changes. The sample periodicity will not change from
the sampling periodicity used by the facility for over 40 years.
Records are maintained and summarized in facility annual reports and
there would be no loss of information. There are no physical changes
to the facility as a result of these administrative changes.
Therefore, the changes would not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Plant safety margins are established through limiting conditions
of operation, limiting safety system settings, and safety limits
specified in the Technical Specifications. The proposed changes do
not alter any of the established safety margins and are
administrative in nature.
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: Melissa J. Lieberman, Deputy Chief Counsel
for NIST, National Institute of Standard and Technology, 100 Bureau
Drive, Gaithersburg, MD 20899.
NRC Branch Chief: Alexander Adams, Jr.
South Carolina Electric and Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1,
Fairfield County, South Carolina
Date of amendment request: April 2, 2013, as supplemented by a
letter dated May 16, 2013.
Description of amendment request: The proposed amendments would
revise the technical specification requirements regarding steam
generator tube inspection and reporting as described in Technical
Specification Task Force (TSTF)-510, ``Revision to Steam Generator
Program Inspection Frequencies and Tube Sample Selection,'' Revision 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Steam Generator (SG) Program to
modify the frequency of verification of SG tube integrity and SG
tube sample selection. A steam generator tube rupture (SGTR) event
is one of the design basis accidents that are analyzed as part of a
plant's licensing basis. The proposed SG tube inspection frequency
and sample selection criteria will continue to ensure that the SG
tubes are inspected such that the probability of a SGTR is not
increased. The consequences of a SGTR are bounded by the
conservative assumptions in the design basis accident analysis. The
proposed change will not cause the consequences of a SGTR to exceed
those assumptions.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the Steam Generator Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
proposed change does not affect the design of the SGs or their
method of operation. In addition, the proposed change does not
impact any other plant system or component.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
also isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a SG
is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change will continue to require monitoring of the physical
condition of the SG tubes such that there will not be a reduction in
the margin of safety compared to the current requirements.
Therefore, it is concluded that 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: J. Hagood Hamilton, Jr., South Carolina
Electric &
[[Page 38084]]
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Branch Chief: Robert J. Pascarelli.
South Carolina Electric and Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1,
Fairfield County, South Carolina
Date of amendment request: April 3, 2013.
Description of amendment request: The proposed amendment would
allow for the extension of the frequency of the containment leak rate
test per Technical Specification 6.8.4(g) from 130-months (10.9 years)
to 15 years.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed exemption involves a permanent 15-year extension to
the current interval for Type A containment testing. The current
test interval of 130 months (10.9 years) would be extended to a
permanent 15-year frequency from the last Type A test. The proposed
extension does not involve a physical change to the plant or a
change in the manner in which the plant is operated or controlled.
The containment is designed to provide an essentially leak tight
barrier against the uncontrolled release of radioactivity to the
environment for postulated accidents. As such, the reactor
containment itself and the testing requirements invoked to
periodically demonstrate the integrity of the reactor containment
exist to ensure the plant's ability to mitigate the consequences of
an accident, and do not involve the prevention or identification of
any precursors of an accident. Therefore, this proposed extension
does not involve a significant increase in the probability of an
accident previously evaluated nor does it create the possibility of
a new or different kind of accident.
The integrity of the reactor containment is subject to two types
of failure mechanisms which can be categorized as (1) activity based
and (2) time based. Activity based failure mechanisms are defined as
degradation due to system and/or component modifications or
maintenance. Local leak rate test requirements and administrative
controls such as configuration management and procedural
requirements for system restoration ensure that containment
integrity is not degraded by plant modifications or maintenance
activities. The design and construction requirements of the
containment itself combined with the containment inspections
performed in accordance with ASME, Section XI, the Maintenance Rule,
and Licensing commitments serve to provide a high degree of
assurance that the containment will not degrade in a manner that is
detectable only by a Type A test.
Based on the above, the proposed extension does not involve a
significant increase in the consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed revision to the TS involves a 15-year permanent
extension to the current interval for Type A containment testing.
