Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 4381-4390 [2011-1480]
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Federal Register / Vol. 76, No. 16 / Tuesday, January 25, 2011 / Notices
Notice of Permit Applications
Received under the Antarctic
Conservation Act of 1978, Public Law
95–541.
ACTION:
The National Science
Foundation (NSF) is required to publish
notice of permit applications received to
conduct activities regulated under the
Antarctic Conservation Act of 1978.
NSF has published regulations under
the Antarctic Conservation Act at Title
45 part 670 of the Code of Federal
Regulations. This is the required notice
of permit applications received.
DATES: Interested parties are invited to
submit written data, comments, or
views with respect to this permit
application by February 24, 2011. This
application may be inspected by
interested parties at the Permit Office,
address below.
ADDRESSES: Comments should be
addressed to Permit Office, Room 755,
Office of Polar Programs, National
Science Foundation, 4201 Wilson
Boulevard, Arlington, Virginia 22230.
FOR FURTHER INFORMATION CONTACT:
Nadene G. Kennedy at the above
address or (703) 292–7405.
SUPPLEMENTARY INFORMATION: The
National Science Foundation, as
directed by the Antarctic Conservation
Act of 1978 (Pub. L. 95–541), as
amended by the Antarctic Science,
Tourism and Conservation Act of 1996,
has developed regulations for the
establishment of a permit system for
various activities in Antarctica and
designation of certain animals and
certain geographic areas as requiring
special protection. The regulations
establish such a permit system to
designate Antarctic Specially Protected
Areas.
The applications received are as
follows:
1. Applicant: R. Natalie P. Goodall,
Sarmiento 44, 9410 Ushuaia, Tierra del
Fuego, ARGENTINA.
Permit Application No. 2011–024.
SUMMARY:
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Activity for Which Permit Is Requested
Take. The applicant plans to salvage
skeletal remains of seabirds (especially
penguins) from the shorelines of South
Georgia, the South Shetlands, the
Antarctic Peninsula and adjacent
islands during visits of scientific, tourist
or supply ships, or tourist yachts. The
collected samples are very useful in the
long-term project, ‘‘Aves y Mamiferos
Marinos Australes’’ (AMMA) (study of
Southern Marine Mammals and Birds)
which have been carried out in Tierra
del Fuego since 1976. Skeletons from
Antarctic waters are especially useful in
comparison with our skeletal collections
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from southern South America. All
collected material will be cleaned,
numbered and deposited in the RNP
collection, which is housed in the
Museo Acatushun de Aves y Mamiferos
Marinos Australes at Estancia
Harberton, Tierra del Fuego
(inaugurated in 2001).
NUCLEAR REGULATORY
COMMISSION
Location
I. Background
South Georgia, the South Shetlands,
the Antarctic Peninsula and adjacent
islands.
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 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 December
30, 2010 to January 12, 2011. The last
biweekly notice was published on
January 11, 2011 (76 FR 1644).
Dates
March 1, 2011 to March 1, 2016.
1. Applicant: R. Natalie P. Goodall,
Sarmiento 44, 410 Ushuaia, Tierra del
Fuego, ARGENTINA.
Permit Application No. 2011–025.
Activity for Which Permit Is Requested
Take. The applicant plans to salvage
skeletal remains of mammals (seals,
dolphins, porpoises, or beaked whales)
from the shorelines of South Georgia,
the South Shetlands, the Antarctic
Peninsula and adjacent islands during
visits of scientific, tourist or supply
ships, or tourist yachts. The collected
samples are very useful in the long-term
project, ‘‘Aves y Mamiferos Marinos
Australes’’ (AMMA) (study of Southern
Marine Mammals and Birds) which
have been carried out in Tierra del
Fuego since 1976. Skeletons from
Antarctic waters are especially useful in
comparison with our skeletal collections
from southern South America. All
collected material will be cleaned,
numbered and deposited in the RNP
collection, which is housed in the
Museo Acatushun de Aves y Mamiferos
Marinos Australes at Estancia
Harberton, Tierra del Fuego
(inaugurated in 2001).
Location
South Georgia, the South Shetlands,
the Antarctic Peninsula and adjacent
islands.
Dates
April 1, 2011 to April 1 2016.
Nadene G. Kennedy,
Permit Officer, Office of Polar Programs.
[FR Doc. 2011–1406 Filed 1–24–11; 8:45 am]
BILLING CODE 7555–01–P
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[NRC–2011–0019]
Biweekly Notice; Applications and
Amendments to Facility Operating
Licenses Involving No Significant
Hazards Considerations
Notice of Consideration of Issuance of
Amendments to Facility Operating
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
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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.
Written comments may be submitted
by mail to the Chief, Rules,
Announcements and Directives Branch
(RADB), TWB–05–B01M, Division of
Administrative Services, Office of
Administration, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001, and should cite the publication
date and page number of this Federal
Register notice. Written comments may
also be faxed to the RADB at 301–492–
3446. Documents may be examined,
and/or copied for a fee, at the NRC’s
Public Document Room (PDR), located
at One White Flint North, Public Room
O1–F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852–2738.
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.
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 Commission’s PDR,
located at One White Flint North, Room
O1–F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852–2738.
Publicly available records will be
accessible from the Agencywide
Documents Access and Management
System’s (ADAMS) Public Electronic
Reading Room on the Internet at the
NRC Web site, 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
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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
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participate fully in the conduct of the
hearing.
If a hearing is requested, the
Commission will make a final
determination on the issue of no
significant hazards consideration. The
final determination will serve to decide
when the hearing is held. If the final
determination is that the amendment
request involves no significant hazards
consideration, the Commission may
issue the amendment and make it
immediately effective, notwithstanding
the request for a hearing. Any hearing
held would take place after issuance of
the amendment. If the final
determination is that the amendment
request involves a significant hazards
consideration, any hearing held would
take place before the issuance of any
amendment.
All documents filed in NRC
adjudicatory proceedings, including a
request for hearing, a petition for leave
to intervene, any motion or other
document filed in the proceeding prior
to the submission of a request for
hearing or petition to intervene, and
documents filed by interested
governmental entities participating
under 10 CFR 2.315(c), must be filed in
accordance with the NRC E-Filing rule
(72 FR 49139, August 28, 2007). The 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 ten
(10) days prior to the filing deadline, the
participant should contact the Office of
the Secretary by e-mail at
hearing.docket@nrc.gov, or by telephone
at 301–415–1677, to request (1) a digital
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
NRC’s public Web site at https://
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Federal Register / Vol. 76, No. 16 / Tuesday, January 25, 2011 / Notices
www.nrc.gov/site-help/e-submittals/
apply-certificates.html. System
requirements for accessing the ESubmittal server are detailed in 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 EIE, users will be
required to install a Web browser plugin from the NRC 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 NRC guidance
available on the NRC public Web site at
https://www.nrc.gov/site-help/esubmittals.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
e-mail notice confirming receipt of the
document. The E-Filing system also
distributes an e-mail notice that
provides access to the document to the
NRC 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
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contacting the NRC Meta System Help
Desk through the ‘‘Contact Us’’ link
located on the NRC Web site at https://
www.nrc.gov/site-help/esubmittals.html, by e-mail 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.
