Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 47123-47131 [2012-19004]
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Federal Register / Vol. 77, No. 152 / Tuesday, August 7, 2012 / Notices
amendments to operating licenses, the
NRC staff determined that the subject
exemption sought involves employment
suitability requirements. The NRC has
determined that this exemption involves
no significant hazards considerations:
(1) The proposed exemption is
administrative in nature and is limited
to allowing a temporary exception from
meeting the requirements of 10 CFR
26.205(c) and (d) during severe weather
to ensure that work hour controls do not
impede the ability to use available staff
resources to respond to a severe weather
event. The proposed exemption does
not make any physical changes to the
facility and does not alter the design,
function or operation of any plant
equipment. Therefore, issuance of this
exemption does not significantly
increase the probability or consequences
of an accident previously evaluated.
(2) The proposed exemption does not
make any changes to the facility and
would not create any new accident
initiators. Therefore, this exemption
does not create the possibility of a new
or different kind of accident from any
accident previously evaluated.
(3) The proposed exemption does not
alter the design, function or operation of
any plant equipment. Therefore, this
exemption does not involve a significant
reduction in the margin of safety.
Based on the above, the NRC has
concluded that the proposed exemption
does not involve a significant hazards
consideration under the standards set
forth in 10 CFR 50.92, and accordingly,
a finding of ‘‘no significant hazards
consideration’’ is justified.
The NRC staff has also determined
that the exemption involves no
significant increase in the amounts, and
no significant change in the types, of
any effluents that may be released
offsite; that there is no significant
increase in individual or cumulative
occupational radiation exposure; that
there is no significant construction
impact; and there is no significant
increase in the potential for or
consequences from a radiological
accident. Furthermore, the requirement
from which the licensee will be
exempted involves scheduling
requirements. Accordingly, the
exemption meets the eligibility criteria
for categorical exclusion set forth in 10
CFR 51.22(c)(25). Pursuant to 10 CFR
51.22(b), no environmental impact
statement or environmental assessment
is required to be prepared in connection
with the issuance of the exemption.
5.0 Conclusion
The Commission has determined that
granting these exemptions is consistent
with 10 CFR 26.207(d), ‘‘Plant
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Emergencies,’’ which allows the
licensee to not meet the requirements of
10 CFR 26.205(c) and (d) during
declared emergencies as defined in the
licensee’s emergency plan. The 10 CFR
Part 26 Statement of Consideration (73
FR 17148; March 31, 2008), states that
‘‘Plant emergencies are extraordinary
circumstances that may be most
effectively addressed through staff
augmentation that can only be
practically achieved through the use of
work hours in excess of the limits of
§ 26.205(c) and (d).’’
Accordingly, the Commission has
determined that, pursuant to 10 CFR
26.9, the exemption is authorized by
law, will not endanger life or property
or the common defense and security,
and is otherwise in the public interest.
Therefore, the Commission hereby
grants the licensee an exemption from
the requirements of 10 CFR 26.205(c)
and (d) for Calvert Cliffs.
Pursuant to 10 CFR 51.32, the
Commission has determined that the
granting of this exemption will not have
a significant effect on the quality of the
human environment.
This exemption is effective upon
issuance.
Dated at Rockville, Maryland, this 31st day
of July 2012.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor
Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. 2012–19268 Filed 8–6–12; 8:45 am]
47123
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 July 12,
2012 to July 25, 2012. The last biweekly
notice was published on July 24, 2012
(77 FR 43374).
ADDRESSES: You may access information
and comment submissions related to
this document, which the NRC
possesses and are publicly available, by
searching on https://www.regulations.gov
under Docket ID NRC–2012–0181. You
may submit comments by any of the
following methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2012–0181. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–492–3668;
email: Carol.Gallagher@nrc.gov.
• 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:
BILLING CODE 7590–01–P
I. Accessing Information and
Submitting Comments
NUCLEAR REGULATORY
COMMISSION
A. Accessing Information
Please refer to Docket ID NRC–2012–
0181 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
https://www.regulations.gov and search
for Docket ID NRC–2012–0181.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may access publicly
available documents online in the NRC
Library at https://www.nrc.gov/readingrm/adams.html. To begin the search,
select ‘‘ADAMS Public Documents’’ and
then select ‘‘Begin Web-based ADAMS
Search.’’ For problems with ADAMS,
please contact the NRC’s Public
Document Room (PDR) reference staff at
1–800–397–4209, 301–415–4737, or by
email to pdr.resource@nrc.gov.
Documents may be viewed in ADAMS
[NRC–2012–0181]
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
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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.
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B. Submitting Comments
Please include Docket ID NRC–2012–
0181 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 in
comment submissions that you do not
want to be publicly disclosed. The NRC
posts all comment submissions at
https://www.regulations.gov as well as
entering the comment submissions into
ADAMS, and the NRC does not 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 in
their comment submissions that they do
not want to be publicly disclosed. Your
request should state that the NRC will
not 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
Title 10 of the Code of Federal
Regulations (10 CFR) 50.92, 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
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considered in making any final
determination.
Normally, the Commission will not
issue the amendment until the
expiration of 60 days after the date of
publication of this notice. The
Commission may issue the license
amendment before expiration of the 60day period provided that its final
determination is that the amendment
involves no significant hazards
consideration. In addition, the
Commission may issue the amendment
prior to the expiration of the 30-day
comment period should circumstances
change during the 30-day comment
period such that failure to act in a
timely way would result, for example in
derating or shutdown of the facility.
Should the Commission take action
prior to the expiration of either the
comment period or the notice period, it
will publish in the Federal Register a
notice of issuance. Should the
Commission make a final No Significant
Hazards Consideration Determination,
any hearing will take place after
issuance. The Commission expects that
the need to take this action will occur
very infrequently.
Within 60 days after the date of
publication of this notice, any person(s)
whose interest may be affected by this
action may file a request for a hearing
and a petition to intervene with respect
to issuance of the amendment to the
subject facility operating license or
combined license. Requests for a
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 https://www.nrc.gov/reading-rm/
doc-collections/cfr/. If a request for a
hearing or petition for leave to intervene
is filed by the above date, the
Commission or a presiding officer
designated by the Commission or by the
Chief Administrative Judge of the
Atomic Safety and Licensing Board
Panel, will rule on the request and/or
petition; and the Secretary or the Chief
Administrative Judge of the Atomic
Safety and Licensing Board will issue a
notice of a hearing or an appropriate
order.
As required by 10 CFR 2.309, a
petition for leave to intervene shall set
forth with particularity the interest of
the petitioner in the proceeding, and
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how that interest may be affected by the
results of the proceeding. The petition
should specifically explain the reasons
why intervention should be permitted
with particular reference to the
following general requirements: (1) The
name, address, and telephone number of
the requestor or petitioner; (2) the
nature of the requestor’s/petitioner’s
right under the Act to be made a party
to the proceeding; (3) the nature and
extent of the requestor’s/petitioner’s
property, financial, or other interest in
the proceeding; and (4) the possible
effect of any decision or order which
may be entered in the proceeding on the
requestor’s/petitioner’s interest. The
petition must also identify the specific
contentions which the requestor/
petitioner seeks to have litigated at the
proceeding.
Each contention must consist of a
specific statement of the issue of law or
fact to be raised or controverted. In
addition, the requestor/petitioner shall
provide a brief explanation of the bases
for the contention and a concise
statement of the alleged facts or expert
opinion which support the contention
and on which the requestor/petitioner
intends to rely in proving the contention
at the hearing. The requestor/petitioner
must also provide references to those
specific sources and documents of
which the petitioner is aware and on
which the requestor/petitioner intends
to rely to establish those facts or expert
opinion. The petition must include
sufficient information to show that a
genuine dispute exists with the
applicant on a material issue of law or
fact. Contentions shall be limited to
matters within the scope of the
amendment under consideration. The
contention must be one which, if
proven, would entitle the requestor/
petitioner to relief. A requestor/
petitioner who fails to satisfy these
requirements with respect to at least one
contention will not be permitted to
participate as a party.
Those permitted to intervene become
parties to the proceeding, subject to any
limitations in the order granting leave to
intervene, and have the opportunity to
participate fully in the conduct of the
hearing.
If a hearing is requested, the
Commission will make a final
determination on the issue of no
significant hazards consideration. The
final determination will serve to decide
when the hearing is held. If the final
determination is that the amendment
request involves no significant hazards
consideration, the Commission may
issue the amendment and make it
immediately effective, notwithstanding
the request for a hearing. Any hearing
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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 EFiling process requires participants to
submit and serve all adjudicatory
documents over the internet, or in some
cases to mail copies on electronic
storage media. Participants may not
submit paper copies of their filings
unless they seek an exemption in
accordance with the procedures
described below.