The reactor containment and the testing requirements invoked to
periodically demonstrate the integrity of the reactor containment
exist to ensure the plant's ability to mitigate the consequences of
an accident and do not involve the prevention or identification of
any precursors of an accident. The proposed TS change does not
involve a physical change to the plant or the manner in which the
plant is operated or controlled.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to the TS involves a 15-year permanent
extension to the current interval for Type A containment testing.
The proposed TS change does not involve a physical change to the
plant or a change in the manner in which the plant is operated or
controlled. The specific requirements and conditions of the Primary
Containment Leak Rate Testing Program, as defined in the TS, exist
to ensure that the degree of reactor containment structural
integrity and leak-tightness that is considered in the plant safety
analysis is maintained. The overall containment leak rate limit
specified by TS is maintained. The proposed change involves only the
extension of the interval between Type A containment leak rate
tests. The proposed surveillance interval extension is bounded by
the 15-year permanent extension currently authorized within NEI 94-
01, Revision 3-A. Type B and C containment leak rate tests will
continue to be performed at the frequency currently required by TS.
Industry experience supports the conclusion that Type B and C
testing detects a large percentage of containment leakage paths and
that the percentage of containment leakage paths that are detected
only by Type A testing is small. The containment inspections
performed in accordance with ASME, Section Xl and the Maintenance
Rule serve to provide a high degree of assurance that the
containment will not degrade in a manner that is detectable only by
Type A testing.
The combination of these factors ensures that the margin of
safety that is in plant safety analysis is maintained. The design,
operation, testing methods and acceptance criteria for Type A, B,
and C containment leakage tests specified in applicable codes and
standards will continue to be met, with the acceptance of this
proposed change, since these are not affected by changes to the Type
A test interval. Therefore, the proposed TS change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. Hagood Hamilton, Jr., South Carolina
Electric & Gas Company, Post Office Box 764, Columbia, South Carolina
29218.
NRC Branch Chief: Robert J. Pascarelli.
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: May 10, 2013.
Description of amendment request: The proposed change would amend
Combined Licenses Nos.: NPF-91 and NPF-92 for Vogtle Electric
Generating Plant (VEGP) Units 3 and 4 by departing from VEGP Units 3
and 4 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by
revising reference document APP-OCS-GEH-520, ``AP1000 Plant Startup
Human Factors Engineering Design Verification Plan,'' from Revision B
to Revision 1. APP-OCS-GEH-520 is incorporated by reference in the
Updated Final Safety Analysis Report (UFSAR) as a means to implement
the activities associated with the human factors engineering
verification and validation.
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 APP-OCS-GEH-520, document confirms aspects of the human
system interface (HSI) and Operation and Control Centers Systems
(OCS) design features that could not be evaluated in other Human
Factors Engineering (HFE) verification and validation (V&V)
activities. It also confirms that the as-built in the plant HSIs,
procedures, and training conform to the design that resulted from
the HFE program. Additionally, it confirms that all HFE-related
issues (including human error discrepancies (HEDs)) documented in
the SmartPlant Foundation (SPF) Human Factors (HF)
[[Page 38085]]
Tracking System are verified as adequately addressed or resolved.
Finally, it confirms the HFE adequacy for risk-important human
actions in the local plant, including the ability for the tasks to
be completed within the time window according to the Probabilistic
Risk Assessment (PRA). The changes to the plan are to clarify the
scope and amend the details of the methodology. The plan does not
affect the plant itself. Changing the plan does not affect
prevention and mitigation of abnormal events, e.g., accidents,
anticipated operational occurrences, earthquakes, floods and turbine
missiles, or their safety or design analyses. The PRA is not
affected. No safety-related Structure, System, or Component (SSC) or
function is adversely affected. The document revision change does
not involve nor interface with any SSC accident initiator or
initiating sequence of events, and thus, the probabilities of the
accidents evaluated in the Updated Final Safety Analysis Report
(UFSAR) are not affected. Because the changes to the plan do not
involve any safety-related SSC or function used to mitigate an
accident, the consequences of the accidents evaluated in the UFSAR
are not affected.
