[Federal Register Volume 77, Number 200 (Tuesday, October 16, 2012)]
[Notices]
[Pages 63343-63355]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2012-25240]
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NUCLEAR REGULATORY COMMISSION
[NRC-2012-0236]
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 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 September 20, 2012 to October 3, 2012. The
last biweekly notice was published on October 2, 2012 (77 FR 60146-
60160).
Addresses: You may access information and comment submissions
related to this document, which the NRC possesses and are publicly
available, by searching on http://www.regulations.gov under Docket ID
NRC-2012-0236. You may submit comments by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0236. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected].
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
Fax comments to: RADB at 301-492-3446.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
Supplementary Information section of this document.
Supplementary Information:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2012-0236 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 are publicly available, by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0236.
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,
[[Page 63344]]
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-2012-0236 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 will post all comment submissions at http://www.regulations.gov as well as enter 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 ``Rules of
Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.
Interested person(s) should consult a current copy of 10 CFR 2.309,
which is available at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. The NRC regulations are accessible electronically from the NRC
Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or a presiding
officer designated by the Commission or by the Chief Administrative
Judge of the Atomic Safety and Licensing Board Panel, will rule on the
request and/or petition; and the Secretary or the Chief Administrative
Judge of the Atomic Safety and Licensing Board will issue a notice of a
hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
[[Page 63345]]
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 the 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 contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC 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. 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
[[Page 63346]]
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].
Carolina Power and Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina.
Date of amendment request: June 19, 2012.
Description of amendment request: The proposed license amendments
would revise the Technical Specifications (TS) for the Brunswick Steam
Electric Plant (BSEP), Units 1 and 2. The TS change proposes to extend
the Completion Time (CT) of TS 3.8.1 Required Action D.4 for an
inoperable diesel generator (DG). A commensurate change is also
proposed to extend the maximum CT of TS 3.8.1 Required Actions C.3 and
D.4. The licensee stated that it will add a supplemental alternating
current power source (i.e., a supplemental diesel generator) with the
capability to power any E-bus within one hour from the Station Blackout
(SBO) event, and with the capacity to bring the affected unit to cold
shutdown, to support this request.
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 DGs are safety related components which provide backup
electrical power supply to the onsite emergency power distribution
system. The proposed changes do not affect the design of the DGs,
the operational characteristics or function of the DGs, the
interfaces between the DGs and other plant systems, or the
reliability of the DGs. The DGs are not accident initiators; the DGs
are designed to mitigate the consequences of previously evaluated
accidents including a loss of offsite power. Extending the CT for a
single DG would not affect the previously evaluated accidents since
the remaining DGs supporting the redundant ESF [engineered safety
feature] systems would continue to be available to perform the
accident mitigation functions. Thus, allowing a DG to be inoperable
for an additional 7 days for performance of maintenance or testing
does not increase the probability of a previously evaluated
accident.
Deterministic and probabilistic risk assessments evaluated the
effect of the proposed TS changes on the availability of an
electrical power supply to the plant emergency safeguards features
systems. These assessments concluded that the proposed TS changes do
not involve a significant increase in the risk of power supply
unavailability.
There is small incremental risk associated with continued
operation for an additional 7 days with one DG inoperable; however,
the calculated impact on risk provides risk metrics consistent with
the acceptance guidelines contained in RG [Regulatory Guide] 1.177
and 1.174 (References 7.2.1 and 7.2.2). This risk is judged to be
reasonably consistent with the risk associated with operations for 7
days with one DG inoperable as allowed by the current TS.
Specifically, the remaining operable DGs and paths are adequate
to supply electrical power to the onsite emergency power
distribution system. A DG is required to operate only if both
offsite power sources fail and there is an event which requires
operation of the plant engineered safety features such as a design
basis accident. The probability of a design basis accident occurring
during this period is low.
The consequences of previously evaluated accidents will remain
the same during the proposed 14-day CT as during the current 7-day
CT. The ability of the remaining TS required DG to mitigate the
consequences of an accident will not be affected since no additional
failures are postulated while equipment is inoperable within the TS
CT. The standby AC [alternating current] power supply for each of
the four safety-related load groups consists of one DG complete with
its auxiliaries, which include the cooling water, starting air,
lubrication, intake and exhaust, and fuel oil systems. The sizing of
the DGs and the loads assigned among them is such that any
combination of three out of four of these DGs is capable of shutting
down the plant safely, maintaining the plant in a safe shutdown
condition, and mitigating the consequences of accident conditions.
Thus this change does not involve a significant increase in the
probability or consequences of a previously analyzed accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve a change in the plant
design, plant configuration, system operation, or procedures
involved with the DGs. The proposed changes allow a DG to be
inoperable for additional time. Equipment will be operated in the
same configuration and manner that is currently allowed and designed
for. The functional demands on credited equipment is unchanged.
There are no new failure modes or mechanisms created due to plant
operation for an extended period to perform DG maintenance or
testing. Extended operation with an inoperable DG does not involve
any modification in the operational limits or physical design of
plant systems. There are no new accident precursors generated due to
the extended CT.
Therefore, the proposed changes do 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.