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 20852–0238, 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 NRC’s
electronic hearing docket which is
available to the public at https://
ehd.nrc.gov/EHD_Proceeding/home.asp,
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.
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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
application for amendment which is
available for public inspection at the
Commission’s PDR, located at One
White Flint North, Room O1–F21, 11555
Rockville Pike (first floor), Rockville,
Maryland 20852–2738. Publicly
available records will be accessible from
the ADAMS Public Electronic Reading
Room on the Internet at the NRC Web
site, 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 PDR Reference staff at 1–800–397–
4209, 301–415–4737, or by e-mail to
pdr.resource@nrc.gov.
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 amendment request: May 20,
2010.
Description of amendment request:
The amendments would revise the
Technical Specifications (TSs) to allow
the reactor building pressure boundary
to be opened under administrative
controls.
Basis for proposed no significant
hazards consideration determination:
As required by Title 10 of the Code of
Federal Regulations (10 CFR), 50.91(a),
the licensee has provided its analysis of
the issue of no significant hazards
consideration, which is presented
below:
Criterion 1:
Does the proposed amendment involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed changes to TS 3.6.10 and TS
3.6.16 have no effect upon accident
probabilities or consequences. The changes
proposed herein will have no impact upon
the Reactor Building or AVS [Annulus
Ventilation System] relative to the
performance of their design functions. These
structures/systems will continue to be
available and will function as designed
during and following all accidents for which
their performance is credited in the plant
safety analyses. The proposed administrative
controls for TS 3.6.16 will ensure the
restoration of the Reactor Building pressure
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boundary when required, thereby enhancing
nuclear safety. No design changes are being
made to the plant itself; therefore, there will
be no impact upon the probability of any
accident occurring. Since the performance of
these systems will not be adversely impacted,
there will be no impact upon accident
consequences.
Criterion 2:
Does the proposed amendment create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed changes to TS 3.6.10 and TS
3.6.16 do not introduce any changes or
mechanisms that create the possibility of a
new or different kind of accident. No design
changes are being made to the plant which
would result in the introduction of new
accident causal mechanisms. The proposed
changes do not introduce any new
equipment, any change to existing
equipment, or any change to the manner in
which the plant is operated. No new effects
or malfunctions will therefore be created.
Criterion 3:
Does the proposed amendment involve a
significant reduction in a margin of safety?
Response: No.
The proposed changes to TS 3.6.10 and TS
3.6.16 maintain the required design margins
of the Reactor Building and AVS for all
accidents for which their function is
assumed. All required General Design
Criteria (GDCs) contained in 10 CFR 50,
Appendix A, ‘‘General Design Criteria for
Nuclear Power Plants’’ will continue to be
satisfied following NRC approval of these
proposed changes. In addition, margin of
safety is related to the confidence in the
fission product barriers to function as
designed during and following an accident.
These barriers include the fuel cladding, the
Reactor Coolant System, and the
Containment System. The changes proposed
in this submittal have no adverse impact
upon the performance of any of these barriers
to perform their design functions during or
following an accident.
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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: Lara S. Nichols,
Associate General Counsel, Duke Energy
Corporation, 526 South Church Street—
EC07H, Charlotte, NC 28202.
NRC Branch Chief: Gloria Kulesa.
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 amendment request:
September 16, 2010.
Description of amendment request:
The amendments would revise
Technical Specification 3.3.2,
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‘‘Engineered Safety Feature Actuation
System (ESFAS) Instrumentation,’’ to
replace the references to the outdated
logic per train per doghouse with
updated references which reflect the
license amendment granted by the U.S.
Nuclear Regulatory Commission staff on
April 2, 2009.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed changes do not adversely
affect accident initiators or precursors nor
alter the design assumptions, conditions, or
configurations of the facility. The proposed
changes do not alter or prevent the ability of
structures, systems and components (SSCs)
to perform their intended function to mitigate
the consequences of an initiating event
within the assumed acceptance limits. In
review of the discussion above (Section 4.1
Significant Hazards Consideration) it can be
concluded the probability or consequences of
any accident previously evaluated are not
increased. This LAR requests administrative
changes only.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
This revision will not impact the accident
analysis. The proposed changes will not alter
the requirements of the ESFAS or its function
during accident conditions. No new or
different accidents result from the changes
proposed. The changes do not involve a
physical alteration of the plant (i.e., no new
or different type of equipment will be
installed) or any changes in methods
governing normal plant operation. The
changes do not alter assumptions made in the
safety analysis. The proposed changes are
consistent with the safety analyses
assumptions. In review of the discussion
above (Section 4.1 Significant Hazards
Consideration) it can be concluded that these
changes do not create the possibility of a new
or different kind of accident from any
accident previously evaluated. This LAR
requests administrative changes only.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
The proposed changes do not alter the
manner in which safety limits, limiting safety
system settings or limiting conditions for
operation are determined. The safety analysis
acceptance criteria are not affected by these
changes. The proposed changes will not
result in plant operation in a configuration
outside the design basis. The proposed
changes do not adversely affect systems that
respond to safely shutdown the plant and to
maintain the plant in a safe shutdown
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condition. In review of the discussion above
(Section 4.1 Significant Hazards
Consideration) it can be concluded that the
proposed changes do not involve a
significant reduction in the margin of safety.
This LAR requests administrative changes
only.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Lara S. Nichols,
Associate General Counsel, Duke Energy
Corporation, 526 South Church Street—
EC07H, Charlotte, NC 28202.
NRC Branch Chief: Gloria Kulesa.
Duke Energy Carolinas, LLC, et al.,
Docket Nos. 50–369, 50–370, McGuire
Nuclear Station, Units 1 and 2,
Mecklenburg County, North Carolina;
50–413 and 50–414, Catawba Nuclear
Station, Units 1 and 2, York County,
South Carolina
Date of amendment request: June 29,
2010.
Description of amendment request:
The amendments would revise
Technical Specification (TS) 3.3.1,
‘‘Reactor Trip System (RTS)
Instrumentation’’ and TS 3.3.2,
‘‘Engineered Safety Feature Actuation
System (ESFAS) Instrumentation.’’
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 specific Technical Specification
changes are associated with (1) the specific
Allowable Values for various RTS and
ESFAS channels, including instrumentation
associated with neutron flux, containment
pressure, pressurizer pressure, pressurizer
water level, reactor coolant flow, reactor
coolant pump underfrequency, steam
generator water level, turbine impulse
pressure, steam line pressure, and reactor
coolant temperature; (2) the addition of
specific requirements to be taken if an
instrument channel setpoint is outside its
predefined as-found tolerance; and (3) the
addition of specific requirements regarding
resetting of an instrument channel setpoint
within an as-left tolerance.