Therefore, there is no 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.
APP-OCS-GEH-520, ``AP1000 Plant Startup Human Factors
Engineering Design Verification Plan'' is the plan to confirm
aspects of the HSI and OCS design features that could not be
evaluated in other HFE V&V activities. The plan also confirms that
the as-built in the plant HSIs, procedures, and training conform to
the design that resulted from the HFE program. Additionally, it
confirms that all HFE-related issues (including HEDs) documented in
the SPF HF Tracking System are verified as adequately addressed or
resolved. Finally, it confirms the HFE adequacy for risk-important
human actions in the local plant, including the ability for the
tasks to be completed within the time window according to the PRA.
These functions support evaluating the HSI and OCS. Therefore, the
changes do not affect the safety-related equipment itself, nor do
they affect equipment which, if it failed, could initiate an
accident or a failure of a fission product barrier. No analysis is
adversely affected. No system or design function or equipment
qualification will be adversely affected by the changes. This
activity will not allow for a new fission product release path, nor
will it result in a new fission product barrier failure mode, nor
create a new sequence of events that would result in significant
fuel cladding failures. In addition, the changes do not result in a
new failure mode, malfunction or sequence of events that could
affect safety or safety-related equipment.
Therefore, this activity does not create the possibility of a
new or different kind of accident than any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
APP-OCS-GEH-520, ``AP1000 Plant Startup Human Factors
Engineering Design Verification Plan'' is the plan to confirm
aspects of the HSI and OCS design features that could not be
evaluated in other HFE V&V activities. The plan also confirms that
the as-built in the plant HSIs, procedures, and training conform to
the design that resulted from the HFE program. Additionally, it
confirms that all HFE-related issues (including HEDs) documented in
the SPF HF Tracking System are verified as adequately addressed or
resolved. Finally, it confirms the HFE adequacy for risk-important
human actions in the local plant, including the ability for the
tasks to be completed within the time windows in the PRA. These
functions support evaluating the HSI and OCS. The proposed changes
to the plan do not affect the design or operation of safety-related
equipment or equipment whose failure could initiate an accident, nor
does the plan adversely affect the interfaces with safety-related
equipment or fission product barriers. No safety analysis or design
basis acceptance limit/criterion is challenged or exceeded by the
requested changes.
Therefore, the 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, Blach & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
Acting NRC Branch Chief: Lawrence Burkhart.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: April 25, 2013.
Description of amendment request: The amendments would revise
Technical Specification (TS) 5.1, ``Site,'' Figures 5.1-1 through 5.1-4
for South Texas Project (STP), Units 1 and 2, to remove identification
of a Visitor's Center building, which has been demolished. The
amendments also would revise Figures 5.1-1, 5.1-3, and 5.1-4 to remove
references to the Emergency Operations Facility (EOF) within the
Nuclear Training Facility, since the EOF was relocated to Center of
Energy Development building located in Bay City, Texas, approximately
12.5 air miles from the plant site in 2009. The EOF was relocated
offsite with an emergency plan change made by the licensee under 10 CFR
50.54(q), ``Emergency plans,'' by concluding that the change did not
represent a decrease in effectiveness of the emergency plan. The
amendments to remove references to the Visitor's Center Building and
EOF from the TSs are administrative in nature.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change is an administrative change to STP TS design
features to remove reference to the Visitor's Center and onsite EOF.
The design function of structures, systems and components (SSC)
important to safety are not impacted by the proposed change. The
proposed change will not initiate an event. The proposed change does
not alter or prevent the ability of SSCs from performing their
intended function to mitigate the consequences of an initiating
event.
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 is an administrative change to STP TS design
features to remove reference to the Visitor's Center and onsite EOF.
The proposed change does not impact create the possibility of a new
or different kind of accident from any accident previously
evaluated. There are no new failure modes or mechanisms associated
with the proposed change. This change does not involve any
modification in operational limits or physical design of equipment
important to safety.
Therefore, the proposed change does not involve a 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 is an administrative change to STP TS design
features to remove reference to the Visitor's Center and onsite EOF.