Currently, if an inoperable DG is not restored to operable
status within 7 days, TS 3.8.1, Condition H, requires the unit to be
in MODE 3 (i.e., HOT SHUTDOWN) within a CT of 12 hours, and to be in
MODE 4 (i.e., COLD SHUTDOWN) within a CT of 36 hours. This TS
Condition is entered on both units resulting in a dual-unit
shutdown. The proposed Technical Specification changes will allow
steady state plant operation at 100 percent power for an additional
7 days for performance of DG planned reliability improvements and
preventive and corrective maintenance.
Deterministic and probabilistic risk assessments evaluated the
effect of the proposed TS changes on the availability of an
electrical power supply to the plant ESF systems. These assessments
concluded that the proposed TS changes do not involve a significant
increase in the risk of power supply unavailability.
The DGs continue to meet their design requirements; there is no
reduction in capability or change in design configuration. The DG
response to LOOP [loss of offsite power], LOCA [loss-of-coolant
accident], SBO [station blackout], or fire is not changed by this
proposed amendment; there is no change to the DG operating
parameters. In the extended CT, as in the existing CT, the remaining
operable DGs and paths are adequate to supply electrical power to
the onsite emergency power distribution system. The proposed change
does not alter a design basis or safety limit; therefore, it does
not significantly reduce the margin of safety. The DGs will continue
to operate per the existing design and regulatory requirements.
[[Page 63347]]
The proposed TS changes do not alter the plant design nor does
it change the assumptions contained in the safety analyses. The
standby AC power system is designed with sufficient redundancy such
that a DG may be removed from service for maintenance or testing.
The remaining DGs are capable of carrying sufficient electrical
loads to satisfy the UFSAR [updated final safety analysis report]
requirements for accident mitigation or unit safe shutdown. The
proposed changes do not impact the redundancy or availability
requirements of offsite power circuits or change the ability of the
plant to cope with a SBO.
Therefore, based on the considerations given above, 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: David T. Conley, Senior Counsel--Manager
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, NC 27602.
NRC Acting Branch Chief: Jessie Quichocho.
Carolina Power and Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina; Florida Power Corporation, et al., Docket No. 50-302, Crystal
River Unit 3 Nuclear Generating Plant, Citrus County, Florida
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina; Carolina Power & Light Company, Docket No. 50-261, H. B.
Robinson Steam Electric Plant, Unit 2, Darlington County, South
Carolina
Date of amendment request: September 12, 2012.
Description of amendment request: The proposed license amendments
would revise the Facility Operating Licenses for the Brunswick Steam
Electric Plant, Units 1 and 2, H. B. Robinson Steam Electric Plant,
Unit 2, Shearon Harris Nuclear Power Plant, Unit 1, and Crystal River
Unit No. 3 Nuclear Generating Plant. The NRC issued license amendments,
dated July 29, 2011, that approved the licensees' cyber security plan
and associated implementation milestone schedule. Milestone 6 requires
the identification, documentation, and implementation of cyber security
controls for critical digital assets that could adversely impact the
design function of physical security target set equipment by no later
than December 31, 2012. The license amendment request would change the
existing facility operating licenses for the Physical Protection/
Security license condition for these plants to reference the change to
an implementation schedule milestone and a proposed Revised Cyber
Security Plan Implementation Schedule for the scope of Milestone 6.
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 the Cyber Security Plan Implementation
Schedule is administrative in nature. This change does not alter
accident analysis assumptions, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected. The proposed
change does not require any plant modifications which affect the
performance capability of the structures, systems, and components
relied upon to mitigate the consequences of postulated accidents and
has no impact on the probability or consequences of an accident
previously evaluated.
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 to the Cyber Security Plan Implementation
Schedule is administrative in nature. This proposed change does not
alter accident analysis assumptions, add any initiators, or affect
the function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected. The proposed
change does not require any plant modifications which affect the
performance capability of the structures, systems, and components
relied upon to mitigate the consequences of postulated accidents and
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Plant safety margins are established through limiting conditions
for operation, limiting safety system settings, and safety limits
specified in the technical specifications. The proposed change to
the Cyber Security Plan Implementation Schedule is administrative in
nature. Because there is no change to these established safety
margins as result of this change, the proposed change does not
involve a significant reduction in a margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Deputy General Counsel,
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A,
Charlotte, NC 28202.
NRC Acting Branch Chief: Jessie Quichocho.
Carolina Power and Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit 2, (HBRSEP) Darlington County, South
Carolina
Date of amendment request: August 6, 2012.
Description of amendment request: The proposed change would revise
the Technical Specification (TS) requirements for inoperable snubbers
by adding Limiting Condition for Operation (LCO) 3.0.8. The change is
consistent with NRC approved Revision 4 to Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
372, ``Addition of LCO 3.0.8, Inoperability of Snubbers.''
