[Federal Register Volume 78, Number 93 (Tuesday, May 14, 2013)]
[Notices]
[Pages 28248-28258]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-11272]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0084]
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 (the
Commission or NRC) is publishing this regular biweekly notice. The Act
requires the
[[Page 28249]]
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 April 18, 2013 to May 1, 2013. The last
biweekly notice was published on April 30, 2013 (78 FR 25310).
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-2013-0084. Address
questions about NRC dockets to Carol 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.
Fax comments to: RADB at 301-492-3446.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2013-0084 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-2013-0084.
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-2013-0084 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
[[Page 28250]]
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC 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
[[Page 28251]]
contacting the NRC Meta System Help Desk through the ``Contact Us''
link located on the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html, by email at [email protected], or by a toll-free
call at 1-866 672-7640. The NRC Meta System Help Desk is available
between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday,
excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants. Filing is considered
complete by first-class mail as of the time of deposit in the mail, or
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. With respect to copyrighted works, except
for limited excerpts that serve the purpose of the adjudicatory filings
and would constitute a Fair Use application, participants are requested
not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the following three
factors in 10 CFR 2.309(c)(1): (i) The information upon which the
filing is based was not previously available; (ii) the information upon
which the filing is based is materially different from information
previously available; and (iii) the filing has been submitted in a
timely fashion based on the availability of the subsequent information.
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS, should contact the NRC PDR Reference staff at 1-800-397-4209,
301-415-4737, or by email to [email protected].
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina;
and Docket Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and
2, Mecklenburg County, North Carolina
Date of amendment request: January 21, 2013.
Description of amendment request: The amendments would revise the
divider barrier seal test coupons' tensile strength in Technical
Specification Surveillance Requirement 3.6.14.4 from ``> 39.7 psi'' to
``> 39.7 lbs.'' This change is an administrative change to correct an
error where the wrong units were used when Catawba and McGuire
converted to Standard Technical Specifications in 1998 using NUREG-
1431, Revision 1.
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.
Divider barrier integrity is necessary to minimize bypassing of
the ice condenser by the hot steam and air mixture released into the
lower compartment during a Design Basis Accident (DBA). This ensures
that most of the gases pass through the ice bed, which condenses the
steam and limits pressure and temperature during the accident
transient. Limiting the pressure and temperature reduces the release
of fission product radioactivity from containment to the environment
in the event of a DBA.
Conducting periodic physical property tests on divider barrier
seal test coupons provides assurance that the seal material has not
degraded in the containment environment, including the effects of
irradiation with the reactor at power. The proposed change to
Technical Specification Surveillance Requirement 3.6.14.4 results in
the correct tensile strength units being listed in this surveillance
requirement. This is considered an editorial change to the Technical
Specifications.
Thus, based on the above, 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 change does not involve a change in the operational
limits or the design capabilities of the containment or containment
systems. The proposed change does not change the function or
operation of plant equipment or introduce any new failure
mechanisms. The technical evaluation that supports this License
Amendment Request included a review of the containment divider
barrier seal capability to which this change is bounded. The
proposed change does not introduce any new or different types of
failure mechanisms; plant equipment will continue to respond as
designed and analyzed.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. The performance of the fuel cladding, the reactor coolant
system and the containment system will not be adversely impacted by
the proposed change since the ability of the divider barrier to
mitigate an analyzed accident has not been adversely impacted by the
proposed change.
Thus, it is concluded that the proposed change does not involve
a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 28252]]
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.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: April 9, 2013.
Description of amendment request: The proposed amendment would
delete certain reporting requirements contained in the Technical
Specifications.
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 changes do not involve the modification of any
plant equipment or affect plant operation. The proposed changes will
have no impact on any safety related structures, systems, or
components. The reporting requirements proposed for deletion are not
required because the requirements are adequately addressed by 10 CFR
50.72 and 10 CFR 50.73, or other regulatory requirements, or are
available on site for NRC review, and are no longer warranted.
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 have no impact on the design, function or
operation of any plant structure, system or component. The proposed
changes do not affect plant equipment or accident analyses. The
reporting requirements proposed for deletion are not required
because the requirements are adequately addressed by 10 CFR 50.72
and 10 CFR 50.73, or other regulatory requirements, or are available
on site for NRC review, and are no longer warranted.