The RTS and ESFAS instrumentation is
accident mitigation equipment and does not
affect the probability of any accident being
initiated. In addition, none of the
abovementioned proposed Technical
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Specification changes affect the probability of
any accident being initiated.
The proposed changes to TS Allowable
Values are based on methodology that is
consistent with the intent of ISA [Instrument
Society of America] Standard RP67.04–1994,
Part II, ‘‘Methodologies for the Determination
of Setpoints for Nuclear Safety Related
Instrumentation,’’ and will preserve
assumptions in the applicable accident
analyses. None of the proposed changes alter
any assumption previously made in the
radiological consequences evaluations, nor
do they affect mitigation of the radiological
consequences of an accident previously
evaluated.
In summary, the proposed changes will 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.
No new accident scenarios, failure
mechanisms, or single failures are introduced
as a result of any of the proposed changes.
The RTS and ESFAS are not capable by itself
of initiating any accident. No physical
changes to the overall plant are being
proposed. No changes to the overall manner
in which the plant is operated are being
proposed. The proposed changes do not
introduce any new failure modes.
Therefore, none of the proposed changes
will create the possibility of a new or
different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
Margin of safety is related to the
confidence in the ability of the fission
product barriers to perform their intended
functions. These barriers include the fuel
cladding, the reactor coolant system pressure
boundary, and the containment barriers. The
proposed changes will not have any impact
on these barriers. Plant actuation features and
Nominal Trip Setpoints will be unchanged
and will actuate prior to exceeding any
analytical limits. No accident mitigating
equipment will be adversely impacted.
Therefore, existing safety margins will be
preserved. None of the proposed changes will
involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Lara S. Nichols,
Associate General Counsel, Duke Energy
Corporation, 526 South Church Street—
EC07H, Charlotte, NC 28202.
NRC Branch Chief: Gloria Kulesa.
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Exelon Generation Company, LLC,
Docket No. 50–461, Clinton Power
Station, Unit No. 1, DeWitt County,
Illinois
Date of amendment request: October
28, 2010.
Description of amendment request:
The proposed amendment would
modify Clinton Power Station Technical
Specifications (TS) Section 3.8.1, ‘‘AC
Sources Operating,’’ by revising certain
Surveillance Requirements (SR) related
to the Division 3 alternating current
(AC) Sources. The Division 3 AC
Sources are independent sources of
offsite and onsite AC power primarily
dedicated to the High-Pressure Core
Spray (HPCS) system. The TS currently
prohibit performing the testing required
by SR 3.8.1.8 and SR 3.8.1.12 in Modes
1 or 2, and prohibit performing the
testing required by SR 3.8.1.11, SR
3.8.1.16, and SR 3.8.1.19 in Modes 1, 2,
or 3. The proposed amendment would
remove these Mode restrictions and
allow all five of the identified SRs to be
performed in any operating Mode for
the Division 3 AC Sources.
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:
EGC has evaluated whether or not a
significant hazards consideration is
involved with the proposed amendment
by focusing on the three standards set
forth in 10 CFR 50.92, ‘‘Issuance of
Amendment,’’ as discussed below:
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The Division 3 (i.e., HPCS) diesel generator
(DG) and its associated emergency loads are
accident mitigating features, not accident
initiators. Therefore, the proposed TS
changes to allow the performance of certain
Division 3 AC Sources surveillance testing in
any plant operating Mode will not
significantly impact the probability of any
previously evaluated accident.
The design of plant equipment is not being
modified by the proposed changes. As such,
the ability of the Division 3 AC Sources to
respond to a design basis accident will not
be adversely impacted by the proposed
changes. Testing procedures include steps to
ensure that injection into the reactor vessel
is precluded. The proposed changes to the TS
surveillance testing requirements for the
Division 3 AC Sources do not affect the
operability requirements for the AC Sources,
as verification of such operability will
continue to be performed as required.
Continued verification of operability
supports the capability of the Division 3 AC
Sources to perform their required functions
of providing emergency power to HPCS
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4385
system equipment, consistent with the plant
safety analyses. Limiting testing to only one
AC Source at a time ensures that design basis
requirements are met. Should a fault occur
while testing the Division 3 AC Sources,
there would be no significant impact on any
accident consequences since the other two
divisional AC Sources and associated
emergency loads would be available to
provide the minimum safety functions
necessary to shut down the unit and
maintain it in a safe shutdown condition.
Therefore, the proposed amendment does
not involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
No changes are being made to the plant
that would introduce any new accident
causal mechanisms. Equipment will be
operated in the same configuration with the
exception of the plant operating mode in
which the Division 3 AC Sources
surveillance testing is conducted.
Performance of these surveillances tests
while online will continue to verify
operability of the Division 3 AC Sources. The
proposed amendment does not impact any
plant systems that are accident initiators and
does not adversely impact any accident
mitigating systems.
Therefore, the proposed amendment does
not create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve
a significant reduction in a margin of safety?
Response: No.
Margin of safety is related to confidence in
the ability of the fission product barriers (fuel
cladding, reactor coolant system, and
primary containment) to perform their design
functions during and following postulated
accidents. The proposed changes to the TS
surveillance testing requirements for the
Division 3 AC Sources do not affect the
operability requirements for the AC Sources,
as verification of such operability will
continue to be performed as required.
Continued verification of operability
supports the capability of the Division 3 AC
Sources to perform their required function of
providing emergency power to HPCS system
equipment, consistent with the plant safety
analyses. Consequently, the performance of
the fission product barriers will not be
adversely impacted by implementation of the
proposed amendment. In addition, the
proposed changes do not alter setpoints or
limits established or assumed by the accident
analysis. Further, performing Division 3 AC
Sources surveillance activities online
increases the Division 3 DG and HPCS
system availability during refueling outages
and allows the testing of the Division 3
systems to be conducted when both Division
1 and 2 systems are required to be
OPERABLE.
Therefore, the proposed amendment 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: Mr. Bradley J.
Fewell, Associate General Counsel,
Exelon Nuclear, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Robert D. Carlson.
Florida Power and Light Company
(FPL), Docket Nos. 50–250 and 50–251,
Turkey Point Plant, Units 3 and 4,
Miami-Dade County, Florida
Date of amendment request: July 16,
2010.
Description of amendment request:
The amendments would revise the
Technical Specifications (TSs) to adopt
Nuclear Regulatory Commission (NRC)approved Revision 3 to Technical
Specification Task Force (TSTF)
Improved Standard Technical
Specification Change Traveler, TSTF–
448, ‘‘Control Room Envelope
Habitability.’’ The proposed
amendments include changes to the TS
requirements related to control room
envelope (CRE) habitability in TS
3/4.7.5, ‘‘Control Room Emergency
Ventilation System (CREVS),’’ and TS
Section 6.8, ‘‘Administrative Controls—
Procedures and Programs.’’ This
submittal satisfies the commitment
identified in FPL’s letter dated August
10, 2007, to adopt the applicable
portions of TSTF–448. Additionally,
this application updates the original
submittal of license amendment request
194 dated September 26, 2008, in
response to an NRC request for
additional information to remove any
reference of unapproved TSTF–508,
which has been done.