The NRC staff issued a notice of availability of a model safety
evaluation and model no significant hazards consideration (NSHC)
determination for referencing in license amendment applications in the
Federal Register on May 4, 2005 (70 FR 23252). The licensee affirmed
the applicability of the model NSHC determination in its application
dated August 6, 2012.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
[[Page 63348]]
snubber if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. Therefore, the probability
of an accident previously evaluated is not significantly increased,
if at all. The consequences of an accident while relying on
allowance provided by proposed LCO 3.0.8 are no different than the
consequences of an accident while relying on the TS required actions
in effect without the allowance provided by proposed LCO 3.0.8.
Therefore, the consequences of an accident previously evaluated are
not significantly affected by this change. The addition of a
requirement to assess and manage the risk introduced by this change
will further minimize possible concerns.
Therefore, 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 change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns.
Thus, 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
the margin of safety?
Response: No.
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG [Regulatory Guide] 1.177. A bounding risk
assessment was performed to justify the proposed TS changes. This
application of LCO 3.0.8 is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The net change to the margin of safety is insignificant.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Manager--Senior Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Acting Branch Chief: Jessie F. Quichocho.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit 2, (HBRSEP) Darlington County, South Carolina
Date of amendment request: August 29, 2012.
Description of amendment request: The proposed change combines two
changes that affect the same Technical Specification (TS) sections into
one license amendment. The first part proposes to implement revisions
consistent with Technical Specification Task Force (TSTF)-510, Revision
2, ``Revision to Steam Generator (SG) Program Inspection Frequencies
and Tube Sample Selection.'' The second part proposes to permanently
revise TS 5.5.9 ``Steam Generator Program'' to exclude portions of the
SG tube below the top of the SG tubesheet from periodic inspections by
implementing the permanent alternate repair criteria ``H.*'' References
2, 3, 8, 23 and 32 referred to in the licensees analysis can be found
in the license amendment request under ADAMS Accession No. ML12275A176.
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 previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The proposed change
modifies steam generator tube inspection frequencies and tube
selection consistent with TSTF-510 and excludes the lower portion of
steam generator tubes from inspection by implementing the alternate
repair criteria (H*) on a permanent basis and does not have a
detrimental impact on the integrity of any plant structure, system,
or component that initiates an analyzed event. The proposed change
will not alter the operation of, or otherwise increase the failure
probability of any plant equipment that initiates an analyzed
accident.
Of the applicable accidents previously evaluated, the limiting
transients with consideration to the proposed change to the SG tube
inspection and repair criteria are the SG tube rupture (SGTR) event
and the main steam line break (MSLB) postulated accident.
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 SG tube
inspection frequency and sample selection criteria will not cause
the consequences of a SGTR to exceed those assumptions.
With respect to the SGTR event, the required structural
integrity margins of the SG tubes and the tube-to-tubesheet joint
over the H* distance will be maintained. Tube rupture in tubes with
cracks within the tubesheet is precluded by the constraint provided
by the presence of the tubesheet and the tube-to-tubesheet joint.
Tube burst cannot occur within the thickness of the tubesheet. The
tube-to-tubesheet joint constraint results from the hydraulic
expansion process, thermal expansion mismatch between the tube and
tube sheet, and from the differential pressure between the primary
and secondary side, and tube sheet rotation. The structural margins
against burst, as discussed in Regulatory Guide [RG] 1.121, ``Bases
for Plugging Degraded PWR [Pressurized-Water Reactor] Steam
Generator Tubes'' (Reference 32) and [Nuclear Energy Institute] NEI
97-06, ``Steam Generator Program Guidelines,'' (Reference 8) are
maintained for both normal and postulated accident conditions.
For the portion of the tube outside of the tubesheet, the
proposed change also has no impact on the structural or leakage
integrity. Therefore, the proposed change does not result in a
significant increase in the probability of the occurrence of a SGTR
accident.
At normal operating pressures, leakage from degradations below
the proposed limited inspection depth is limited by the tube-to-
tubesheet crevice. Consequently, negligible normal operating leakage
is expected from degradation below the inspected depth within the
tubesheet region. The consequences of an SGTR event are affected by
the primary to secondary leakage flow during the event. However,
primary to secondary leakage flow through a postulated tube that has
been pulled out of the tubesheet is not affected by the proposed
changes since the tubesheet enhances the tube integrity in the
region of the hydraulic expansion by precluding tube deformation
beyond its initial hydraulically expanded outside diameter.
Therefore, the proposed change does not result in a significant
increase in the consequences of an SGTR. In addition, the selected
H* value envelopes the depth within the tubesheet required to
prevent a tube pullout.
The probability of a MSLB event is unaffected by the potential
failure of a SG tube as the failure of a tube is not an initiator
for a MSLB event. Therefore the proposed SG tube inspection
frequency and sample selection criteria and the structural integrity
margins of the SG tubes and the tube-to-tubesheet joint over the H*
distance do not increase the probability of a MSLB event.
[[Page 63349]]
The leak rate factor of 1.82 for HBRSEP, for a postulated MSLB,
has been calculated as shown in References 2, 3 and 23. HBRSEP Unit
No. 2 will apply the factor of 1.82 to the normal operating leakage
associated with the tubesheet expansion region in the condition
monitoring and operational assessment. Through application of the
limited tube sheet inspection scope, the existing operating leakage
limit provides assurance that excessive leakage (i.e., greater than
accident analysis assumptions) will not occur.