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 proposed changes do not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analyses. There is no change being made to
safety analysis assumptions, safety limits or limiting safety system
settings that would adversely affect plant safety as a result of the
proposed changes. Margins of safety are unaffected by deletion of
the reporting requirements.
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: J. Bradley Fewell, Esquire, Associate
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Meena K. Khanna.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Units 1 and 2, Salem County, New Jersey
Date of amendment request: November 30, 2012.
Description of amendment request: The proposed amendment would
revise the Emergency Plan to remove references to the backup plant vent
extended range noble gas radiation monitoring (R45) indication,
recording, and alarm capability in the emergency response facilities.
The R45 indicators have become obsolete and unreliable. The R45 is a
backup to the R41 for plant vent intermediate and high range noble gas
radiation monitoring indicators. The accident sampling function of the
R45 will be maintained.
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 plant vent noble gas indicators are not an initiator of or a
precursor to any accident or transient. The proposed change to the
Emergency Plan to delete the backup (R45) noble gas indicators does
not impact any design function of the Salem Radiation Monitoring
System. The backup (R45) plant vent radiation monitors do not
perform any accident mitigating function. The modification of the
R45 noble gas indicators does not alter or modify the function of
systems used to mitigate the consequences of any design basis
accident.
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 proposed changes to the Emergency Plan to delete the backup
plant vent noble gas indicators (R45) does not introduce any new
accident precursors and does not involve any physical plant
alterations or changes in the methods governing normal plant
operation that could initiate a new or different kind of accident.
The R45 noble gas indicators only provide indication of the effluent
release through the plant vent release path.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is related to the ability of the fission
product barriers (fuel cladding, reactor coolant system, and primary
containment) to perform their design functions during and following
postulated accidents. The proposed amendment does not alter
setpoints or limits established or assumed by the accident analyses.
The R45 plant vent radiation monitor provides indication only. The
elimination of the backup R45 noble gas indicator does not reduce
the margin of safety since the remaining R41 noble gas indicator
will continue to provide the accident indication capability. The
accident sampling capability of the R45 will remain.
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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Meena K. Khanna.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: February 28, 2013.
Description of amendment request: The proposed amendment would
revise Technical Specification Section 3.6.5 by adding a different
limitation on the
[[Page 28253]]
containment average air temperature. The revised Technical
Specification Section 3.6.5 would read as follows:
``Containment average air temperature shall be
<125[emsp14][deg]F.''
To support this proposed change, the licensee revised the accident
analyses that were impacted by the increase in initial containment air
temperature or increase in safety injection accumulator temperature,
which are located in the Ginna containment, and are expected to be at
the same temperature as containment air. The impact of the change in
the containment air temperature was addressed by revising the Loss of
Coolant Accident (LOCA) and a Main Steam Line break containment
response analyses. The change in SI accumulator temperature was
reflected in the re-evaluated core response to a large break LOCA
(LBLOCA) and a small break LOCA. The combined impact on the post-LOCA
long term cooling analyses was also re-assessed.
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 increase the containment average air
temperature limit to 125[emsp14][deg]F, from 120[emsp14][deg]F, does
not alter the assumed initiators to any analyzed event. The
probability of an accident previously evaluated will not be
increased by this proposed change. This proposed change will not
affect radiological dose consequence analyses. The 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 allowable average containment air
temperature resulting in a higher calculated peak containment
pressure. The 10 CFR Part 50, Appendix J containment leak rate
testing program will continue to ensure that containment leakage
remains within the leakage rate assumed in the offsite dose
consequence analyses. The acceptable leakage corresponds to the peak
allowable containment pressure of 60 psig. The radiological dose
consequence analyses assume a certain source term, which is not
affected by the change in allowable average containment air
temperature. All core limitations set forth in 10 CFR 50.46 continue
to be met. The consequences of an accident previously evaluated will
not be increased by this proposed change.
Therefore, operation of the facility in accordance with the
proposed change to the containment average air temperature limit
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 for a higher allowable containment
average air temperature to that currently in the TS Section 3.6.5.