To comply with the procedural
requirements of E-Filing, at least 10
days prior to the filing deadline, the
participant should contact the Office of
the Secretary by email at
hearing.docket@nrc.gov, or by telephone
at 301–415–1677, to request (1) a digital
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 NRCissued digital ID certificate). Based upon
this information, the Secretary will
establish an electronic docket for the
hearing in this proceeding if the
Secretary has not already established an
electronic docket.
Information about applying for a
digital ID certificate is available on the
NRC’s public Web site at https://
www.nrc.gov/site-help/e-submittals/
apply-certificates.html. System
requirements for accessing the ESubmittal server are detailed in the
NRC’s ‘‘Guidance for Electronic
Submission,’’ which is available on the
agency’s public Web site at https://
www.nrc.gov/site-help/esubmittals.html. Participants may
attempt to use other software not listed
on the Web site, but should note that the
NRC’s E-Filing system does not support
unlisted software, and the NRC Meta
System Help Desk will not be able to
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offer assistance in using unlisted
software.
If a participant is electronically
submitting a document to the NRC in
accordance with the E-Filing rule, the
participant must file the document
using the NRC’s online, Web-based
submission form. In order to serve
documents through the Electronic
Information Exchange System, users
will be required to install a Web
browser plug-in from the NRC’s Web
site. Further information on the Webbased submission form, including the
installation of the Web browser plug-in,
is available on the NRC’s public Web
site at https://www.nrc.gov/site-help/esubmittals.html.
Once a participant has obtained a
digital ID certificate and a docket has
been created, the participant can then
submit a request for hearing or petition
for leave to intervene. Submissions
should be in Portable Document Format
(PDF) in accordance with the NRC
guidance available on the NRC’s public
Web site at https://www.nrc.gov/sitehelp/e-submittals.html. A filing is
considered complete at the time the
documents are submitted through the
NRC’s E-Filing system. To be timely, an
electronic filing must be submitted to
the E-Filing system no later than 11:59
p.m. Eastern Time on the due date.
Upon receipt of a transmission, the EFiling system time-stamps the document
and sends the submitter an email notice
confirming receipt of the document. The
E-Filing system also distributes an email
notice that provides access to the
document to the NRC’s Office of the
General Counsel and any others who
have advised the Office of the Secretary
that they wish to participate in the
proceeding, so that the filer need not
serve the documents on those
participants separately. Therefore,
applicants and other participants (or
their counsel or representative) must
apply for and receive a digital ID
certificate before a hearing request/
petition to intervene is filed so that they
can obtain access to the document via
the E-Filing system.
A person filing electronically using
the agency’s adjudicatory E-Filing
system may seek assistance by
contacting the NRC Meta System Help
Desk through the ‘‘Contact Us’’ link
located on the NRC’s Web site at
https://www.nrc.gov/site-help/esubmittals.html, by email at
MSHD.Resource@nrc.gov, or by a tollfree call at 1–866 672–7640. The NRC
Meta System Help Desk is available
between 8 a.m. and 8 p.m., Eastern
Time, Monday through Friday,
excluding government holidays.
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Participants who believe that they
have a good cause for not submitting
documents electronically must file an
exemption request, in accordance with
10 CFR 2.302(g), with their initial paper
filing requesting authorization to
continue to submit documents in paper
format. Such filings must be submitted
by: (1) First class mail addressed to the
Office of the Secretary of the
Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001, Attention: Rulemaking and
Adjudications Staff; or (2) courier,
express mail, or expedited delivery
service to the Office of the Secretary,
Sixteenth Floor, One White Flint North,
11555 Rockville Pike, Rockville,
Maryland, 20852, Attention:
Rulemaking and Adjudications Staff.
Participants filing a document in this
manner are responsible for serving the
document on all other participants.
Filing is considered complete by firstclass mail as of the time of deposit in
the mail, or by courier, express mail, or
expedited delivery service upon
depositing the document with the
provider of the service. A presiding
officer, having granted an exemption
request from using E-Filing, may require
a participant or party to use E-Filing if
the presiding officer subsequently
determines that the reason for granting
the exemption from use of E-Filing no
longer exists.
Documents submitted in adjudicatory
proceedings will appear in the NRC’s
electronic hearing docket which is
available to the public at https://
ehd1.nrc.gov/ehd/, unless excluded
pursuant to an order of the Commission,
or the presiding officer. Participants are
requested not to include personal
privacy information, such as social
security numbers, home addresses, or
home phone numbers in their filings,
unless an NRC regulation or other law
requires submission of such
information. 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. Nontimely filings will not be entertained
absent a determination by the presiding
officer that the petition or request
should be granted or the contentions
should be admitted, based on a
balancing of the factors specified in 10
CFR 2.309(c)(1)(i)–(viii).
For further details with respect to this
license amendment application, see the
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application for amendment which is
available for public inspection at the
NRC’s PDR, located at One White Flint
North, Room O1–F21, 11555 Rockville
Pike (first floor), Rockville, Maryland
20852. Publicly available documents
created or received at the NRC are
accessible electronically through
ADAMS in the NRC Library at https://
www.nrc.gov/reading-rm/adams.html.
Persons who do not have access to
ADAMS or who encounter problems in
accessing the documents located in
ADAMS, should contact the NRC’s PDR
Reference staff at 1–800–397–4209, 301–
415–4737, or by email to
pdr.resource@nrc.gov.
Carolina Power and Light Company,
Docket No. 50–261, H. B. Robinson
Steam Electric Plant, Unit 2, (HBRSEP)
Darlington County, South Carolina
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Date of amendment request: June 8,
2012.
Description of amendment request:
The proposed change would revise the
Technical Specifications (TSs) 3.1.4,
‘‘Rod Group Alignment Limits,’’ and TS
3.1.7, ‘‘Rod Position Indication,’’ to
allow up to 1 hour of soak time
following substantial rod movement
during which individual rod position
indicators may not be within its limits.
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.
This license amendment request proposes
to allow up to one hour of soak time
following substantial rod movement during
which time the rod position indication may
be outside its limits. This would allow an
additional hour for rod position indication to
be inoperable or a control rod to be
misaligned prior to entry into a TS LCO
[Limiting Condition for Operation] Condition
and Required Actions. RPI [Rod Position
Indicators] instrumentation is not an
assumed accident initiator; however, the
HBRSEP, Unit No. 2 safety analyses consider
two types of rod misalignment events, static
misalignment and a dropped rod.
The safety analyses show that for the static
misalignment event, without any operator
intervention, a single fully withdrawn rod
event does not result in any fuel pin failure;
therefore, the static rod misalignment event
is not time dependent and an additional
hour, with the misalignment undetected and
unmitigated does not increase the
consequences of the event. Multiple rod
misalignment events are bounded by the
single rod misalignment analyses and
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therefore an additional hour would not have
any impact on this event.
The safety analyses also show that a single
dropped rod event, without any operator
intervention, does not result in any fuel pin
failure; therefore, the rod drop event is not
time dependent and an additional hour with
the misalignment undetected and
unmitigated does not increase the
consequences of the event. Multiple rod drop
events cause the reactor to trip and therefore
an additional hour would not have any
impact on that event.
Although this license amendment request
may allow a misaligned rod to be undetected
for an additional hour, the additional time for
discovery does not change the probability of
a misaligned control rod event because the
one hour time extension does not affect the
control rod drive system features that would
result in either type of misalignment.
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 previously evaluated?
Response: No.
This proposed change does not alter the
design, function, or operation of any plant
component and does not install any new or
different equipment. No new accident
scenarios, transient precursors, failure
mechanisms, or limiting single failures are
introduced as a result of these changes. No
new equipment performance burdens are
imposed.
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 the margin of safety?
Response: No.
The RPI system is an instrumentation
system that provides indication to the
operators that a control rod may be
misaligned. Inoperable individual RPI
instrumentation does not, by itself in any
way, harm or impact reactor operation.
Inoperable rod position indication may
impair the ability of the operators to detect
a misaligned rod. However, the impact of
inoperable RPI instrumentation may be offset
by availability of other indications that a rod
is misaligned such as nuclear
instrumentation indication that reactor
power has shifted to one side of the core or
thermocouple indication that the core
temperatures increased in one region of the
core and/or decreased in another region of
the core. Based on plant experience, the
likelihood of a misaligned rod at HBRSEP,
Unit No. 2 is considered to be small and the
likelihood of a misaligned rod coincident
with inoperable rod position indication
during the allowed one hour extension is
even smaller. In addition, these proposed
changes may enhance plant safety and
reliability because the one hour soak time
will allow the operators and engineers to
focus on monitoring the reactor performance
without unnecessary entry into TS LCO
Conditions and Required Actions.
The proposed change does not involve a
significant reduction in a margin of safety.
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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.