When the TS operational leak rate limit of 75 [gallons per day]
gpd or about 0.052 gallons per minute (gpm) through any one SG is
multiplied by the MSLB leak rate factor applicable to HBRSEP Unit
No. 2 of 1.82 (Table 9-7 in [Westinghouse Commercial Atomic Power
Report] WCAP-17091-P, Reference 3) the maximum primary to secondary
accident induced leak rate is less than 0.095 gpm and is bounded by
the value of 0.11 gpm through the faulted SG used in the MSLB
accident analyses. Since the existing limit on operational leakage
continues to ensure that the MSLB assumed accident induced leakage
will not be exceeded, the consequences of a MSLB accident are not
increased.
For the condition monitoring assessment, the component of
leakage from the prior cycle from below the H* distance will be
multiplied by a factor of 1.82 and added to the total leakage from
any other source and compared to the allowable accident induced leak
rate. For the operational assessment, the difference in the leakage
between the allowable leakage and the calculated accident induced
leakage from sources other than the tubesheet expansion region will
be divided by 1.82 and compared to the observed operational leakage.
Based on the above, 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 modifies steam generator tube inspection
frequencies and tube selection consistent with TSTF-510 and excludes
the lower portion of steam generator tubes from inspection by
implementing the alternate repair criteria (H*) on a permanent
basis. The proposed change does not introduce any new equipment,
create new failure modes for existing equipment, or create any new
limiting single failures resulting from 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. Plant operation will not
be altered, and all safety functions will continue to perform as
previously assumed in accident analyses.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the change involve a significant reduction in the 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 but will limit inspection within the tubesheet to the
portion of the tube from the top of the tubesheet to a distance H*
below the top of the tubesheet.
The proposed change modifies steam generator tube inspection
frequencies and tube selection consistent with TSTF-510 and limits
required inspection to the safety significant portion of the steam
generator tubes. WCAP-17345, Rev. 2 (Reference 2) identifies the
specific inspection depth (H*) below which any type of tube
degradation is shown to have no impact on the performance criteria
in NEI 97-06 Rev. 3, ``Steam Generator Program Guidelines''
(Reference 8) and TS 5.5.9, ``Steam Generator (SG) Program.''
Changes associated with inspection frequency and tube selection
criteria are consistent with TSTF-510 and are based on recent
industry experience and are more effective in managing the frequency
of verification of tube integrity and sample selection than those
required by current TSs.
The proposed change maintains the required structural margins of
the SG tubes for both normal and accident conditions. Nuclear Energy
Institute 97-06, ``Steam Generator Program Guidelines'' (Reference
8), and NRC Regulatory Guide 1.121 ``Bases for Plugging Degraded PWR
Steam Generator Tubes'' (Reference 32), are used as the bases in the
development of the limited tubesheet inspection depth methodology
for determining that SG tube integrity considerations are maintained
within acceptable limits. Regulatory Guide 1.121 describes a method
acceptable to the NRC for meeting General Design Criteria (GDC) 14,
``Reactor Coolant Pressure Boundary,'' GDC 15, ``Reactor Coolant
System Design,'' GDC 31, ``Fracture Prevention of Reactor Coolant
Pressure Boundary,'' and GDC 32, ``Inspection of Reactor Coolant
Pressure Boundary,'' by reducing the probability and consequences of
a SGTR. Regulatory Guide 1.121 concludes that by determining the
limiting safe conditions for tube wall degradation, the probability
and consequences of a SGTR are reduced. This Regulatory Guide uses
safety factors on loads for tube burst that are consistent with the
requirements of Section III of the American Society of Mechanical
Engineers (ASME) Code.
For axially oriented cracking located within the tubesheet, tube
burst is precluded due to the presence of the tubesheet. For
circumferentially oriented cracking, Westinghouse WCAP-17091-P, Rev.
0 (Reference 3) and WCAP-17345, Rev. 2 (Reference 2) define a length
of degradation-free expanded tubing that provides the necessary
resistance to tube pullout due to the pressure induced forces, with
applicable safety factors applied. Application of the limited hot
and cold leg tubesheet inspection criteria will preclude
unacceptable primary to secondary leakage during all plant
conditions. When the TS operational leak rate limit of 75 gpd or
about 0.052 gpm through any one SG is multiplied by the MSLB leak
rate factor applicable to HBRSEP Unit No. 2 of 1.82 (Table 9-7 in
WCAP-17091-P (Reference 3) the maximum primary to secondary accident
induced leak rate is less than 0.095 gpm and is bounded by the value
of 0.11 gpm through the faulted SG used in the MSLB accident
analyses.
Therefore, the proposed change does not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the license'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: David T. Conley, Manager--Senior Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Acting Branch Chief: Jessie F. Quichocho.
Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power
Station (KPS), Kewaunee County, Wisconsin
Date of amendment request: July 30, 2012.
Description of amendment request: The proposed amendment would
modify KPS Technical Specifications requirements regarding steam
generator tube inspections and reporting as described in Technical
Specification Task Force (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 change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
[[Page 63350]]
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
the 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: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc.,
120 Tredegar Street, Richmond, VA 23219.
NRC Acting Branch Chief: Istvan Frankl.