The calculated peak containment temperature and pressure remain
below the containment design temperature and pressure of
286[emsp14][deg]F and 60 psig. 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 Section 3.6.5 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 calculated peak containment pressure and temperature remain
below the containment design pressure and temperature of 60 psig and
286[emsp14][deg]F, respectively. The penalties applied to the BE
[best estimate] LBLOCA analysis result in the limitations set forth
in 10 CFR 50.46 continuing to be met. Since the radiological
consequence analyses are based on the maximum allowable containment
leakage rate, which is not being revised, the change in the
calculated peak containment pressure and temperature and changes in
core response do not represent a significant change in the margin of
safety. The longterm impact of the peak containment temperature
following a design basis accident exceeding the EQ profile by
2[emsp14][deg]F with respect to the current licensing basis is
negligible.
Therefore, operation of the facility in accordance with the
proposed change to increase the allowable containment average air
temperature from 120[emsp14][deg]F to 125[emsp14][deg]F 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: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Group, LLC, 750 East Pratt Street, 17 Floor,
Baltimore, MD 21202.
NRC Branch Chief: Sean Meighan, Acting.
Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Houston
County, Alabama
Date of amendment request: January 23, 2013.
Description of amendment request: The proposed change would revise
Technical Specification (TS) Section 5.5.9, ``Steam Generator (SG)
Program,'' 5.6.10, ``Steam Generator Tube Inspection Report,'' and the
Steam Generator Tube Integrity specification (LCO 3.4.17). The proposed
changes are needed to address implementation issues associated with the
inspection periods, and address other administrative changes and
clarifications.
The proposed amendment is consistent with TSTF-510, Revision 2,
``Revision to Steam Generator Program Inspection Frequencies and Tube
Sample Selection.''
In addition, this proposed amendment corrects the indenting for FNP
TS Section 5.5.9.a at the top of page 5.5-6. This change is purely
administrative, and has no technical impact on the TS.
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 license amendment 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 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 a SGTR is not
increased. The consequences of a SGTR are bounded by the
conservative assumptions in the design 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 previously evaluated?
Response: No.
The proposed changes to the SG 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
[[Page 28254]]
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.
Based on the above, SNC concludes that the proposed change
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Branch Chief: Robert J. Pascarelli.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: January 23, 2013.
Description of amendment request: The proposed change would revise
Technical Specification Section 5.5.9, ``Steam Generator (SG)
Program,'' 5.6.10, ``Steam Generator Tube Inspection Report,'' and the
Steam Generator Tube Integrity specification (LCO 3.4.17). The proposed
changes are needed to address implementation issues associated with the
inspection periods, and address other administrative changes and
clarifications.
The proposed amendment is consistent with TSTF-510, Revision 2,
``Revision to Steam Generator Program Inspection Frequencies and Tube
Sample Selection.''
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 license amendment 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 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 a SGTR is not
increased. The consequences of a SGTR are bounded by the
conservative assumptions in the design 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 previously evaluated?
Response: No.
The proposed changes to the SG 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.
Based on the above, SNC concludes that the proposed change
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street NE., Atlanta,
Georgia 30308-2216.
NRC Branch Chief: Robert Pascarelli.
Southern Nuclear Operating Company, Inc., Docket Nos.: 52-025 and 52-
026, Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke
County, Georgia
Date of amendment request: March 25, 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-120, ``AP1000 Human Factors
Design Engineering Verification Plan,'' from Revision B to Revision 0.
APP-OCS-GEH-120 is incorporated by reference in the updated 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?
[[Page 28255]]
Response: No.
Design verification provides a final check of the adequacy of
the Human System Interface (HSI) Resources and Operation and Control
Centers System (OCS) design. The changes do not affect the plant
itself, and so there is no change to the probability or consequences
of an accident previously evaluated. Changing the design
verification 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 as the purpose of the plan is simply to
verify implementation of design criteria. The Probabilistic Risk
Assessment is not affected. No safety-related structure, system,
component (SSC) or function is adversely affected. The 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 UFSAR are not affected. Because the
changes 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.
Design verification provides a final check of the adequacy of
the HSI Resources and Operation and Control Centers System design.
The changes do not affect the plant itself, and so there is no new
or different kind of accident from any accident previously
evaluated. Therefore, the changes do not affect safety-related
equipment, nor does it 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 is 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 from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The changes to the design verification plan provide a final
check of the adequacy of the HSI Resources and Operation and Control
Centers System design. The changes do not affect the assessments or
the plant itself. The changes do not affect safety-related equipment
or equipment whose failure could initiate an accident, nor does it
adversely interface with safety-related equipment or fission product
barriers. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the requested change.