Florida Power and Light Company,
Docket Nos. 50–250 and 50–251, Turkey
Point Plant, Units 3 and 4, Miami-Dade
County, Florida
Date of amendment request: April 30,
2012.
Description of amendment request:
This amendment request proposes to
permanently revise technical
specification (TS) 6.8.4.j, Steam
Generator (SG) Surveillance Program, to
exclude portions of the SG tube below
the top of the SG tubesheet from
periodic tube inspections.
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 that alters the SG inspection and
reporting criteria 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 steam
line break (SLB) postulated accident.
Addressing 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-totubesheet 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
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tubesheet, and from the differential pressure
between the primary and secondary side, and
tubesheet rotation. The structural margins
against burst, as discussed in Regulatory
Guide (RG) 1.121, ‘‘Bases for Plugging
Degraded PWR [Pressurized-Water Reactors]
Steam Generator Tubes’’ [Reference 7] and
NEI [Nuclear Energy Institute] 97–06, ‘‘Steam
Generator Program Guidelines’’, [Reference 3]
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 primary water stress corrosion cracking
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 not affected by the primary to secondary
leakage flow during the event as primary to
secondary leakage flow through a postulated
tube that has been pulled out of the tubesheet
is essentially equivalent to a tube rupture.
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 SLB is unaffected by
the potential failure of a SG tube as the
failure of a tube is not an initiator for a SLB
event.
The leak rate factor of 1.82 for Turkey
Point Units 3 and 4, for a postulated SLB, has
been calculated as shown in References 2, 9
and 19. Turkey Point Units 3 and 4 will
apply the factor of 1.82 to the normal
operating leakage associated with the
tubesheet expansion region in the condition
monitoring (CM) and operational assessment
(OA). Through application of the limited
tubesheet inspection scope, the existing
operating leakage limit provides assurance
that excessive leakage (i.e., greater than
accident analysis assumptions) will not
occur. Multiplying the TS operational leak
rate limit of 150 gpd (at room temperature)
through any one SG by a factor of 1.82 shows
that the maximum primary to secondary
accident induced leak rate is limited to 273
gpd. This leakage rate is bounded by the
current licensing basis assumed primary to
secondary accident leak rate of 0.20 gpm (288
gpd) through any one SG for SLB. Since the
existing limit on operational leakage
continues to ensure that the SLB assumed
accident induced leakage will not be
exceeded, the consequences of a SLB
accident are not increased.
For the CM 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 OA, the
difference in the leakage between the
allowable leakage and the calculated accident
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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 change create the possibility of
a new or different kind of accident from any
accident previously evaluated?
Response: No
The proposed change that alters the SG
inspection and reporting criteria does not
introduce any new equipment, create new
failure modes for existing equipment, or
create any new limiting single failures. 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 a margin of safety?
Response: No
The proposed change defines the safety
significant portion of the tube that must be
inspected and repaired. WCAP–17345, Rev. 2
[Reference 9] identifies the specific
inspection depth 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 3] and TS 6.8.4.j,
‘‘Steam Generator (SG) Program.’’
The proposed change that alters the SG
inspection and reporting criteria 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 3], and NRC Regulatory Guide
(RG) 1.121, ‘‘Bases for Plugging Degraded
PWR Steam Generator Tubes’’ [Reference 7],
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. RG 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. RG 1.121 concludes
that by determining the limiting safe
conditions for tube wall degradation, the
probability and consequences of a SGTR are
reduced. This RG 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 2] and WCAP–17345, Rev. 2
[Reference 9] define a length of degradationfree expanded tubing that provides the
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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. The SLB leak rate factor for
Turkey Point Units 3 and 4 is 1.82 (Table 9–
7 in WCAP–17091–P). Multiplying the TS
operational leak rate limit of 150 gpd through
any one SG by the leak rate factor of 1.82
shows that the maximum primary to
secondary accident induced leak rate is
limited to 273 gpd. This leakage rate is
bounded by the current licensing basis
assumed primary to secondary accident leak
rate of 0.20 gpm (288 gpd) through any one
SG for SLB.
Therefore, the proposed change does not
involve a significant reduction in any margin
of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 50.92(c) are satisfied.
Therefore, the NRC staff proposes to
determine that the amendment request
involves no significant hazards
consideration.
Attorney for licensee: M.S. Ross,
Attorney, Florida Power & Light, P.O.
Box 14000, Juno Beach, Florida 33408–
0420.
NRC Acting Branch Chief: Jessie F.
Quichocho.
Nebraska Public Power District, Docket
No. 50–298, Cooper Nuclear Station,
Nemaha County, Nebraska
Date of amendment request: May 30,
2012.
Description of amendment request:
The proposed amendment would revise
Technical Specification Section 2.0,
‘‘Safety Limits.’’ Specifically, the
proposed amendment would revise two
recirculation loop and single
recirculation loop Safety Limit
Minimum Critical Power Ratio
(SLMCPR) values to reflect results of a
cycle-specific calculation.
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. Do the proposed changes involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
Four accidents have been evaluated
previously as reflected in the CNS [Cooper
Nuclear Station] Updated Safety Analysis
Report (USAR). These four accidents are (1)
loss-of-coolant, (2) control rod drop, (3) main
steam line break, and (4) fuel handling. The
probability of an evaluated accident is
derived from the probabilities of the
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individual precursors to that accident.
Changing the SLMCPR values does not
increase the probability of an evaluated
accident. The change does not require any
physical modifications to the plant or any
components, nor does it require a change in
plant operation. Therefore, no individual
precursors of an accident are affected.
The consequences of an evaluated accident
are determined by the operability of plant
systems designed to mitigate those
consequences. This proposed change makes
no modification to the design or operation of
the systems that are used in mitigation of
accidents. Limits have been established,
consistent with Nuclear Regulatory
Commission (NRC) approved methods, to
ensure that fuel performance during normal,
transient, and accident conditions is
acceptable. The proposed change to the
values of the SLMCPR continues to
conservatively establish this safety limit such
that the fuel is protected during normal
operation and during any plant transients or
anticipated operational occurrences.
Based on the above, NPPD [Nebraska
Public Power District] concludes that the
proposed change does not involve a
significant increase in the probability or
consequences of an accident previously
evaluated.
2. Do the proposed changes create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
Creation of the possibility of a new or
different kind of accident from an accident
previously evaluated would require creation
of precursors of that accident. New accident
precursors may be created by modification of
the plant configuration or changes in how the
plant is operated. The proposed change does
not involve a modification of the plant
configuration or in how the plant is operated.
The proposed change to the SLMCPR values
assures that safety criteria are maintained.
Based on the above, NPPD concludes that
the proposed change does not create the
possibility of a new or different kind of
accident from any previously evaluated.
3. Do the proposed changes involve a
significant reduction in a margin of safety?
Response: No.
The values of the proposed SLMCPR
provides a margin of safety by ensuring that
no more than 0.1% of fuel rods are expected
to be in boiling transition if the Minimum
Critical Power Ratio limit is not violated. The
proposed change will ensure the appropriate
level of fuel protection is maintained.
Additionally, operational limits are
established based on the proposed SLMCPR
to ensure that the SLMCPR is not violated
during all modes of operation. This will
ensure that the fuel design safety criteria are
met (i.e., that at least 99.9% of the fuel rods
do not experience transition boiling during
normal operation as well as anticipated
operational occurrences).
Based on the above, NPPD concludes that
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
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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. John C.
McClure, Nebraska Public Power
District, Post Office Box 499, Columbus,
NE 68602–0499.
NRC Branch Chief: Michael T.
Markley.
Omaha Public Power District, Docket
No. 50–285, Fort Calhoun Station, Unit
1, Washington County, Nebraska
Date of amendment request: February
10, 2012.
Description of amendment request:
The proposed amendment would
establish the limiting condition for
operation (LCO) requirements for the
reactor protective system (RPS)
actuation circuits in Technical
Specification (TS) 2.15,
‘‘Instrumentation and Control Systems.’’
Specifically, the proposed change:
renumbers LCO 2.15(1) through 2.15(4)
to 2.15.1(1) through 2.15.1(4),
renumbers LCO 2.15(5) to LCO 2.15.3
with an associated Table 2–6, and
implements a new LCO 2.15.2 for the
RPS logic and trip initiation channels.
The Table of Contents will also be
revised to reflect the renumbering and
addition of the LCO for the RPS logic
and trip initiation channels and the new
Table 2–6.
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 amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The reactor protective system logic and trip
initiation channels meets Criterion 3 of 10
CFR 50.36 for inclusion into Technical
Specification (TS) as a component that is part
of the primary success path and which
functions or actuates to mitigate a design
basis accident or transient. The TSs currently
does not have limiting conditions for
operations (LCO) specific for this circuitry,
but does contain surveillance requirements.