The NRC staff published a notice of
opportunity for comment in the Federal
Register on October 17, 2006 (71 FR
61075), on possible amendments
adopting TSTF–448, including a model
safety evaluation and model no
significant hazards consideration
(NSHC) determination, using the
consolidated line-item improvement
process. The NRC staff subsequently
issued a notice of availability of the
models for referencing in license
amendment applications in the Federal
Register on January 17, 2007 (72 FR
2022). The licensee affirmed the
applicability of the following NSHC
determination in its application dated
July 16, 2010.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), an
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analysis of the issue of no significant
hazards consideration is presented
below:
Criterion 1: The Proposed Change Does Not
Involve a Significant Increase in the
Probability or Consequences of an Accident
Previously Evaluated.
The proposed change does not adversely
affect accident initiators or precursors nor
alter the design assumptions, conditions, or
configuration of the facility. The proposed
change does not alter or prevent the ability
of structures, systems, and components
(SSCs) to perform their intended function to
mitigate the consequences of an initiating
event within the assumed acceptance limits.
The proposed change revises the TS for the
CRE emergency ventilation system, which is
a mitigation system designed to minimize
unfiltered air leakage into the CRE and to
filter the CRE atmosphere to protect the CRE
occupants in the event of accidents
previously analyzed. An important part of
the CRE emergency ventilation system is the
CRE boundary. The CRE emergency
ventilation system is not an initiator or
precursor to any accident previously
evaluated. Therefore, the probability of any
accident previously evaluated is not
increased. Performing tests to verify the
operability of the CRE boundary and
implementing a program to assess and
maintain CRE habitability ensure that the
CRE emergency ventilation system is capable
of adequately mitigating radiological
consequences to CRE occupants during
accident conditions, and that the CRE
emergency ventilation system will perform as
assumed in the consequence analyses of
design basis accidents. Thus, the
consequences of any accident previously
evaluated are not increased.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
Criterion 2: The Proposed Change Does Not
Create the Possibility of a New or Different
Kind of Accident from any Accident
Previously Evaluated.
The proposed change does not impact the
accident analysis. The proposed change does
not alter the required mitigation capability of
the CRE emergency ventilation system, or its
functioning during accident conditions as
assumed in the licensing basis analyses of
design basis accident radiological
consequences to CRE occupants. No new or
different accidents result from performing the
new surveillance or following the new
program. The proposed change does not
involve a physical alteration of the plant (i.e.,
no new or different type of equipment will
be installed) or a significant change in the
methods governing normal plant operation.
The proposed change does not alter any
safety analysis assumptions and is consistent
with current plant operating practice.
Therefore, this change does not create the
possibility of a new or different kind of
accident from any accident previously
evaluated.
Criterion 3: The Proposed Change Does Not
Involve a Significant Reduction in the Margin
of Safety.
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The proposed change does not alter the
manner in which safety limits, limiting safety
system settings or limiting conditions for
operation are determined. The proposed
change does not affect safety analysis
acceptance criteria. The proposed change
will not result in plant operation in a
configuration outside the design basis for an
unacceptable period of time without
compensatory measures. The proposed
change does not adversely affect systems that
respond to safely shut down the plant and to
maintain the plant in a safe shutdown
condition.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
The NRC staff proposes to determine
that the amendment request involves no
significant hazards consideration.
Attorney for licensee: M.S. Ross,
Attorney, Florida Power & Light, P.O.
Box 14000, Juno Beach, Florida 33408–
0420.
NRC Branch Chief: Douglas A.
Broaddus.
Nebraska Public Power District, Docket
No. 50–298, Cooper Nuclear Station,
Nemaha County, Nebraska
Date of amendment request: October
29, 2010.
Description of amendment request:
The proposed amendment would revise
Technical Specification (TS) 3.8.4, ‘‘DC
[Direct Current] Sources—Operating,’’
and TS 3.8.6, ‘‘Battery Cell Parameters.’’
Specifically, the proposed changes
would replace non-conservative
minimum voltages in Surveillance
Requirement 3.8.4.1 for the 125 volt
direct current (V DC) and 250 V DC
essential batteries, and the nonconservative battery specific gravity
values listed in TS Table 3.8.6–1,
‘‘Battery Cell Parameter Requirements.’’
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.
Performing surveillances that verify
terminal voltage and specific gravity of
batteries is not a precursor of any accident
previously evaluated. Restoring battery limits
to conservative values does not significantly
affect the method of performing the
surveillances, such that the probability of an
accident would be affected. Therefore, the
proposed changes do not result in a
significant increase in the probability of an
accident previously evaluated.
Restoring battery limits to conservative
values so that batteries are maintained in
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accordance with plant design basis ensures
they provide the power assumed in design
basis accident mitigation calculations.
Therefore, the change does not involve a
significant increase in the consequences of an
accident previously evaluated.
NPPD [Nebraska Public Power District]
concludes that the proposed changes do 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.
The proposed change does not involve any
modification to the plant or equipment or
how they are operated. Therefore, NPPD
concludes that these proposed changes do
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 proposed change will continue to
ensure station batteries are able to perform
their design function as assumed in
calculations that evaluate their function
during design basis accidents. The proposed
change actually increases the margin of safety
by restoring conservatisms inherent in
battery design and manufacturer’s
recommendations. Based on this, the ability
of CNS [Cooper Nuclear Station] to mitigate
the design basis accidents that rely on
operation of the station batteries is not
adversely impacted. Therefore, NPPD
concludes that these proposed changes do
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: 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
No. 1, Washington County, Nebraska
Date of amendment request: July 12,
2010.
Description of amendment request:
This notice is being reissued in its
entirety due to missing statements from
the description of the amendment
request in the notice published in the
Federal Register on December 28, 2010
(75 FR 81671). The proposed
amendment would modify Item 1 of
Table 2–5, ‘‘Instrumentation Operating
Requirements for Other Safety Feature
Functions,’’ of Technical Specification
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(TS) 2.15, ‘‘Instrumentation and Control
Systems,’’ to provide new Note (e), and
Surveillance Requirement (SR) Items 1
and 2 of Table 3–3, ‘‘Minimum
Frequencies for Checks, Calibrations
and Testing of Miscellaneous
Instrumentation and Controls,’’ of TS
3.1, ‘‘Instrumentation and Control,’’
which pertain to operability of the
primary and secondary control element
assembly (CEA) position indication
system (CEAPIS) channels. A new SR is
proposed for Item 4 of Table 3–3 of TS
3.1, which will verify the position of
CEAs each shift. The proposed
amendment will ensure that CEA
alignment is maintained during power
operations so that the power
distribution and reactivity limits
defined by the design power peaking
and shutdown margin (SDM) limits are
preserved. The proposed amendment
would also revise TS 2.10.2(7)c
regarding actions to be taken when the
regulating CEA groups are inserted
below the Long Term Insertion Limit.