The proposed change does not impact TS safety limits, TS limiting
safety system set points, or the results of any of the safety
analyses.
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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that
[[Page 38086]]
the request for amendments involves no significant hazards
consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
Previously Published Notices 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 following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: May 22, 2013.
Brief description of amendment request: The proposed amendment
would revise the WBN Unit 1 Technical Specifications (TSs) to allow a
one-time extension to the Completion Time for TS Limiting Condition for
Operation (LCO) 3.6.6 Required Action A.1 from 72 hours to 7 days for
an inoperable Containment Spray (CS) Train B. This change is necessary
to provide sufficient time to replace a leaking mechanical seal on CS
Pump 1B-B. The pump repair is currently scheduled for the week of June
24, 2013. TVA requested this TS change under exigent circumstances and
that the NRC expedites the review to support approval by June 22, 2013.
Date of publication of individual notice in Federal Register: June
3, 2013 (78 FR 33117).
Expiration date of individual notice: June 17, 2013 (public
comments); August 2, 2013 (hearing requests).
Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the NRC's Public Document Room (PDR), located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through the Agencywide Documents Access and
Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to
[email protected].
Carolina Power and Light Company, et al., Docket No. 50-261, H.B.
Robinson Steam Electric Plant, Unit 2, Darlington County, South
Carolina
Date of application for amendment: September 6, 2012, as
supplemented by letter dated December 7, 2012.
Brief Description of amendment: The amendment revised the Technical
Specifications (TSs) to eliminate Function 14, Steam Generator Water
Level-Low Coincident with Steam Flow/Feedwater Flow Mistmatch, from the
HBRSEP TS Table 3.3.1-1, ``Reactor Protection System Instrumentation.''
Date of issuance: May 29, 2013.
Effective date: As of date of issuance and shall be implemented
prior exiting the scheduled fall 2013 refueling outage.
Amendment No.: 234.
Renewed Facility Operating License No. DPR-23: Amendment changed
the license and TSs.
Date of initial notice in Federal Register: November 27, 2012 (77
FR 70840). The supplement dated December 7, 2012, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 29, 2013.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit 2, New London County, Connecticut
Date of amendment request: July 21, 2010.
Description of amendment request: The proposed amendment revised
the Technical Specification (TS) 3/4.9.3.1, ``Decay Time'' for
Millstone Power Station, Unit 2 (MPS2). The proposed change revises TS
3/4.9.3.1 by reducing the minimum decay time for irradiated fuel prior
to movement in the reactor vessel from 150 hours to 100 hours. The
licensee requested a reduction in the minimum decay time requirement to
provide additional flexibility in outage planning such that irradiated
fuel can be moved from the reactor vessel to the spent fuel pool
earlier in an outage.
Date of issuance: June 4, 2013.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 315.
Renewed Facility Operating License No. DPR-65: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: April 2, 2013 (78 FR
19749). The supplemental letter dated July 19, 2011, 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
[[Page 38087]]
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 4, 2013.
No significant hazards consideration comments received: No.
South Carolina Electric and Gas. Docket Nos. 52-027 and 52-028, Virgil
C. Summer Nuclear Station (VCSNS), Units 3 and 4, Fairfield County,
South Carolina
Date of amendment request: February 14, 2013.
Brief description of amendment: The amendment authorizes a
departure from the Virgil C. Summer Nuclear Station, Units 2 and 3
plant-specific Design Control Document (DCD) material incorporated into
the Updated Final Safety Analysis Report (UFSAR) to revise Figure
3.8.8-1, Sheet 1, Note 2.
Date of issuance: May 23, 2013.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: Unit 2-3, and Unit 3-3.
Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised
the Facility Combined Licenses.
Date of initial notice in Federal Register: March 4, 2013 (78 FR
14126).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 23, 2013.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 14th day of June 2013.
For The Nuclear Regulatory Commission.
John D. Monninger,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2013-14880 Filed 6-24-13; 8:45 am]
BILLING CODE 7590-01-P