The addition of LCOs provides additional
restrictions on the operation of the plant and
provides required actions and time limits if
these components are incapable of
performing their function. As such, the
proposed change does not increase the
probability of an accident. The proposed
changes do not alter the physical design of
the RPS, or any other plant structure, system
or component (SSC) at Fort Calhoun Station
(FCS).
The proposed changes conform to the
Nuclear Regulatory Commission’s (NRC’s)
PO 00000
Frm 00101
Fmt 4703
Sfmt 4703
regulatory guidance regarding the content of
plant TS as identified in 10 CFR 50.36 and
NRC publication NUREG 1432.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed TS changes do not alter the
physical design, safety limits, or safety
analysis assumptions associated with the
operation of the plant. Hence, the proposed
changes do not introduce any new accident
initiators, nor do they reduce or adversely
affect the capabilities of any plant structure
or system in the performance of their safety
function.
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 TS operability requirements for the
RPS logic and trip initiation channels ensure
there is adequate components operable to
assure safe reactor operation and are
necessary to ensure safety systems
accomplish their safety function for design
basis accident events. The proposed TS
would revise the applicability for when the
RPS logic and trip initiation channels are
required to be operable to include whenever
control element assemblies (CEAs) are
capable of being withdrawn and the reactor
coolant system (RCS) is not at refueling boron
concentration. When the RCS boron
concentration is at refueling boron
concentration, or when no more than one
trippable control rod is capable of being
withdrawn, the RPS function is already
fulfilled. These proposed TS changes for the
RPS are aligned with the applicability and
operability requirements provided in NUREG
1432.
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: David A. Repka,
Esq., Winston & Strawn, 1700 K Street,
NW., Washington, DC 20006–3817.
NRC Branch Chief: Michael T.
Markley.
ZionSolutions LLC, Docket Nos. 50–295
and 50–304, Zion Nuclear Power Station
(Zion), Units 1 and 2, Lake County,
Illinois
Date of amendment request: May 31,
2012.
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Description of amendment request:
The proposed amendments would
approve methods of analysis, use of the
upgraded fuel handling building crane
system as a single-failure proof crane,
and a NUREG 0612 compliant heavy
loads handling program.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the change involve a significant
increase in the probability or consequences
of an accident previously evaluated?
Response: No.
The existing DSAR [Defueled Safety
Analysis Report] analysis assumes that a
spent fuel cask drop occurs. In this analysis,
the physics of the drop, coupled with
concrete bumpers on the cask loading pit and
pool edge were used to demonstrate that a
postulated drop of the spent fuel cask near
the Spent Fuel Pool neither impacted the
spent fuel directly nor damaged the pool
structure in a manner that adversely affected
the spent fuel, when a cask was to be
handled in the cask loading pit. The
proposed License Amendment Request to
operate a single-failure proof Fuel Building
Crane demonstrates that no analysis is
required for the cask drop event based on the
design and the associated programmatic
controls. A drop of the spent fuel cask
handled with a single-failure proof crane
(designed to ASME NOG–1 [‘‘Rules for
Construction of Overhead and Gantry Cranes
(Top Running Bridge, Multiple Girder)’’] and
compliant with NUREG–0554 [‘‘SingleFailure-Proof Cranes for Nuclear Power
Plants’’, ML110450636]), operated in
accordance with the administrative controls
of NUREG–0612 [‘‘Control of Heavy Loads at
Nuclear Power Plants,’’ ML070250180] has
an acceptably low probability so as to
effectively preclude consideration of the
event. The risk of such a drop event using the
new single-failure proof crane operated in
accordance with the Heavy Loads Program
procedures, qualitatively, is lower than the
event previously analyzed which postulate
the event without evaluation of its
likelihood.
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 location and design functions of the
Fuel Building crane are not changed from
those currently described in the DSAR.
Because the new crane has a single-failure
proof design the uncontrolled lowering, or
drop, of a heavy load will not be considered
credible. Evaluations show that individual
malfunctions or component failures of the
crane will not result in load drop. The new
single-failure proof crane[’s] primary use[s]
will be to move a loaded or unloaded
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MAGNASTOR transfer cask between the cask
loading pit [and] the decontamination pit,
and transfer [the cask] to the low profile cart
rail transport in the Fuel Handling Building.
No components that are classified as
Important to the Defueled Condition, other
than the Fuel Building crane, will be affected
by these movements. Based on the design
and programmatic controls on the crane, no
load will lower uncontrollably or drop in or
around the spent fuel pool or near an open
cask containing spent fuel nor will a cask
containing spent fuel drop or be lowered
uncontrollably during operation of the crane.
Hence no new accidents will be initiated.
Therefore, the proposed change will not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the change involve a significant
reduction in a margin of safety?
Response: No.
This proposed License Amendment
Request involves the replacement of the
existing non-single-failure proof Fuel
Building Crane with a new single-failure
proof crane. The new crane has been
designed to meet the specifications found in
ASME NOG–1–2004, which has been
endorsed by the NRC in RIS 2005–25, as
supplemented, as an acceptable means of
meeting the criteria in NUREG–0554,
‘‘Single-failure Proof Cranes for Nuclear
Power Plants.’’ to provide adequate
protection and safety margin against the
uncontrolled lowering of the lifted load. The
occurrence of a cask load drop accident is
considered not credible when the load is
lifted with a single-failure proof lifting
system meeting the guidance in NUREG–
0612, ‘‘Control of Heavy Loads at Nuclear
Power Plants’’ Section 5.1.6, ‘‘Single-FailureProof Handling Systems.’’ As a result, the
proposed change, replacing the existing nonsingle-failure proof crane, has no adverse
impact on stored spent fuel, or structural
integrity of the pool.
The configuration of the crane and the
primary load, a spent fuel cask containing
spent fuel, is changed from that of the DSAR.
The specific analysis dealing with a drop of
the cask will no longer be applicable and
[will be] removed from the DSAR, since the
new single-proof crane makes that event of
low enough probability to not be considered
credible. The maximum critical lift capacity
of the crane has not been changed, though
the load to be lifted is larger. The structural
analyses of the crane and its support
structure, however, show acceptable margin
under the acceptance criteria of NOG–I for
operation of the crane.
Therefore, the proposed change does not
involve a significant reduction in the margin
of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
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: Russ Workman,
Deputy General Counsel,
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47129
EnergySolutions, 423 West 300 South,
Suite 200, Salt Lake City, UT 84101.
NRC Branch Chief: Bruce Watson.
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 Commission’s Public Document
Room (PDR), located at One White Flint
North, Room O1–F21, 11555 Rockville
Pike (first floor), Rockville, Maryland
20852. Publicly available documents
created or received at the NRC are
accessible electronically through the
Agencywide Documents Access and
Management System (ADAMS) in the
NRC Library at https://www.nrc.gov/
reading-rm/adams.html. If you do not
have access to ADAMS or if there are
problems in accessing the documents
located in ADAMS, contact the PDR’s
Reference staff at 1–800–397–4209, 301–
415–4737 or by email to
pdr.resource@nrc.gov.
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Duke Energy Carolinas, LLC, et al.,
Docket Nos. 50–413 and 50–414,
Catawba Nuclear Station, Units 1 and 2,
York County, South Carolina
Date of application for amendments:
July 21, 2011.
Brief description of amendments: The
amendments revised Technical
Specifications 3.3.2, ‘‘Engineered Safety
Feature Actuation System (ESFAS)
Instrumentation,’’ 3.5.4, ‘‘Refueling
Water Storage Tank (RWST),’’ and 3.6.6,
‘‘Containment Spray System.’’
Date of issuance: July 25, 2012.
Effective date: As of the date of
issuance and shall be implemented
within 30 days from the date of
issuance.
Amendment Nos.: Unit 1–269 and
Unit 2–265.
Renewed Facility Operating License
Nos. NPF–35 and NPF–52: Amendments
revised the licenses and the technical
specifications.
Date of initial notice in Federal
Register: March 20, 2012 (77 FR 16274).
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated July 25, 2012.
No significant hazards consideration
comments received: No.
mstockstill on DSK4VPTVN1PROD with NOTICES
Entergy Nuclear Operations, Inc.,
Docket Nos. 50–247 and 50–286, Indian
Point Nuclear Generating Units 2 and 3
(IP2 and IP3), Westchester County, New
York
Date of application for amendment:
July 8, 2009, as supplemented by letters
dated September 28, 2009, October 26,
2009, October 5, 2010, October 28, 2010,
July 28, 2011, August 23, 2011, October
28, 2011, December 15, 2011, January
11, 2012, March 2, 2012, April 23, 2012,
and May 7, 2012.
Brief description of amendment: The
amendment authorizes the transfer of
spent fuel from the IP3 spent fuel pool
to the IP2 spent fuel pool, using a
newly-designed shielded transfer
canister, for further transfer to the onsite Independent Spent Fuel Storage
Installation.