Therefore, there is no 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.
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 25, 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-220, ``AP1000 Human Factors
Engineering Task Support Verification Plan,'' from Revision B to
Revision 0. APP-OCS-GEH-220 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 HFE Task Support Verification Plan is one of several
verification and validation (V&V) activities performed on human-
system interface (HSI) resources and the Operation and Control
Centers System (OCS), where applicable. The Task Support
Verification Plan is used to assess and verify displays and
activities related to normal and emergency operation. The changes
are to the Task Support Verification Plan to clarify the scope and
amend the details of the methodology. The Task Support Verification
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
Probabilistic Risk Assessment is not affected. No safety-related
structure, system, component (SSC) or function is adversely
affected. The 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 UFSAR are not
affected. Because the changes 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.
The changes to the Task Support Verification Plan change
information related to validation and verification on Human System
Interface and Operational Control Centers. 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 from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The changes to the Task Support Verification Plan affect the
validation and verification on the Human System Interface and the
Operational Control Centers. Therefore, the changes do not affect
the plant itself. These changes do not affect the design or
operation of safety-related equipment or equipment whose failure
could initiate an accident, nor does it adversely interface with
safety-related equipment or fission product barriers. No safety
analysis or design basis acceptance limit/criterion is challenged or
exceeded by the requested change.
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.
[[Page 28256]]
Attorney for licensee: Mr. M. Stanford Blanton, Blach & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
Acting 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: April 5, 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-420, ``AP1000 Human Factors
Engineering Discrepancy Resolution Process,'' from Revision B to
Revision 0. APP-OCS-GEH-420 is incorporated by reference in the 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 HFE Discrepancy Resolution Process is used to capture and
resolve Human Engineering Discrepancies (HEDs) identified during the
Human Factors Engineering (HFE) verification and validation (V&V)
activities. These discrepancy resolution process activities are used
to support the final check of the adequacy of the HFE design of the
Human-System Interface (HSI) resources and the Operation and Control
Centers Systems (OCS) design. The discrepancy resolution process
activities are performed as part of the V&V activities against the
final configuration and control documentation, simulator or
installed target system. The changes are to the Discrepancy
Resolution Process to clarify the scope and amend the details of the
methodology. The Discrepancy Resolution Process does not affect the
plant itself. Changing the Discrepancy Resolution Process 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. No safety-
related structure, system, component (SSC) or function is adversely
affected. The document revision 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 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.
The changes to the Discrepancy Resolution Process information
are related to discrepancy resolution of HEDs during the HFE V&V
activities on the HSI and the 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.
The changes to the Discrepancy Resolution Process affect
discrepancy resolution of HEDs during the HFE V&V activities on the
HSI and the OCS. Therefore, the changes do not affect the
assessments or the plant itself. These changes do not affect the
design or operation of safety-related equipment or equipment whose
failure could initiate an accident, nor does it adversely interface
with safety-related equipment or fission product barriers. No safety
analysis or design basis acceptance limit/criterion is challenged or
exceeded by the requested change.
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.
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].
[[Page 28257]]
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: April 17, 2012.
Brief description of amendment: The amendment revised the VYNPS
Technical Specification (TS) 3.5.A.5 and TS 4.5.A.5 to change the
normal position of the recirculation pump discharge bypass valves from
``open'' to ``closed''; and therefore, the valves' safety function to
close in support of accident mitigation is eliminated. The amendment
also revised the TSs to require the valves to remain closed and their
position to be verified once per operating cycle.
Date of Issuance: April 26, 2013.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 257.
Facility Operating License No. DPR-28: The amendment revised the
License and TSs.
Date of initial notice in Federal Register: October 2, 2012 (77 FR
60150).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated April 26, 2013.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: February 22, 2012, and
supplemented by letter dated.
March 8, 2013.