Entergy Nuclear Operations, Inc., Docket No 50-286, Indian Point
Nuclear Generating Unit 3, Westchester County, New York
Date of amendment request: August 14, 2012.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.5.4, ``Refueling Water Storage Tank,''
such that the non-seismically qualified piping of the spent fuel pool
purification system may be connected to the refueling water storage
tank (RWST) seismic piping by manual operation of a RWST seismically
qualified boundary valve under administrative controls for a limited
period of time (i.e., 14 days per fuel cycle for filtration for removal
of suspended solids from the RWST water). This change will only be
applicable until Refueling Outage 18 (Spring 2015) ends.
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 SFP [spent fuel pool] Purification Loop to re-
circulate the RWST 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 SFP Purification Loop to
the RWST would not affect the probability of these accidents
occurring. The SFP Purification Loop is not credited for safe
shutdown of the plant or accident mitigation. Administrative
controls ensure that the SFP Purification Loop can be isolated as
necessary 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 Purification Loop to allow
recirculation and purification has not been altered. Procedures for
the operation of the plant have not been revised to create the
possibility of a new or different type of accident. Contingent upon
manual operator action, a SFP Purification Loop line break will not
result in a loss of the RWST safety function. Similarly, an active
or passive failure in the SFP Purification Loop will not be
significantly different whether aligned to the SFP or the RWST.
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 is not credited for safe shutdown of
the plant or accident mitigation. Adequate 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 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.
NRC Branch Chief: George Wilson.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit 3, Westchester County, New York
Date of amendment request: August 14, 2012.
Description of amendment request: The amendment changes the
licensing basis regarding the emergency diesel generator (EDG) fuel oil
storage requirements. The current licensing basis, which requires
sufficient EDG fuel oil to operate two EDGs at minimum safeguards for
seven days, will be revised to provide sufficient EDG fuel oil to
operate three EDGs at modified rated capacity for seven days. The
Conditions, Required Actions, and Completion Times of Technical
Specification (TS) 3.8.3, ``Diesel Fuel Oil, and Starting Air,'' will
be modified
[[Page 63351]]
to be more consistent with the Nuclear Regulatory Commission's improved
Standard Technical Specifications. Finally, TS 3.8.3 will be modified
to relocate specific numerical values for EDG fuel oil storage
requirements from the TSs to the TS Bases in accordance with TS Task
Force (TSTF) 501 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change relocates the volume of diesel fuel oil
required to support 7 day operation of the onsite diesel generators
to licensee control, revises the action statement to reflect the
volume equivalent to a 6 day supply and locates the volume in the TS
Bases under licensee control, consolidates surveillance requirements
and recalculates the fuel oil volume required for the EDG. The
revised specific volume of fuel oil equivalent to a 7 and 6 day
supply is calculated consistent with the NRC approved methodology
described in Regulatory Guide 1.137, Revision 1, ``Fuel Oil Systems
for Standby Diesel Generators'' and ANSI N195-1976, ``Fuel Oil
Systems for Standby Diesel Generators.'' Because the requirement to
maintain a 7 day supply of diesel fuel oil is not changed and is
revised to be more consistent with the assumptions in the accident
analyses, the consolidated surveillances are more conservative, and
the actions taken when the volume of fuel oil is less than a 6 day
supply have not changed, neither the probability or the consequences
of any accident previously evaluated will be affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The change does not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. The change
does not alter assumptions made in the safety analysis but ensures
that the diesel generator operates as assumed in the accident
analysis. The proposed change is consistent with the safety analysis
assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change relocates the volume of diesel fuel oil
required to support 7 day operation of the onsite diesel generators,
revises the action statement to reflect the volume equivalent to a 6
day supply, locates the volume in the TS Bases under licensee
control, consolidates surveillance requirements and recalculates the
fuel oil volume required for the EDG. Although the bases for the
existing limits on diesel fuel oil are changed, no change is made to
the accident analysis assumptions and no margin of safety is reduced
as part of this change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: George Wilson.
South Carolina Electric and Gas Docket Nos.: 52-027 and 52-028, Virgil
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: August 29, 2012.
Description of amendment request: The proposed change would amend
Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear
Station (VCSNS) Units 2 and 3, respectively, by adding four non-Class
1E containment electrical penetration assemblies (EPAs). Containment
EPAs are a passive extension of containment which provide the passage
of the electric conductors through a single aperture in the nuclear
containment structure, while providing a pressure barrier between the
inside and the outside of the containment structure.