The TS would be revised to require
actions to be taken when either time
interval is exceeded, which would also
make TS 2.10.2(7)c more consistent
with Combustion Engineering (CE)
Standard Technical Specifications
(STS).
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed amendment involve
a significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed amendment will allow plant
operation to continue when a CEAPIS
channel is inoperable by requiring prompt
verification of CEA positions following CEA
movement. CEAs are most likely to become
misaligned during movement and therefore,
this change will cause CEA alignment errors
to be promptly detected and corrected. It is
appropriate to clarify that CEAPIS channels
are not subject to the requirements of TS
2.15(1), (2), and (3) as they are not designed
to be placed in trip or bypass, nor are they
engineered safety feature (ESF) or isolation
logic subsystems.
The proposed amendment does not alter
the requirements of TS 2.15(4) regarding the
rod block function of the secondary CEAPIS
channel. Should the secondary CEAPIS
channel or its rod block function be
inoperable, several additional CEA deviation
events are possible. However, this situation
is already addressed by TS 2.15(4), which
requires the CEAs (rods) to be maintained
fully withdrawn with the control rod drive
system mode switch in the off position
except when manual motion of CEA Group
4 is required to control axial power
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4387
distribution. This is the same position that
the CEAs must be in (fully withdrawn) when
the plant is at power (Mode 1) in order to
utilize distributed control system (DCS) core
mimic to CHANNEL CHECK the CEAPIS
channels.
If it was not possible to use DCS core
mimic to verify the primary CEAPIS channel
as would be the case if CEA Group 4 was
inserted to control axial power distribution,
then the primary CEAPIS channel would be
declared inoperable when the CHANNEL
CHECK could not be accomplished. The
plant would then be placed in hot shutdown
(Mode 3) within 12 hours in accordance with
TS 2.15(4). Therefore, although the proposed
amendment will allow a CEAPIS channel to
be inoperable indefinitely, there is no
significant increase in the probability or
consequences of an accident as the
requirements of TS 2.15(4) will continue to
be met. This serves to prevent the type of
CEA deviation events that the rod block
function was designed for.
Replacing the current method of verifying
CEAPIS data with the defined term
CHANNEL CHECK is an improvement that
provides additional flexibility without
weakening the intent of the surveillance. As
a result, when it is feasible to obtain CEA
position indication from DCS core mimic
(i.e., when the CEAs are either fully inserted
or fully withdrawn), the primary and
secondary CEAPIS channels will be
compared with DCS core mimic indication as
well as each other.
As an additional means of verifying CEA
positions, DCS core mimic indication
provides added confidence that the CEAs are
in the indicated positions. Should the
primary or secondary CEAPIS channel
become inoperable, the accuracy and
reliability of DCS core mimic indication is
assured by its previous comparison with both
OPERABLE channels. Comparison of the
OPERABLE CEAPIS channel with DCS core
mimic will satisfy the required CHANNEL
CHECK and allow continued operation while
the inoperable channel is repaired. The
proposed amendment ensures that the CEA
alignment required by TS 2.10.2(4) is met
each shift by requiring all full length
(shutdown and regulating) CEAs to be
positioned within 12 inches of all other CEAs
in the group.
The change proposed for TS 2.10.2(7)c
incorporates more conservative wording to
ensure that the regulating CEA groups are
maintained within the Long Term Insertion
Limit. The proposed change will ensure that
corrective actions are taken if either time
interval is exceeded and makes TS 2.10.2(7)c
more consistent with CE STS.
The proposed amendment does not alter
the plant configuration, require new plant
equipment to be installed, alter accident
analysis assumptions, add any initiators, or
affect the function of plant systems or the
manner in which systems are operated,
maintained, modified, tested, or inspected.
As an additional means of verifying
primary and secondary CEAPIS data, DCS
core mimic indication increases confidence
in the reliability of CEAPIS data.
The proposed amendment will help
minimize unplanned shutdowns that can
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cause plant transients yet continues to ensure
that power distribution and reactivity limits
are maintained. Therefore, it is concluded
that this change does not involve a
significant increase in the probability or
consequences of an accident previously
evaluated.
2. Does the proposed amendment create
the possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not change
the design function or operation of the
primary or secondary CEAPIS channels. If
one CEAPIS channel should become
inoperable, the position of CEAs will be
verified within 15 minutes of any CEA
movement to quickly detect and correct CEA
alignment errors. Data from each CEAPIS
channel will continue to be compared to the
other channel each shift as before. However,
a CHANNEL CHECK will require that
CEAPIS channel data also be compared with
DCS core mimic indication when it is
available. Thus, when the CEAPIS channels
are required to be OPERABLE, there will be
at least two means of verifying the position
of CEAs or else appropriate actions must be
taken. The CEA alignment required by TS
2.10.2(4) is assured by requiring verification
each shift that all full length (shutdown and
regulating) CEAs are positioned within 12
inches of all other CEAs in the group.
No changes are proposed to testing and
calibration of the CEAPIS channels and these
requirements will continue to ensure that
they are capable of performing their design
function. Use of the defined term CHANNEL
CHECK is an appropriate surveillance
method as it requires that the channel be
compared with other independent channels
measuring the same variable where feasible.
DCS core mimic is a diverse, accurate and
reliable means of verifying CEA positions
when the CEAs are fully inserted or fully
withdrawn. The change proposed for TS
2.10.2(7)c ensures that appropriate corrective
actions are taken when the regulating CEA
groups are below the Long Term Insertion
Limit in excess of either of the specified time
intervals.
No new structures, systems, or components
(SSCs) are being installed, and no credible
new failure mechanisms, malfunctions, or
accident initiators are created. Therefore, the
proposed amendment 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.
When a CEAPIS channel is inoperable, the
proposed amendment allows plant operation
to continue but requires more frequent
verification of CEA positions following any
CEA movement, which is when CEAs are
most likely to become misaligned. This will
enable CEA alignment errors to be detected
and corrected more promptly. As CEAPIS
channels are not designed to be placed in trip
or bypass, nor are they engineered safety
feature (ESF) or isolation logic subsystems, it
is appropriate to clarify that TS 2.15(1), (2),
and (3) do not apply. FCS normally operates
with the CEAs fully withdrawn and
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maintains reactivity control by adjusting
reactor coolant system (RCS) boric acid
concentration. When the CEAs are fully
withdrawn (or fully inserted), DCS core
mimic indication provides accurate and
reliable indication of CEA positions suitable
for comparison with the primary and
secondary CEAPIS channels. Thus, even with
one CEAPIS channel inoperable, a diverse
means of verifying the accuracy of the
OPERABLE CEAPIS channel will be
available. The accuracy and reliability of DCS
core mimic is assured by testing conducted
each refueling outage with continued
assurance provided by comparison with
primary and secondary CEAPIS each shift.