Date of issuance: July 13, 2012.
Effective date: As of the date of
issuance, and shall be implemented
within 30 days.
Amendment No.: 268 and 246.
Facility Operating License Nos. DPR–
26 and DPR–64: The amendment
revised the License and the Technical
Specifications.
Date of initial notice in Federal
Register: January 21, 2010 (75 FR
3497).
The supplements provided additional
information that clarified the
application but did not expand the
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scope of the application as originally
noticed.
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated July 13, 2012.
Entergy Operations, Inc., System Energy
Resources, Inc., South Mississippi
Electric Power Association, and Entergy
Mississippi, Inc., Docket No. 50–416,
Grand Gulf Nuclear Station, Unit 1,
Claiborne County, Mississippi
Date of application for amendment:
September 8, 2010, as supplemented by
letters dated November 18, 2010,
November 23, 2010, February 23, 2011
(four letters), March 9, 2011 (two
letters), March 22, 2011, March 30,
2011, March 31, 2011, April 14, 2011,
April 21, 2011, May 3, 2011, May 5,
2011, May 11, 2011, June 8, 2011, June
15, 2011, June 21, 2011, June 23, 2011,
July 6, 2011, July 28, 2011, August 25,
2011, August 29, 2011, August 30, 2011,
September 2, 2011, September 9, 2011,
September 12, 2011, September 15,
2011, September 26, 2011, October 10,
2011, October 24, 2011, November 14,
2011, November 25, 2011, November 28,
2011, December 19, 2011, February 6,
2012, February 15, 2012, February 20,
2012, March 13, 2012, March 21, 2012,
April 5, 2012, April 18, 2012 (two
letters), April 26, 2012, May 9, 2012,
and June 12, 2012.
Brief description of amendment: The
amendment increased the maximum
steady-state reactor core power level
from 3,898 megawatts thermal (MWt) to
4,408 MWt, which is an increase of
approximately 15 percent from the
original licensed thermal power level of
3,833 MWt. The proposed increase in
power level is considered an extended
power uprate.
Date of issuance: July 18, 2012.
Effective date: As of the date of
issuance and shall be implemented
within 120 days of issuance.
Amendment No: 191.
Facility Operating License No. NPF–
29: The amendment revised the Facility
Operating License and Technical
Specifications.
Date of initial notice in Federal
Register: January 10, 2011 (76 FR
1464). The supplemental letters dated
November 18, 2010, November 23, 2010,
February 23, 2011 (four letters), March
9, 2011 (two letters), March 22, 2011,
March 30, 2011, March 31, 2011, April
14, 2011, April 21, 2011, May 3, 2011,
May 5, 2011, May 11, 2011, June 8,
2011, June 15, 2011, June 21, 2011, June
23, 2011, July 6, 2011, July 28, 2011,
August 25, 2011, August 29, 2011,
August 30, 2011, September 2, 2011,
September 9, 2011, September 12, 2011,
September 15, 2011, September 26,
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Fmt 4703
Sfmt 4703
2011, October 10, 2011, October 24,
2011, November 14, 2011, November 25,
2011, November 28, 2011, December 19,
2011, February 6, 2012, February 15,
2012, February 20, 2012, March 13,
2012, March 21, 2012, April 5, 2012,
April 18, 2012 (two letters), April 26,
2012, May 9, 2012, and June 12, 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 July 18, 2012.
No significant hazards consideration
comments received: No.
Nine Mile Point Nuclear Station, LLC,
Docket Nos. 50–220, and 50–410, Nine
Mile Point Nuclear Station, Units 1 and
2, Oswego County, New York
Date of application for amendments:
July 20, 2011, as supplemented on
November 3, 2011, and January 12,
2012.
Brief description of amendments: The
amendments revised the NMP1
Technical Specification (TS) Section
5.1, ‘‘Site,’’ and associated TS Figure
5.1–1, ‘‘Site Boundaries, Nine Mile
Point–Unit 1,’’ and the NMP2 TS Figure
4.1–1, ‘‘Site Area and Land Portion of
Exclusion Area Boundaries,’’ to reflect
the transfer of a portion of the Nine Mile
Point Nuclear Station, LLC (NMPNS)
site real property located outside of the
NMPNS Protected Area but within the
current NMPNS Owner Controlled Area,
as well as specified easements over the
remainder of the NMPNS site, to Nine
Mile Point 3 Nuclear Project, LLC
(NMP3), a subsidiary of UniStar Nuclear
Energy, LLC.
Date of issuance: July 12, 2012.
Effective date: As of the date of
issuance to be implemented within 30
days.
Amendment Nos.: 212 for Unit 1 and
142 for Unit 2.
Renewed Facility Operating License
Nos. DPR–63 and NPF–69: Amendments
revised the License and Technical
Specifications.
Date of initial notice in Federal
Register: December 27, 2011 (76 FR
80977).
The supplements dated November 3,
2011, and January 12, 2012, provided
additional information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the Nuclear
Regulatory Commission (NRC) staff’s
initial proposed no significant hazards
consideration determination noticed in
E:\FR\FM\07AUN1.SGM
07AUN1
Federal Register / Vol. 77, No. 152 / Tuesday, August 7, 2012 / Notices
the Federal Register on December 27,
2011 (76 FR 80977).
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated July 12, 2012.
No significant hazards consideration
comments received: No.
Dated at Rockville, Maryland, this 26th day
of July 2012.
For the Nuclear Regulatory Commission.
Louise Lund,
Deputy Director, Division of Operating
Reactor Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. 2012–19004 Filed 8–6–12; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[NRC–2012–0002]
Notice of Sunshine Act Meeting
AGENCY HOLDING THE MEETINGS: Nuclear
Regulatory Commission.
DATES: Weeks of August 6, 13, 20, 27,
September 3, 10, 2012.
PLACE: Commissioners’ Conference
Room, 11555 Rockville Pike, Rockville,
Maryland.
STATUS: Public and closed.
Week of August 6, 2012
Tuesday, August 7, 2012
8:55 a.m. Affirmation Session (Public
Meeting) (Tentative)
a. Calvert Cliffs Nuclear Project, L.L.C.
(Calvert Cliffs Nuclear Power Plant,
Unit 3), et al., Petition to Suspend
Final Decisions on Reactor License
Applications Pending Completion
of Remanded Waste Confidence
Proceeding (June 18, 2012)
(Tentative)
This meeting will be webcast live at
the Web address—www.nrc.gov.
9 a.m. Briefing on the Status of
Lessons Learned from the
Fukushima Dai-ichi Accident
(Public Meeting) (Contact: John
Monninger, 301–415–0610)
This meeting will be webcast live at
the Web address—www.nrc.gov.
mstockstill on DSK4VPTVN1PROD with NOTICES
Week of August 13, 2012—Tentative
Week of August 20, 2012—Tentative
There are no meetings scheduled for
the week of August 20, 2012.
Week of August 27, 2012—Tentative
There are no meetings scheduled for
the week of August 27, 2012.
16:52 Aug 06, 2012
Jkt 226001
Week of September 10, 2012—Tentative
Tuesday, September 11, 2012
9 a.m. Briefing on Economic
Consequences (Public Meeting)
(Contact: Richard Correia, 301–251–
7430)
This meeting will be webcast live at
the Web address—www.nrc.gov.
*
*
*
*
*
*The schedule for Commission
meetings is subject to change on short
notice. To verify the status of meetings,
call (recording)—301–415–1292.
Contact person for more information:
Rochelle Bavol, 301–415–1651.
*
*
*
*
*
The NRC Commission Meeting
Schedule can be found on the Internet
at: https://www.nrc.gov/public-involve/
public-meetings/schedule.html.
*
*
*
*
*
The NRC provides reasonable
accommodation to individuals with
disabilities where appropriate. If you
need a reasonable accommodation to
participate in these public meetings, or
need this meeting notice or the
transcript or other information from the
public meetings in another format (e.g.
braille, large print), please notify Bill
Dosch, Chief, Work Life and Benefits
Branch, at 301–415–6200, TDD: 301–
415–2100, or by email at
william.dosch@nrc.gov. Determinations
on requests for reasonable
accommodation will be made on a caseby-case basis.
*
*
*
*
*
This notice is distributed
electronically to subscribers. If you no
longer wish to receive it, or would like
to be added to the distribution, please
contact the Office of the Secretary,
Washington, DC 20555 (301–415–1969),
or send an email to
darlene.wright@nrc.gov.
Dated: August 2, 2012.
Richard J. Laufer,
Technical Coordinator, Office of the
Secretary.
[FR Doc. 2012–19380 Filed 8–3–12; 11:15 am]
BILLING CODE 7590–01–P
There are no meetings scheduled for
the week of August 13, 2012.