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 additional containment EPAs are a passive extension of
containment and provide a pathway for communication [sic, passage]
of non-Class 1E electrical signals [sic, conductors] between the
Auxiliary Building and Containment. The proposed containment
electrical penetration assemblies are similar in form, fit and
function to the current non-Class 1E containment electrical
penetration assemblies. The maximum allowable leakage rate allowed
by Technical Specifications is also unchanged. The new EPAs will
meet the same design function as current EPAs; therefore, the
additional penetrations 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 containment electrical penetration assemblies are
similar in form, fit, and function to the current non-Class 1E
containment electrical penetration assemblies. The new EPAs will
meet the same design function as current EPAs. Because the new EPAs
are virtually identical in design and function to the current EPAs,
no new type of failure modes exist.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed containment electrical penetration assemblies are
similar in form, fit and function to the current non-Class 1E
containment electrical penetration assemblies. The additional
containment electrical penetration assemblies are an engineered
passive extension of containment, and, therefore, do not affect
containment or its ability to perform its design function. The
addition of the new EPAs does not exceed or alter a design basis or
safety limit and, therefore, does not significantly reduce 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: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Mark E. Tonacci.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of amendment request: July 31, 2012.
Description of amendment request: The proposed amendment would
modify the Technical Specifications requirements regarding steam
generator tube inspections and reporting as described in TSTF-510,
Revision 2, ``Revision to Steam Generator Program Inspection
Frequencies and Tube Sample Selection.''
[[Page 63352]]
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. The proposed change to reporting requirements and
clarifications of the existing requirements have no affect on the
probability or consequences of SGTR.
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. Based on the
above, Dominion 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: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
NRC Branch Chief: Robert J. Pascarelli.
Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the NRC's Public Document Room (PDR), located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through the Agencywide Documents Access and
Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by email to
[email protected].
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit 3, New London County, Connecticut
Date of amendment request: November 17, 2011.
Description of amendment request: The proposed amendment would add
Optimized ZIRLO\TM\ as an allowable fuel rod cladding material and add
the Westinghouse topical report on Optimized ZIRLO\TM\ to the Millstone
Power Station, Unit 3 Technical Specifications. In addition, a
typographical error would be corrected.
Date of issuance: September 24, 2012.
Effective date: As of the date of issuance, and shall be
implemented within 60 days. Amendment No.: 253.
Renewed Facility Operating License No. NPF-69: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: May 15, 2012 (77 FR
28629).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 24, 2012.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit 3, New London County, Connecticut
Date of amendment request: July 23, 2012.
Description of amendment request: The proposed amendment would
conform the Millstone Power Station Unit 3 (MPS3) licenses to reflect a
name change for Central Vermont Public Service Corporation (CVPS)
resulting
[[Page 63353]]
from a subsequent restructuring in which CVPS will be consolidated with
Gaz M[eacute]tro's other electric utility subsidiary in Vermont, Green
Mountain Power Corporation.
Date of issuance: October 3, 2012.
Effective date: As of the date of issuance, and shall be
implemented within 30 days. Amendment No.: 254.
Renewed Facility Operating License No. NPF-49: Amendment revised
the License.
Date of initial notice in Federal Register: July 20, 2012 (77 FR
42768).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 21, 2012.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: September 26, 2011, as
supplemented April 16, 2012.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to incorporate aspects of NUREG-1431,
``Standard Technical Specifications--Westinghouse [Electric Company]
Plants,'' STS 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,''
Condition E, regarding Diesel Generator starting air receiver pressure
limits.
Date of issuance: September 28, 2012.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1--267 and Unit 2--247.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: May 29, 2012 (77 FR
31659).
The supplement dated April 16, 2012, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 28, 2012.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: December 5, 2011.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.3.1, Table 3.3.1-1, ``Reactor Trip System
Instrumentation,'' Function 16(e) to replace the phrase ``Turbine
Impulse Pressure'' with ``Turbine Inlet Pressure.''
Date of issuance: October 1, 2012.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1--268 and Unit 2--248.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: August 9, 2012 (77 FR
47677).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 1, 2012.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket No. 50-269, Oconee Nuclear Station,
Unit 1, Oconee County, South Carolina
Date of application of amendments: April 3, 2012
Brief description of amendments: The amendment revised the
Technical Specifications related to the integrated leak rate test of
the reactor containment building.
Date of Issuance: October 1, 2012.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 381.
Renewed Facility Operating License No. DPR-38: Amendment revised
the license and the technical specifications.
Date of initial notice in Federal Register: July 10, 2012, 77 FR
40651.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 1, 2012.
No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1 (RBS), West Feliciana
Parish, Louisiana
Date of amendment request: September 12, 2011, as supplemented by
letters dated October 13, 2011, March 22, 2012, and April 3, 2012.
Brief description of amendment: The amendment approved a change to
the site Emergency Plan to relocate the existing backup emergency
operations facility for RBS from its current location at the Entergy
Operations-Baton Rouge Division Office, located at 1509 Government
Street in Baton Rouge, Louisiana, approximately 23 miles southeast of
RBS, to the Entergy Customer Service Center, located at 5564 Essen Lane
in Baton Rouge, Louisiana, approximately 28 miles southeast of RBS.
Date of issuance: September 24, 2012.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 175.
Facility Operating License No. NPF-47: The amendment revised the
RBS Emergency Plan.
Date of initial notice in Federal Register: December 27, 2011 (76
FR 80975). The supplemental letters dated October 13, 2011, March 22,
2012, and April 3, 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 September 24, 2012.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: August 16, 2011, as supplemented
by letters dated March 30, June 13, August 1, August 16, and September
14, 2012.