The change also ensures that the CEA
alignment required by TS 2.10.2(4) is met
each shift by requiring all full length
(shutdown and regulating) CEAs to be
positioned within 12 inches of all other CEAs
in the group. The proposed amendment does
not alter the TS 2.15(4) requirement to place
the reactor in hot shutdown in the event that
both CEAPIS channels are inoperable. The
change proposed for TS 2.10.2(7)c
incorporates more conservative wording to
ensure that the regulating CEA groups are
maintained within the Long Term Insertion
Limit.
The proposed amendment will help
minimize unplanned shutdowns that can
cause plant transients yet continues to ensure
that power distribution and reactivity limits
are maintained. The proposed amendment
does not alter the plant configuration, require
new plant equipment to be installed, alter
accident analysis assumptions, add any
initiators, or affect the function of plant
systems or the manner in which systems are
operated, maintained, modified, tested, or
inspected. Therefore, the proposed
amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: David A. Repka,
Esq., Winston & Strawn, 1700 K Street,
NW., Washington, DC 20006–3817.
NRC Branch Chief: Michael T.
Markley.
Omaha Public Power District, Docket
No. 50–285, Fort Calhoun Station, Unit
No. 1, Washington County, Nebraska
Date of amendment request: August
16, 2010, as supplemented by letter
dated September 27, 2010.
Description of amendment request:
The proposed amendment would
remove the Technical Specification (TS)
limiting condition for operation (LCO)
2.15, ‘‘Instrumentation and Control
Systems,’’ Table 2–5, ‘‘Instrumentation
Operating Requirements for Other
Safety Feature Functions,’’ Items 3, 4,
and 5, the associated Notes a, b, c, and
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d, and the associated footnote, for
power-operated relief valve (PORV) and
pressurizer safety valve (PSV) acoustic
position indication and tail pipe
temperature from the Fort Calhoun
Station (FCS) TS. The proposed
amendment would also revise the
surveillance requirement (SR), TS 3.1,
‘‘Instrumentation and Control,’’ Table 3–
3, ‘‘Minimum Frequencies for Checks,
Calibrations and Testing of
Miscellaneous Instrumentation and
Controls,’’ Items 21, 23, and 24 for PORV
Operation and Acoustic Position
Indication, Safety Valve Acoustic
Position Indication, and PORV/Safety
Valve Tail Pipe Temperature,
respectively. Specifically, Table 3–3,
Item 21 will be revised to reflect the
performance of the PORV operation
channel functional test on its existing
refueling frequency and deletes the
monthly frequency denoted in the TS
for the acoustic position indication
which would also be more aligned with
NUREG–1432, ‘‘Standard Technical
Specifications, Combustion Engineering
Plants,’’ Revision 3, for PORV operation;
and Items 21, 23, and 24 will be revised
to relocate the acoustic position
indication and tail pipe temperature
indication SRs from the FCS TS. In
conjunction with the proposed TS
changes, operability and surveillance
requirements for the acoustic position
indication and tail pipe temperature
indication instrumentation would be
incorporated into the FCS Updated
Safety Analysis Report (USAR) and
associated plant procedures.
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 safety valve acoustic position
indication does not affect the operation of its
associated spring-loaded safety valve. As
such, the proposed change does not increase
the probability of an accident. The acoustic
monitor and tail pipe temperature indication
are only two of the indications used to
identify that a safety valve is open. Other
indications are available to the operators and
alarm in the control room. The acoustic
monitor is only one of the indications that
the abnormal and emergency procedures
direct operators to use to diagnose the
opening of a safety valve. The failure of the
power operated relief valve (PORV)/safety
valve position instrumentation is not
assumed to be an initiator of any analyzed
event in the Updated Safety Analysis Report
(USAR). The proposed changes do not alter
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the physical design of the PORVs/safety
valves or any other plant structure, system or
component (SSC). The changes would
remove the PORV/safety valve position
indicator operability and surveillance
requirements from the Fort Calhoun Station
(FCS) Technical Specifications (TS), and
incorporate the requirements for this
instrumentation into a licensee-controlled
document under the control of 10 CFR 50.59.
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 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 instrumentation is not needed for
manual operator actions necessary for safety
systems to accomplish their safety function
for the design basis accident events. The
acoustic position indicator and tail-pipe
temperature instrumentation provides only
alarm and PORV/safety valve position
indication, and does not provide an input to
any automatic trip function. Diverse means
are available to monitor PORV/safety valve
position, and operability and surveillance
requirements will be established in a
licensee-controlled document to ensure the
reliability of the PORV/safety valve position
monitoring capability. Changes to these
requirements will be subject to the controls
of 10 CFR 50.59, providing the appropriate
level of regulatory control. In addition, the
PORV operation is currently tested on a
refueling frequency, which is aligned with
the surveillance 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.
VerDate Mar<15>2010
18:40 Jan 24, 2011
Jkt 223001
Attorney for licensee: David A. Repka,
Esq., Winston & Strawn, 1700 K Street,
NW., Washington, DC 20006–3817.
NRC Branch Chief: Michael T.
Markley.
Previously Published Notices of
Consideration of Issuance of
Amendments to Facility Operating
Licenses, Proposed No Significant
Hazards Consideration Determination,
and Opportunity for a Hearing
The following notices were previously
published as separate individual
notices. The notice content was the
same as above. They were published as
individual notices either because time
did not allow the Commission to wait
for this biweekly notice or because the
action involved exigent circumstances.
They are repeated here because the
biweekly notice lists all amendments
issued or proposed to be issued
involving no significant hazards
consideration.
For details, see the individual notice
in the Federal Register on the day and
page cited. This notice does not extend
the notice period of the original notice.
Entergy Nuclear Operations, Inc.,
Docket No. 50–255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of amendment request: January
27, 2010.
Brief description of amendment
request: The amendment revises Section
2.E. of the Palisades Nuclear Plant (PNP)
Renewed Facility Operating License to
remove the name of the former operator
of the plant in the title of the PNP
physical security plan and replace it
with Entergy Nuclear. The change also
removes the security plan revision
number and the date the plan was
submitted to the Nuclear Regulatory
Commission.
Date of publication of individual
notice in Federal Register: November
18, 2010 (75 FR 70708).
Expiration date of individual: January
17, 2011
Notice of Issuance of Amendments to
Facility Operating 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
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4389
10 CFR chapter I, which are set forth in
the license amendment.
Notice of Consideration of Issuance of
Amendment to Facility Operating
License, 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, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville,
Maryland. Publicly available records
will be accessible from the Agencywide
Documents Access and Management
System (ADAMS) Public Electronic
Reading Room on the internet at the
NRC Web site, 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
Reference staff at 1 (800) 397–4209,
(301) 415–4737 or by e-mail to
pdr.resource@nrc.gov.