VerDate Mar<15>2010
Week of September 3, 2012—Tentative
There are no meetings scheduled for
the week of September 3, 2012.
SECURITIES AND EXCHANGE
COMMISSION
Sunshine Act Meeting Notice
Notice is hereby given, pursuant to
the provisions of the Government in the
Sunshine Act, Public Law 94–409, that
the Securities and Exchange
PO 00000
Frm 00104
Fmt 4703
Sfmt 4703
47131
Commission will hold a Closed Meeting
on Thursday, August 9, 2012 at 2 p.m.
Commissioners, Counsel to the
Commissioners, the Secretary to the
Commission, and recording secretaries
will attend the Closed Meeting. Certain
staff members who have an interest in
the matters also may be present.
The General Counsel of the
Commission, or his designee, has
certified that, in his opinion, one or
more of the exemptions set forth in 5
U.S.C. 552b(c)(3), (5), (7), 9(B) and (10)
and 17 CFR 200.402(a)(3), (5), (7), 9(ii)
and (10), permit consideration of the
scheduled matters at the Closed
Meeting.
Commissioner Gallagher, as duty
officer, voted to consider the items
listed for the Closed Meeting in a closed
session.
The subject matter of the Closed
Meeting scheduled for Thursday,
August 9, 2012 will be:
Institution and settlement of injunctive
actions;
Institution and settlement of
administrative proceedings;
A litigation matter; and
Other matters relating to enforcement
proceedings.
At times, changes in Commission
priorities require alterations in the
scheduling of meeting items.
For further information and to
ascertain what, if any, matters have been
added, deleted or postponed, please
contact: The Office of the Secretary at
(202) 551–5400.
Dated: August 2, 2012.
Lynn M. Powalski,
Deputy Secretary.
[FR Doc. 2012–19340 Filed 8–3–12; 11:15 am]
BILLING CODE 8011–01–P
SECURITIES AND EXCHANGE
COMMISSION
[Release No. 34–67552; File No. SR–
NYSEArca–2012–55]
Self-Regulatory Organizations; NYSE
Arca, Inc.; Order Granting Approval of
Proposed Rule Change Relating to the
Listing and Trading of the STARTM
Global Buy-Write ETF Under NYSE
Arca Equities Rule 8.600
August 1, 2012.
I. Introduction
On May 31, 2012, NYSE Arca, Inc.
(‘‘Exchange’’ or ‘‘NYSE Arca’’) filed
with the Securities and Exchange
Commission (‘‘Commission’’), pursuant
to Section 19(b)(1) of the Securities
Exchange Act of 1934 (‘‘Act’’ or
E:\FR\FM\07AUN1.SGM
07AUN1
]]>Agencies
[Federal Register Volume 77, Number 152 (Tuesday, August 7, 2012)]
[Notices]
[Pages 47123-47131]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2012-19004]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2012-0181]
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 July 12, 2012 to July 25, 2012. The last
biweekly notice was published on July 24, 2012 (77 FR 43374).
ADDRESSES: You may access information and comment submissions related
to this document, which the NRC possesses and are publicly available,
by searching on https://www.regulations.gov under Docket ID NRC-2012-
0181. You may submit comments by any of the following methods:
Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2012-0181. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: Carol.Gallagher@nrc.gov.
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-2012-0181 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 https://www.regulations.gov and search for Docket ID NRC-2012-0181.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly available documents online in the NRC
Library at https://www.nrc.gov/reading-rm/adams.html. To begin the
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to pdr.resource@nrc.gov. Documents may be viewed in
ADAMS
[[Page 47124]]
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-2012-0181 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 in comment submissions that you do not want to be publicly
disclosed. The NRC posts all comment submissions at https://www.regulations.gov as well as entering the comment submissions into
ADAMS, and the NRC does not 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 in their comment submissions
that they do not want to be publicly disclosed. Your request should
state that the NRC will not 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 Title 10 of the Code of Federal
Regulations (10 CFR) 50.92, 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 https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or a presiding
officer designated by the Commission or by the Chief Administrative
Judge of the Atomic Safety and Licensing Board Panel, will rule on the
request and/or petition; and the Secretary or the Chief Administrative
Judge of the Atomic Safety and Licensing Board will issue a notice of a
hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing
[[Page 47125]]
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 hearing.docket@nrc.gov, 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 https://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at https://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with the NRC guidance
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC's Web site
at https://www.nrc.gov/site-help/e-submittals.html, by email at
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants. Filing is considered
complete by first-class mail as of the time of deposit in the mail, or
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information. 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. Non-timely filings
will not be entertained absent a determination by the presiding officer
that the petition or request should be granted or the contentions
should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
For further details with respect to this license amendment
application, see the
[[Page 47126]]
application for amendment which is available for public inspection at
the NRC's PDR, located at One White Flint North, Room O1-F21, 11555
Rockville Pike (first floor), Rockville, Maryland 20852. Publicly
available documents created or received at the NRC are accessible
electronically through ADAMS in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who
encounter problems in accessing the documents located in ADAMS, should
contact the NRC's PDR Reference staff at 1-800-397-4209, 301-415-4737,
or by email to pdr.resource@nrc.gov.
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: June 8, 2012.
Description of amendment request: The proposed change would revise
the Technical Specifications (TSs) 3.1.4, ``Rod Group Alignment
Limits,'' and TS 3.1.7, ``Rod Position Indication,'' to allow up to 1
hour of soak time following substantial rod movement during which
individual rod position indicators may not be within its limits.
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.
This license amendment request proposes to allow up to one hour
of soak time following substantial rod movement during which time
the rod position indication may be outside its limits. This would
allow an additional hour for rod position indication to be
inoperable or a control rod to be misaligned prior to entry into a
TS LCO [Limiting Condition for Operation] Condition and Required
Actions. RPI [Rod Position Indicators] instrumentation is not an
assumed accident initiator; however, the HBRSEP, Unit No. 2 safety
analyses consider two types of rod misalignment events, static
misalignment and a dropped rod.
The safety analyses show that for the static misalignment event,
without any operator intervention, a single fully withdrawn rod
event does not result in any fuel pin failure; therefore, the static
rod misalignment event is not time dependent and an additional hour,
with the misalignment undetected and unmitigated does not increase
the consequences of the event. Multiple rod misalignment events are
bounded by the single rod misalignment analyses and therefore an
additional hour would not have any impact on this event.
The safety analyses also show that a single dropped rod event,
without any operator intervention, does not result in any fuel pin
failure; therefore, the rod drop event is not time dependent and an
additional hour with the misalignment undetected and unmitigated
does not increase the consequences of the event. Multiple rod drop
events cause the reactor to trip and therefore an additional hour
would not have any impact on that event.
Although this license amendment request may allow a misaligned
rod to be undetected for an additional hour, the additional time for
discovery does not change the probability of a misaligned control
rod event because the one hour time extension does not affect the
control rod drive system features that would result in either type
of misalignment.
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 previously evaluated?
Response: No.
This proposed change does not alter the design, function, or
operation of any plant component and does not install any new or
different equipment. No new accident scenarios, transient
precursors, failure mechanisms, or limiting single failures are
introduced as a result of these changes. No new equipment
performance burdens are imposed.
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
the margin of safety?
Response: No.
The RPI system is an instrumentation system that provides
indication to the operators that a control rod may be misaligned.
Inoperable individual RPI instrumentation does not, by itself in any
way, harm or impact reactor operation. Inoperable rod position
indication may impair the ability of the operators to detect a
misaligned rod. However, the impact of inoperable RPI
instrumentation may be offset by availability of other indications
that a rod is misaligned such as nuclear instrumentation indication
that reactor power has shifted to one side of the core or
thermocouple indication that the core temperatures increased in one
region of the core and/or decreased in another region of the core.
Based on plant experience, the likelihood of a misaligned rod at
HBRSEP, Unit No. 2 is considered to be small and the likelihood of a
misaligned rod coincident with inoperable rod position indication
during the allowed one hour extension is even smaller. In addition,
these proposed changes may enhance plant safety and reliability
because the one hour soak time will allow the operators and
engineers to focus on monitoring the reactor performance without
unnecessary entry into TS LCO Conditions and Required Actions.
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: 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.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: April 30, 2012.
Description of amendment request: This amendment request proposes
to permanently revise technical specification (TS) 6.8.4.j, Steam
Generator (SG) Surveillance Program, to exclude portions of the SG tube
below the top of the SG tubesheet from periodic tube inspections.
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
that alters the SG inspection and reporting criteria 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 steam line break (SLB) postulated accident.