Brief description of amendment: The amendment revises the James A.
FitzPatrick's (JAF's) current licensing basis, in the Updated Final
Safety Analysis Report, to support installation of new reserve station
service transformers (RSST) with on-load tap changers (OLTC). The new
RSSTs with OLTCs will compensate for the wider range of offsite power
voltage variations so that acceptable voltages at the safety-related
equipment will be better maintained. The new RSSTs provided with OLTCs
would facilitate operations in the automatic mode.
The OLTCs are sub-components of two new RSSTs that will be
installed at JAF during the refueling outage scheduled for September
2012.
Date of issuance: September 26, 2012.
Effective date: As of the date of issuance, and shall be
implemented
[[Page 63354]]
within 90 days The implementation of the amendment shall include
revision of the Updated Final Safety Analysis Report as described in
the licensee's application for this amendment.
Amendment No.: 302.
Renewed Facility Operating License No. DPR-59: The amendment
revised the License and the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: November 15, 2011 (76
FR 70768).
The supplements dated March 30, June 13, August 1, August 16, and
September 14, 2012 provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the NRC 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 September 26, 2012.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station (BVPS), Units 1 and 2, Beaver County,
Pennsylvania; Docket No. 50-346, Davis-Besse Nuclear Power Station
(DBNPS), Unit 1, Ottawa County, Ohio; Docket No. 50-440, Perry Nuclear
Power Plant (PNPP), Unit 1, Lake County, Ohio
Date of application for amendments: September 20, 2011, as
supplemented by letter dated June 21, 2012.
Brief description of amendments: The proposed amendments would
change the licenses of BVPS, Units 1 and 2, DBNPS, and PNPP to reflect
the name change of an owner licensee from ``FirstEnergy Nuclear
Generation Corp.'' to ``FirstEnergy Nuclear Generation, LLC.'' The
proposed amendment is administrative in nature. The proposed amendment
will also correct errors regarding the name of FirstEnergy Nuclear
Generation Corp in the DBNPS and PNNP Facility Operating Licenses.
Date of issuance: October 2, 2012.
Effective date: At date of issuance.
Amendment Nos.: No. 290 to Facility Operating License No. DPR-66
and No. 177 to Facility Operating License No. NPF-73 for BVPS, Units 1
and 2, respectively, and No. 286 to Facility Operating License No. NPF-
3 for DBNPS, and No. 161 to Facility Operating License No. NPF-58 for
PNPP.
Facility Operating License Nos. DPR-66, NPF-73, NPF-3, and NPF-58:
The amendments revised the Facility Operating Licenses.
Date of initial notice in Federal Register: October 18, 2011 (76 FR
64391).
The supplemental letter dated June 21, 2012, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed finding of no significant hazards
consideration determination published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 2, 2012.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-389, St. Lucie
Plant, Unit 2, St. Lucie County, Florida.
Date of application for amendment: February 25, 2011, as
supplemented by the letters dated November 4 and December 8, 2011, and
April 30 and May 4 and 7, 2012.
Brief description of amendment: This amendment raises the maximum
fuel enrichment for fresh fuel storage from a maximum of 4.5 weight
percent uranium-235 to a maximum lattice averaged value of 4.6 weight
percent uranium-235. The Technical Specification changes associated
with fuel stored in the spent fuel pool include increasing the maximum
initial enrichment from 4.5 weight percent uranium-235 to a maximum
planar average initial enrichment of 4.6 weight percent uranium-235,
credit for empty storage locations, credit for use of
METAMICTM inserts, credit for installation of full-length
full-strength, five-fingered control element assemblies, and definition
of three special configurations referred to in the nuclear criticality
safety analysis as inspection and maintenance configurations.
Date of issuance: September 18, 2012.
Effective date: This license amendment is effective as of its date
of issuance and shall be implemented within 60 days.
Amendment No: 162.
Renewed Facility Operating License No. NPF-16: Amendment revised
the Operating License and the Technical Specifications.
Date of initial notice in Federal Register: September 1, 2011 (76
FR 54503).
The supplemental letters provided additional information that
clarified the application and did not expand the scope of the
application as originally noticed and published in the Federal
Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 18, 2012.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-389, St. Lucie
Plant, Unit 2, St. Lucie County, Florida.
Date of application for amendment: February 25, 2011 as
supplemented by letters dated February 25, May 24, July 22, August 18
(three letters), August 25 (three letters), August 29, September 2,
September 8 (two letters), September 22, October 5, October 10, October
12 (two letters), October 31, November 2, November 3, November 4,
November 7, November 14 (three letters), November 23 (three letters),
December 8, December 14, December 20, December 27, December 29, 2011,
January 14, 2012 (two letters), January 18 (two letters), January 21
(two letters), February 29, March 6 (two letters), March 8, March 15,
March 16, March 17 (two letters), March 25, March 31 (two letters),
April 5 (two letters), April 6, April 10, April 19 (seven letters),
April 30, May 4, May 7, May 18, and July 23, 2012.