Duke Energy Carolinas, LLC, Docket
Nos. 50–269, 50–270, and 50–287,
Oconee Nuclear Station, Units 1, 2, and
3, Oconee County, South Carolina
Date of application of amendments:
May 30, 2008, as supplemented by
letters dated October 31, 2008, January
30, 2009, February 9, 2009, February 23,
2009, May 31, 2009, August 3, 2009,
September 29, 2009, and November 30,
2009. By letter dated April 14, 2010, the
licensee resubmitted the application
and superseded the contents of the
application submitted by letter dated
May 30, 2008, as supplemented October
31, 2008. This resubmitted application,
however, does not supersede the
supplements dated January 30, 2009,
February 9, 2009, February 23, 2009,
May 31, 2009, August 3, 2009,
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September 29, 2009, and November 30,
2009. By letters dated September 13,
2010, September 27, 2010, October 14,
2010, November 19, 2010, and
December 22, 2010, the licensee
supplemented the April 14, 2010
application.
Brief description of amendments: The
amendments revised the licenses and
Technical Specifications to allow the
licensee to maintain a fire protection
program in accordance with 10 CFR
50.48(c) for the Oconee Nuclear Station,
Units 1, 2, and 3.
Date of Issuance: December 29, 2010.
Effective date: As of the date of
issuance and shall be fully implemented
prior to January 1, 2013.
Amendment Nos.: Unit 1—371, Unit
2—373, Unit 3—372.
Renewed Facility Operating License
Nos. DPR–38, DPR–47, and DPR–55:
Amendments revised the licenses and
the Technical Specifications.
Date of initial notice in Federal
Register: October 28, 2010 (75 FR
66395).
The supplements dated September 13,
2010, September 27, 2010, October 14,
2010, November 19, 2010, and
December 22, 2010, provided additional
information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the staff’s original
proposed no significant hazards
consideration determination.
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated December 29,
2010.
No significant hazards consideration
comments received: No.
Entergy Gulf States Louisiana, LLC, and
Entergy Operations, Inc., Docket No.
50–458, River Bend Station, Unit 1,
West Feliciana Parish, Louisiana
Date of amendment request: July 22,
2010.
Brief description of amendment: The
amendment revised Limiting Condition
for Operation (LCO) 3.10.1, ‘‘Inservice
Leak and Hydrostatic Testing
Operation,’’ and the associated Bases, to
expand its scope to include provisions
for temperature excursions greater than
200 degrees Fahrenheit as a
consequence of inservice leak and
hydrostatic testing, and as a
consequence of scram time testing
initiated in conjunction with an
inservice leak or hydrostatic test, while
considering operational conditions to be
in Mode 4. The change is consistent
with NRC-approved Technical
Specification Task Force (TSTF)
Improved Standard Technical
Specifications Change Traveler, TSTF–
VerDate Mar<15>2010
18:40 Jan 24, 2011
Jkt 223001
484, ‘‘Use of TS 3.10.1 for Scram Time
Testing Activities,’’ that was announced
in the Federal Register on October 27,
2006 (71 FR 63050), as part of the
Consolidated Line Item Improvement
Process (CLIIP).
Date of issuance: January 5, 2011.
Effective date: As of the date of
issuance and shall be implemented 60
days from the date of issuance.
Amendment No.: 170.
Facility Operating License No. NPF–
47: The amendment revised the Facility
Operating License and Technical
Specifications.
Date of initial notice in Federal
Register: October 5, 2010 (75 FR
61524).
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated January 5, 2011.
No significant hazards consideration
comments received: No.
Exelon Generation Company, LLC,
Docket No. 50–289, Three Mile Island
Nuclear Station, Unit 1 (TMI–1),
Dauphin County, Pennsylvania
Date of application for amendment:
March 24, 2010, supplemented by
letters dated July 29, 2010, and
September 27, 2010.
Brief description of amendment: The
changes revise the TMI–1 technical
specifications to relocate certain
surveillance frequencies to a licenseecontrolled program through the
implementation of Nuclear Energy
Institute 04–10, ‘‘Risk-Informed
Technical Specifications Initiative 5b,
Risk-Informed Method for Control of
Surveillance Frequencies.’’ The changes
are consistent with U.S. Nuclear
Regulatory Commission (NRC)-approved
Technical Specifications Task Force
(TSTF) Standard Technical
Specifications change TSTF–425,
‘‘Relocate Surveillance Frequencies to
Licensee Control—Risk Informed
Technical Specifications Task Force
Initiative 5b,’’ Revision 3.
Date of issuance: January 12, 2011.
Effective date: Immediately, and shall
be implemented within 120 days.
Amendment No.: 274.
Facility Operating License No. DPR–
50. Amendment revised the license and
the technical specifications.
Date of initial notice in Federal
Register: May 18, 2010 (75 FR 27829).
The supplements dated July 29, 2010,
and September 27, 2010, provided
additional information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the NRC staff’s
original proposed no significant hazards
determination. The Commission’s
related evaluation of the amendment is
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contained in a Safety Evaluation dated
January 12, 2011.
No significant hazards consideration
comments received: No.
Dated at Rockville, Maryland, this 13th day
of January 2011.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor
Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. 2011–1480 Filed 1–24–11; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[NRC–2009–0263]
Draft Regulatory Guide: Comment
Period Extension and Correction
Nuclear Regulatory
Commission.
ACTION: Notice of Reissuance and
Availability of Draft Regulatory Guide
(DG)–1229; Comment Period Extension
and Correction.
AGENCY:
FOR FURTHER INFORMATION CONTACT:
Aaron Szabo, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001, telephone: 301–415–1985 or
e-mail: Aaron.Szabo@nrc.gov.
SUMMARY: On January 13, 2011, the U.
S. Nuclear Regulatory Commission
(NRC) published a notice in the Federal
Register (76 FR 2425) announcing the
reissuance and availability of Draft
Regulatory Guide (DG)–1229, titled
‘‘Assuring the Availability of Funds for
Decommissioning Nuclear Reactors.’’
This Federal Register notice stated that
electronic copies of DG–1229 were
available in the NRC’s Agencywide
Documents Access and Management
System (ADAMS) (https://www.nrc.gov/
reading-rm/adams.html), under
Accession No. ML103350136 and that
the regulatory analysis was available
under ML103350166. The ADAMS
accession numbers assigned to DG–1229
and noted in 76 FR 2425 are incorrect.
Due to this error, the comment period
has been extended to allow the public
access the correct version.