Addressing 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
[[Page 47127]]
tubesheet, and from the differential pressure between the primary
and secondary side, and tubesheet rotation. The structural margins
against burst, as discussed in Regulatory Guide (RG) 1.121, ``Bases
for Plugging Degraded PWR [Pressurized-Water Reactors] Steam
Generator Tubes'' [Reference 7] and NEI [Nuclear Energy Institute]
97-06, ``Steam Generator Program Guidelines'', [Reference 3] 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 primary water stress
corrosion cracking 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 not affected by the primary to secondary leakage flow
during the event as primary to secondary leakage flow through a
postulated tube that has been pulled out of the tubesheet is
essentially equivalent to a tube rupture. 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 SLB is unaffected by the potential failure
of a SG tube as the failure of a tube is not an initiator for a SLB
event.
The leak rate factor of 1.82 for Turkey Point Units 3 and 4, for
a postulated SLB, has been calculated as shown in References 2, 9
and 19. Turkey Point Units 3 and 4 will apply the factor of 1.82 to
the normal operating leakage associated with the tubesheet expansion
region in the condition monitoring (CM) and operational assessment
(OA). Through application of the limited tubesheet inspection scope,
the existing operating leakage limit provides assurance that
excessive leakage (i.e., greater than accident analysis assumptions)
will not occur. Multiplying the TS operational leak rate limit of
150 gpd (at room temperature) through any one SG by a factor of 1.82
shows that the maximum primary to secondary accident induced leak
rate is limited to 273 gpd. This leakage rate is bounded by the
current licensing basis assumed primary to secondary accident leak
rate of 0.20 gpm (288 gpd) through any one SG for SLB. Since the
existing limit on operational leakage continues to ensure that the
SLB assumed accident induced leakage will not be exceeded, the
consequences of a SLB accident are not increased.
For the CM 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 OA,
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 change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No
The proposed change that alters the SG inspection and reporting
criteria does not introduce any new equipment, create new failure
modes for existing equipment, or create any new limiting single
failures. 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 a margin
of safety?
Response: No
The proposed change defines the safety significant portion of
the tube that must be inspected and repaired. WCAP-17345, Rev. 2
[Reference 9] identifies the specific inspection depth 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 3] and TS 6.8.4.j, ``Steam Generator (SG)
Program.''
The proposed change that alters the SG inspection and reporting
criteria 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 3], and NRC
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR Steam
Generator Tubes'' [Reference 7], 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. RG 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.
RG 1.121 concludes that by determining the limiting safe conditions
for tube wall degradation, the probability and consequences of a
SGTR are reduced. This RG 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 2] and WCAP-17345, Rev. 2 [Reference 9] 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. The SLB leak rate factor for Turkey Point Units 3 and 4
is 1.82 (Table 9-7 in WCAP-17091-P). Multiplying the TS operational
leak rate limit of 150 gpd through any one SG by the leak rate
factor of 1.82 shows that the maximum primary to secondary accident
induced leak rate is limited to 273 gpd. This leakage rate is
bounded by the current licensing basis assumed primary to secondary
accident leak rate of 0.20 gpm (288 gpd) through any one SG for SLB.
Therefore, the proposed change does not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Acting Branch Chief: Jessie F. Quichocho.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: May 30, 2012.
Description of amendment request: The proposed amendment would
revise Technical Specification Section 2.0, ``Safety Limits.''
Specifically, the proposed amendment would revise two recirculation
loop and single recirculation loop Safety Limit Minimum Critical Power
Ratio (SLMCPR) values to reflect results of a cycle-specific
calculation.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Four accidents have been evaluated previously as reflected in
the CNS [Cooper Nuclear Station] Updated Safety Analysis Report
(USAR). These four accidents are (1) loss-of-coolant, (2) control
rod drop, (3) main steam line break, and (4) fuel handling. The
probability of an evaluated accident is derived from the
probabilities of the
[[Page 47128]]
individual precursors to that accident. Changing the SLMCPR values
does not increase the probability of an evaluated accident. The
change does not require any physical modifications to the plant or
any components, nor does it require a change in plant operation.
Therefore, no individual precursors of an accident are affected.
The consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. This proposed change makes no modification to the
design or operation of the systems that are used in mitigation of
accidents. Limits have been established, consistent with Nuclear
Regulatory Commission (NRC) approved methods, to ensure that fuel
performance during normal, transient, and accident conditions is
acceptable. The proposed change to the values of the SLMCPR
continues to conservatively establish this safety limit such that
the fuel is protected during normal operation and during any plant
transients or anticipated operational occurrences.
Based on the above, NPPD [Nebraska Public Power District]
concludes that the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Creation of the possibility of a new or different kind of
accident from an accident previously evaluated would require
creation of precursors of that accident. New accident precursors may
be created by modification of the plant configuration or changes in
how the plant is operated. The proposed change does not involve a
modification of the plant configuration or in how the plant is
operated. The proposed change to the SLMCPR values assures that
safety criteria are maintained.
Based on the above, NPPD concludes that the proposed change does
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The values of the proposed SLMCPR provides a margin of safety by
ensuring that no more than 0.1% of fuel rods are expected to be in
boiling transition if the Minimum Critical Power Ratio limit is not
violated. The proposed change will ensure the appropriate level of
fuel protection is maintained. Additionally, operational limits are
established based on the proposed SLMCPR to ensure that the SLMCPR
is not violated during all modes of operation. This will ensure that
the fuel design safety criteria are met (i.e., that at least 99.9%
of the fuel rods do not experience transition boiling during normal
operation as well as anticipated operational occurrences).
Based on the above, NPPD concludes that the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit 1, Washington County, Nebraska
Date of amendment request: February 10, 2012.
Description of amendment request: The proposed amendment would
establish the limiting condition for operation (LCO) requirements for
the reactor protective system (RPS) actuation circuits in Technical
Specification (TS) 2.15, ``Instrumentation and Control Systems.''
Specifically, the proposed change: renumbers LCO 2.15(1) through
2.15(4) to 2.15.1(1) through 2.15.1(4), renumbers LCO 2.15(5) to LCO
2.15.3 with an associated Table 2-6, and implements a new LCO 2.15.2
for the RPS logic and trip initiation channels. The Table of Contents
will also be revised to reflect the renumbering and addition of the LCO
for the RPS logic and trip initiation channels and the new Table 2-6.
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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The reactor protective system logic and trip initiation channels
meets Criterion 3 of 10 CFR 50.36 for inclusion into Technical
Specification (TS) as a component that is part of the primary
success path and which functions or actuates to mitigate a design
basis accident or transient. The TSs currently does not have
limiting conditions for operations (LCO) specific for this
circuitry, but does contain surveillance requirements. The addition
of LCOs provides additional restrictions on the operation of the
plant and provides required actions and time limits if these
components are incapable of performing their function. As such, the
proposed change does not increase the probability of an accident.
The proposed changes do not alter the physical design of the RPS, or
any other plant structure, system or component (SSC) at Fort Calhoun
Station (FCS).
The proposed changes conform to the Nuclear Regulatory
Commission's (NRC's) regulatory guidance regarding the content of
plant TS as identified in 10 CFR 50.36 and NRC publication NUREG
1432.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed TS changes do not alter the physical design, safety
limits, or safety analysis assumptions associated with the operation
of the plant. Hence, the proposed changes do not introduce any new
accident initiators, nor do they reduce or adversely affect the
capabilities of any plant structure or system in the performance of
their safety function.
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 TS operability requirements for the RPS logic and trip
initiation channels ensure there is adequate components operable to
assure safe reactor operation and are necessary to ensure safety
systems accomplish their safety function for design basis accident
events. The proposed TS would revise the applicability for when the
RPS logic and trip initiation channels are required to be operable
to include whenever control element assemblies (CEAs) are capable of
being withdrawn and the reactor coolant system (RCS) is not at
refueling boron concentration. When the RCS boron concentration is
at refueling boron concentration, or when no more than one trippable
control rod is capable of being withdrawn, the RPS function is
already fulfilled. These proposed TS changes for the RPS are aligned
with the applicability and operability requirements provided in
NUREG 1432.
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: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power
Station (Zion), Units 1 and 2, Lake County, Illinois
Date of amendment request: May 31, 2012.