Brief description of amendments: The proposed amendments would
increase the licensed core power level for St. Lucie Unit 2 from 2070
megawatts thermal (MWt) to 3020 MWt. This represents a net increase in
the core thermal power of approximately 11.85 percent, including a 10-
percent power uprate and a 1.7 percent measurement uncertainty
recapture, over the current licensed thermal power level and is defined
as an extended power uprate. The proposed amendments would change the
renewed facility operating license and the technical specifications
(TSs) to support operation at the increased core thermal power level,
including changes to the maximum licensed reactor core thermal power,
reactor core safety limits, and reactor protection system and
engineered safety feature actuation system limiting safety system
settings. Additional TS changes include reactor coolant system heatup
and cooldown limitations, accumulator and refueling water storage tank
boron concentrations, main steam safety valve lift settings, emergency
diesel generator fuel storage and core operating limits report
references. A complete list of the proposed TS changes and the
licensee's basis for change can be found in Attachment 1 of the
licensee's application (ADAMS Accession No. ML110730116).
Date of issuance: September 24, 2012.
[[Page 63355]]
Effective date: This license amendment is effective as of its date
of issuance and shall be implemented within 60 days.
Amendment No.: 163.
Renewed Facility Operating License No. NPF-16: Amendment revised
the Operating License and the Technical Specifications.
Date of initial notice in Federal Register: September 1, 2011 (76
FR 54503).
The supplemental letters provided additional information that
clarified the application and did not expand the scope of the
application as originally noticed and published in the Federal
Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 24, 2012.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: September 16, 2011, as supplemented by
letters dated May 2, May 24, and September 17, 2012.
Brief description of amendment: The amendment revised the Cooper
Nuclear Station Technical Specifications (TS) and Operating License to
implement a 24-month fuel cycle and adopt TS Task Force (TSTF) Traveler
TSTF-493, Revision 4, ``Clarify Application of Setpoint Methodology for
LSSS [Limiting Safety System Settings] Functions,'' Option A.
Specifically, the amendment revised certain TS Surveillance Requirement
frequencies that are specified as ``18 months'' by changing them to
``24 months'' in accordance with the guidance provided in NRC Generic
Letter 91-04, ``Changes in Technical Specification Surveillance
Intervals to Accommodate a 24-Month Fuel Cycle.''
Date of issuance: September 28, 2012.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 242.
Renewed Facility Operating License No. DPR-46: Amendment revised
the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 6, 2012 (77 FR
13371). The supplemental letters dated May 2, May 24, and September 17,
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 September 28, 2012.
No significant hazards consideration comments received: No.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center (DAEC), Linn County, Iowa
Date of application for amendments: May 1, 2012, as supplemented by
letters dated June 27, 2012, and July 26, 2012.
Brief description of amendments: The amendment revises existing TS
3.3.5.1, on a one-time basis only, by adding a note to TS Table
3.3.5.1-1, Function 1d, Modes 4 and 5. This one-time license amendment
enables DAEC to re-coat the internal surface of the Suppression Chamber
during Refueling Outage 23.
Date of issuance: September 27, 2012.
Effective date: This license amendment is effective as of the date
of issuance and shall be implemented within 30 days from date of
issuance.
Amendment No.: 283.
Renewed Facility Operating License No. DPR-49: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 10, 2012 (77 FR
40654).
The licensee's June 27, 2012, and July 26, 2012, supplemental
letters contained clarifying information, did not change the scope of
the original amendment request, did not change the NRC staff's initial
proposed finding of no significant hazards consideration determination,
and did not expand the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 27, 2012.
No significant hazards consideration comments received: Yes.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendment: June 17, 2011, supplemented by
letters dated July 27, 2011, November 14, 2011, March 23, April 26, May
15, May 24, and June 26, 2012 (TS-SQN-2011-07).
Brief description of amendment: The amendments revised the
licensing basis and the Technical Specifications (TSs) to permit the
use of a more robust AREVA Advanced W17 high thermal performance fuel.
This new fuel has been selected to address fuel assembly distortion and
its resultant fuel-handling issues. The AREVA Advanced W17 HTP fuel
assembly design consists of standard uranium dioxide fuel pellets with
gadolinium oxide burnable poison and M5TM cladding. The new
fuel design ensures mechanical compatibility with the existing fuel,
reactor core, control rods, steam supply system, and fuel-handling
system. The transition from the existing fuel (AREVA Mark-BW) to new
fuel is planned to occur over two refueling cycles for each unit.
Date of issuance: September 26, 2012.
Effective date: As of the date of issuance and shall be implemented
prior to startup from fall 2013 refueling outage (RFO) for Unit 1, and
prior to startup from fall 2012 RFO for Unit 2.
Amendment Nos.: 331 and 324.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the License and TSs.
Date of initial notice in Federal Register: August 23, 2011 (76 FR
52703). The supplement letters dated July 27, 2011, November 14, 2011,
March 23, April 26, May 15, May 24, and June 26, 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.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 26, 2012.
No significant hazards consideration comments received: No.
For The Nuclear Regulatory Commission.
Dated at Rockville, Maryland, this 5th day of October 2012.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2012-25240 Filed 10-15-12; 8:45 am]
BILLING CODE 7590-01-P