Brief description of amendment: FirstEnergy Nuclear Operating
Company, the licensee for the Perry Nuclear Power Plant Unit 1 (PNPP),
requested a license amendment to revise PNPP's Technical Specifications
(TS) 3.10.1, and the associated TS Bases, to expand its scope to
include provisions for temperature excursions greater than 200 degrees
Fahrenheit ([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 leak or hydrostatic test, while
considering operational conditions to be in MODE 4. This change is
consistent with the U.S. Nuclear Regulatory Commission (NRC)-approved
Revision 0 to Technical Specification Task Force (TSTF) Improved
Standard TS Change Traveler, TSTF-484, ``Use of TS 3.10.1 for Scram
Time Testing Activities.''
Date of issuance: April 18, 2013.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 163.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: July 24, 2012 (77 FR
43377). The March 8, 2013 supplement contained clarifying information
and did not change the NRC staff's initial proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 18, 2013.
No significant hazards consideration comments received: No.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, Rockingham County, New Hampshire
Date of amendment request: May 14, 2010, as supplemented by letters
dated August 24, 2010, September 16, 2011, March 15, 2012, July 2, 2012
and January 31, 2013.
Description of amendment request: The changes revise the Seabrook
Station Technical Specifications (TSs) governing the Containment
Enclosure Emergency Air Cleanup System (CEEACS). The amendment changes
TS Surveillance Requirement (SR) 4.6.5.1.d.4 so that it will
demonstrate integrity of the containment enclosure building rather than
operability of CEEACS. The amendment relocates SR 4.6.5.1.d.4 with
modifications to new SR 4.6.5.2.b. Additionally, the amendment makes
some minor wording changes, deletes a definition, and removes a moot
footnote.
Date of issuance: April 23, 2013.
Effective date: As of its date of issuance and shall be implemented
within 30 days.
Amendment No.: 136.
Facility Operating License No. NPF-86: The amendment revised the
Technical Specifications and the License.
Date of initial notice in Federal Register: July 13, 2010 (75 FR
39979). The notice was reissued in its entirety to include a revised
description of the amendment request on April 17, 2012 (77 FR 22815).
The notice was reissued again in its entirety to include a revised
description of the amendment request on July 24, 2012 (77 FR 43378).
The supplement dated January 31, 2013, provided additional information
that clarified the application, did not expand the scope of the
application as noticed, and did not change the NRC staff's proposed no
significant hazards consideration determination as published in the
Federal Register on July 24, 2012.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 23, 2013.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of application for amendments: August 31, 2012, as
supplemented on December 6, 2012.
Brief description of amendments: The amendments revise Technical
Specifications (TSs) 3.6.6, 3.7.5, 3.8.1, 3.8.9, and TS Example 1.3-3
by eliminating second completion times from the TSs in accordance with
TS Task Force Traveler (TSTF)-439, ``Eliminate Second Completion Times
Limiting Time from Discovery of Failure to Meet an LCO [Limiting
Condition for Operation].'' In addition, the amendment makes an
administrative change to TS 3.6.6 by removing an obsolete note
associated with Condition A.
Date of issuance: April 24, 2013.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 169 and 151.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the licenses and the TSs.
Date of initial notice in Federal Register: December 11, 2012 (77
FR 73690). The supplemental letter dated December 6, 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 amendments is contained
in a Safety Evaluation dated April 24, 2013.
No significant hazards consideration comments received: No.
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: August 1, 2012, as supplemented by
letter dated April 15, 2013.
Brief description of amendment: The amendments revised Technical
Specification (TS) Table 3.3-10, ``Accident Monitoring
Instrumentation,'' with respect to the required actions and
[[Page 28258]]
allowed outage times for inoperable instrumentation for Neutron Flux
(Extended Range) and Neutron Flux--Startup Rate (Extended Range)
(Instrument Nos. 19 and 23). The required actions have been revised to
enhance plant reliability by reducing exposure to unnecessary shutdowns
and increase operational flexibility by allowing more time to implement
required repairs for inoperable instrumentation. The changes are
consistent with requirements generically approved as part of NUREG-
1431, Standard Technical Specifications, Westinghouse Plants, Revision
4 (TS 3.3.3, ``Post Accident Monitoring (PAM) Instrumentation'').
Date of issuance: April 25, 2013.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1--200; Unit 2--198.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: October 2, 2012 (77 FR
60154). The supplemental letter dated April 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 as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 25, 2013.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 6th day of May 2013.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2013-11272 Filed 5-13-13; 8:45 am]
BILLING CODE 7590-01-P