[[Page 47129]]
Description of amendment request: The proposed amendments would
approve methods of analysis, use of the upgraded fuel handling building
crane system as a single-failure proof crane, and a NUREG 0612
compliant heavy loads handling program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The existing DSAR [Defueled Safety Analysis Report] analysis
assumes that a spent fuel cask drop occurs. In this analysis, the
physics of the drop, coupled with concrete bumpers on the cask
loading pit and pool edge were used to demonstrate that a postulated
drop of the spent fuel cask near the Spent Fuel Pool neither
impacted the spent fuel directly nor damaged the pool structure in a
manner that adversely affected the spent fuel, when a cask was to be
handled in the cask loading pit. The proposed License Amendment
Request to operate a single-failure proof Fuel Building Crane
demonstrates that no analysis is required for the cask drop event
based on the design and the associated programmatic controls. A drop
of the spent fuel cask handled with a single-failure proof crane
(designed to ASME NOG-1 [``Rules for Construction of Overhead and
Gantry Cranes (Top Running Bridge, Multiple Girder)''] and compliant
with NUREG-0554 [``Single-Failure-Proof Cranes for Nuclear Power
Plants'', ML110450636]), operated in accordance with the
administrative controls of NUREG-0612 [``Control of Heavy Loads at
Nuclear Power Plants,'' ML070250180] has an acceptably low
probability so as to effectively preclude consideration of the
event. The risk of such a drop event using the new single-failure
proof crane operated in accordance with the Heavy Loads Program
procedures, qualitatively, is lower than the event previously
analyzed which postulate the event without evaluation of its
likelihood.
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 location and design functions of the Fuel Building crane are
not changed from those currently described in the DSAR. Because the
new crane has a single-failure proof design the uncontrolled
lowering, or drop, of a heavy load will not be considered credible.
Evaluations show that individual malfunctions or component failures
of the crane will not result in load drop. The new single-failure
proof crane['s] primary use[s] will be to move a loaded or unloaded
MAGNASTOR transfer cask between the cask loading pit [and] the
decontamination pit, and transfer [the cask] to the low profile cart
rail transport in the Fuel Handling Building. No components that are
classified as Important to the Defueled Condition, other than the
Fuel Building crane, will be affected by these movements. Based on
the design and programmatic controls on the crane, no load will
lower uncontrollably or drop in or around the spent fuel pool or
near an open cask containing spent fuel nor will a cask containing
spent fuel drop or be lowered uncontrollably during operation of the
crane. Hence no new accidents will be initiated.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
This proposed License Amendment Request involves the replacement
of the existing non-single-failure proof Fuel Building Crane with a
new single-failure proof crane. The new crane has been designed to
meet the specifications found in ASME NOG-1-2004, which has been
endorsed by the NRC in RIS 2005-25, as supplemented, as an
acceptable means of meeting the criteria in NUREG-0554, ``Single-
failure Proof Cranes for Nuclear Power Plants.'' to provide adequate
protection and safety margin against the uncontrolled lowering of
the lifted load. The occurrence of a cask load drop accident is
considered not credible when the load is lifted with a single-
failure proof lifting system meeting the guidance in NUREG-0612,
``Control of Heavy Loads at Nuclear Power Plants'' Section 5.1.6,
``Single-Failure-Proof Handling Systems.'' As a result, the proposed
change, replacing the existing non-single-failure proof crane, has
no adverse impact on stored spent fuel, or structural integrity of
the pool.
The configuration of the crane and the primary load, a spent
fuel cask containing spent fuel, is changed from that of the DSAR.
The specific analysis dealing with a drop of the cask will no longer
be applicable and [will be] removed from the DSAR, since the new
single-proof crane makes that event of low enough probability to not
be considered credible. The maximum critical lift capacity of the
crane has not been changed, though the load to be lifted is larger.
The structural analyses of the crane and its support structure,
however, show acceptable margin under the acceptance criteria of
NOG-I for operation of the crane.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this 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: Russ Workman, Deputy General Counsel,
EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT
84101.
NRC Branch Chief: Bruce Watson.
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 Commission's Public Document Room (PDR), located at One White Flint
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. Publicly available documents created or received at the
NRC are accessible electronically through the Agencywide Documents
Access and Management System (ADAMS) in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by
email to pdr.resource@nrc.gov.
[[Page 47130]]
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: July 21, 2011.
Brief description of amendments: The amendments revised Technical
Specifications 3.3.2, ``Engineered Safety Feature Actuation System
(ESFAS) Instrumentation,'' 3.5.4, ``Refueling Water Storage Tank
(RWST),'' and 3.6.6, ``Containment Spray System.''
Date of issuance: July 25, 2012.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1-269 and Unit 2-265.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: March 20, 2012 (77 FR
16274).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 25, 2012.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Units 2 and 3 (IP2 and IP3), Westchester
County, New York
Date of application for amendment: July 8, 2009, as supplemented by
letters dated September 28, 2009, October 26, 2009, October 5, 2010,
October 28, 2010, July 28, 2011, August 23, 2011, October 28, 2011,
December 15, 2011, January 11, 2012, March 2, 2012, April 23, 2012, and
May 7, 2012.
Brief description of amendment: The amendment authorizes the
transfer of spent fuel from the IP3 spent fuel pool to the IP2 spent
fuel pool, using a newly-designed shielded transfer canister, for
further transfer to the on-site Independent Spent Fuel Storage
Installation.
Date of issuance: July 13, 2012.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 268 and 246.
Facility Operating License Nos. DPR-26 and DPR-64: The amendment
revised the License and the Technical Specifications.
Date of initial notice in Federal Register: January 21, 2010 (75 FR
3497).
The supplements provided additional information that clarified the
application but did not expand the scope of the application as
originally noticed.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 13, 2012.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: September 8, 2010, as
supplemented by letters dated November 18, 2010, November 23, 2010,
February 23, 2011 (four letters), March 9, 2011 (two letters), March
22, 2011, March 30, 2011, March 31, 2011, April 14, 2011, April 21,
2011, May 3, 2011, May 5, 2011, May 11, 2011, June 8, 2011, June 15,
2011, June 21, 2011, June 23, 2011, July 6, 2011, July 28, 2011, August
25, 2011, August 29, 2011, August 30, 2011, September 2, 2011,
September 9, 2011, September 12, 2011, September 15, 2011, September
26, 2011, October 10, 2011, October 24, 2011, November 14, 2011,
November 25, 2011, November 28, 2011, December 19, 2011, February 6,
2012, February 15, 2012, February 20, 2012, March 13, 2012, March 21,
2012, April 5, 2012, April 18, 2012 (two letters), April 26, 2012, May
9, 2012, and June 12, 2012.
Brief description of amendment: The amendment increased the maximum
steady-state reactor core power level from 3,898 megawatts thermal
(MWt) to 4,408 MWt, which is an increase of approximately 15 percent
from the original licensed thermal power level of 3,833 MWt. The
proposed increase in power level is considered an extended power
uprate.
Date of issuance: July 18, 2012.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No: 191.
Facility Operating License No. NPF-29: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 10, 2011 (76 FR
1464). The supplemental letters dated November 18, 2010, November 23,
2010, February 23, 2011 (four letters), March 9, 2011 (two letters),
March 22, 2011, March 30, 2011, March 31, 2011, April 14, 2011, April
21, 2011, May 3, 2011, May 5, 2011, May 11, 2011, June 8, 2011, June
15, 2011, June 21, 2011, June 23, 2011, July 6, 2011, July 28, 2011,
August 25, 2011, August 29, 2011, August 30, 2011, September 2, 2011,
September 9, 2011, September 12, 2011, September 15, 2011, September
26, 2011, October 10, 2011, October 24, 2011, November 14, 2011,
November 25, 2011, November 28, 2011, December 19, 2011, February 6,
2012, February 15, 2012, February 20, 2012, March 13, 2012, March 21,
2012, April 5, 2012, April 18, 2012 (two letters), April 26, 2012, May
9, 2012, and June 12, 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 July 18, 2012.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220, and 50-410,
Nine Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York
Date of application for amendments: July 20, 2011, as supplemented
on November 3, 2011, and January 12, 2012.
Brief description of amendments: The amendments revised the NMP1
Technical Specification (TS) Section 5.1, ``Site,'' and associated TS
Figure 5.1-1, ``Site Boundaries, Nine Mile Point-Unit 1,'' and the NMP2
TS Figure 4.1-1, ``Site Area and Land Portion of Exclusion Area
Boundaries,'' to reflect the transfer of a portion of the Nine Mile
Point Nuclear Station, LLC (NMPNS) site real property located outside
of the NMPNS Protected Area but within the current NMPNS Owner
Controlled Area, as well as specified easements over the remainder of
the NMPNS site, to Nine Mile Point 3 Nuclear Project, LLC (NMP3), a
subsidiary of UniStar Nuclear Energy, LLC.
Date of issuance: July 12, 2012.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 212 for Unit 1 and 142 for Unit 2.
Renewed Facility Operating License Nos. DPR-63 and NPF-69:
Amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: December 27, 2011 (76
FR 80977).
The supplements dated November 3, 2011, and January 12, 2012,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the Nuclear Regulatory Commission (NRC) staff's initial proposed
no significant hazards consideration determination noticed in
[[Page 47131]]
the Federal Register on December 27, 2011 (76 FR 80977).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 12, 2012.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 26th day of July 2012.
For the Nuclear Regulatory Commission.
Louise Lund,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2012-19004 Filed 8-6-12; 8:45 am]
BILLING CODE 7590-01-P