SUPPLEMENTARY INFORMATION: The NRC
issued a notice of reissuance and
availability of DG–1229, ‘‘Assuring the
Availability of Funds for
Decommissioning Nuclear Reactors’’ on
January 13, 2011. The ADAMS
accession numbers for the regulatory
analysis and the draft regulatory guide
noted on page 2426 of volume 76,
‘‘further information’’ section were
incorrect. The content should read ‘‘The
regulatory analysis is available
E:\FR\FM\25JAN1.SGM
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Agencies
[Federal Register Volume 76, Number 16 (Tuesday, January 25, 2011)]
[Notices]
[Pages 4381-4390]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2011-1480]
=======================================================================
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2011-0019]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. 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
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 December 30, 2010 to January 12, 2011. The
last biweekly notice was published on January 11, 2011 (76 FR 1644).
Notice of Consideration of Issuance of Amendments to Facility Operating
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
[[Page 4382]]
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.
Written comments may be submitted by mail to the Chief, Rules,
Announcements and Directives Branch (RADB), TWB-05-B01M, Division of
Administrative Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be faxed to the RADB at 301-492-3446.
Documents may be examined, and/or copied for a fee, at the NRC's Public
Document Room (PDR), located at One White Flint North, Public Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852-
2738.
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.
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 Commission's PDR, located at One White Flint North, Room O1-F21,
11555 Rockville Pike (first floor), Rockville, Maryland 20852-2738.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, 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 held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, any hearing held
would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule (72 FR 49139,
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
ten (10) days prior to the filing deadline, the participant should
contact the Office of the Secretary by e-mail at
hearing.docket@nrc.gov, or by telephone at 301-415-1677, to request (1)
a digital 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 NRC's public Web site at https://
[[Page 4383]]
www.nrc.gov/site-help/e-submittals/apply-certificates.html. System
requirements for accessing the E-Submittal server are detailed in 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 EIE, users will be required to install a Web
browser plug-in from the NRC 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 NRC guidance
available on the NRC 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 e-mail notice confirming receipt of the document. The
E-Filing system also distributes an e-mail notice that provides access
to the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at https://www.nrc.gov/site-help/e-submittals.html, by e-mail 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
20852-0238, 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
NRC's electronic hearing docket which is available to the public at
https://ehd.nrc.gov/EHD_Proceeding/home.asp, 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 application for amendment which is available for
public inspection at the Commission's PDR, located at One White Flint
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852-2738. Publicly available records will be accessible from
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web
site, 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 PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov.
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 amendment request: May 20, 2010.
Description of amendment request: The amendments would revise the
Technical Specifications (TSs) to allow the reactor building pressure
boundary to be opened under administrative controls.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR), 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
Criterion 1:
Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to TS 3.6.10 and TS 3.6.16 have no effect
upon accident probabilities or consequences. The changes proposed
herein will have no impact upon the Reactor Building or AVS [Annulus
Ventilation System] relative to the performance of their design
functions. These structures/systems will continue to be available
and will function as designed during and following all accidents for
which their performance is credited in the plant safety analyses.
The proposed administrative controls for TS 3.6.16 will ensure the
restoration of the Reactor Building pressure
[[Page 4384]]
boundary when required, thereby enhancing nuclear safety. No design
changes are being made to the plant itself; therefore, there will be
no impact upon the probability of any accident occurring. Since the
performance of these systems will not be adversely impacted, there
will be no impact upon accident consequences.
Criterion 2:
Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to TS 3.6.10 and TS 3.6.16 do not introduce
any changes or mechanisms that create the possibility of a new or
different kind of accident. No design changes are being made to the
plant which would result in the introduction of new accident causal
mechanisms. The proposed changes do not introduce any new equipment,
any change to existing equipment, or any change to the manner in
which the plant is operated. No new effects or malfunctions will
therefore be created.
Criterion 3:
Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes to TS 3.6.10 and TS 3.6.16 maintain the
required design margins of the Reactor Building and AVS for all
accidents for which their function is assumed. All required General
Design Criteria (GDCs) contained in 10 CFR 50, Appendix A, ``General
Design Criteria for Nuclear Power Plants'' will continue to be
satisfied following NRC approval of these proposed changes. In
addition, margin of safety is related to the confidence in the
fission product barriers to function as designed during and
following an accident. These barriers include the fuel cladding, the
Reactor Coolant System, and the Containment System. The changes
proposed in this submittal have no adverse impact upon the
performance of any of these barriers to perform their design
functions during or following an accident.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC
28202.
NRC Branch Chief: Gloria Kulesa.
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 amendment request: September 16, 2010.
Description of amendment request: The amendments would revise
Technical Specification 3.3.2, ``Engineered Safety Feature Actuation
System (ESFAS) Instrumentation,'' to replace the references to the
outdated logic per train per doghouse with updated references which
reflect the license amendment granted by the U.S. Nuclear Regulatory
Commission staff on April 2, 2009.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configurations of the facility. The proposed changes do not alter or
prevent the ability of structures, systems and components (SSCs) to
perform their intended function to mitigate the consequences of an
initiating event within the assumed acceptance limits. In review of
the discussion above (Section 4.1 Significant Hazards Consideration)
it can be concluded the probability or consequences of any accident
previously evaluated are not increased. This LAR requests
administrative changes only.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This revision will not impact the accident analysis. The
proposed changes will not alter the requirements of the ESFAS or its
function during accident conditions. No new or different accidents
result from the changes proposed. The changes do not involve a
physical alteration of the plant (i.e., no new or different type of
equipment will be installed) or any changes in methods governing
normal plant operation. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analyses assumptions. In review of the discussion above
(Section 4.1 Significant Hazards Consideration) it can be concluded
that these changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This LAR requests administrative changes only.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not affected by these changes. The proposed changes will not
result in plant operation in a configuration outside the design
basis. The proposed changes do not adversely affect systems that
respond to safely shutdown the plant and to maintain the plant in a
safe shutdown condition. In review of the discussion above (Section
4.1 Significant Hazards Consideration) it can be concluded that the
proposed changes do not involve a significant reduction in the
margin of safety. This LAR requests administrative changes only.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC
28202.
NRC Branch Chief: Gloria Kulesa.
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-369, 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina; 50-
413 and 50-414, Catawba Nuclear Station, Units 1 and 2, York County,
South Carolina
Date of amendment request: June 29, 2010.
Description of amendment request: The amendments would revise
Technical Specification (TS) 3.3.1, ``Reactor Trip System (RTS)
Instrumentation'' and TS 3.3.2, ``Engineered Safety Feature Actuation
System (ESFAS) Instrumentation.''
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 specific Technical Specification changes are associated with
(1) the specific Allowable Values for various RTS and ESFAS
channels, including instrumentation associated with neutron flux,
containment pressure, pressurizer pressure, pressurizer water level,
reactor coolant flow, reactor coolant pump underfrequency, steam
generator water level, turbine impulse pressure, steam line
pressure, and reactor coolant temperature; (2) the addition of
specific requirements to be taken if an instrument channel setpoint
is outside its predefined as-found tolerance; and (3) the addition
of specific requirements regarding resetting of an instrument
channel setpoint within an as-left tolerance.
The RTS and ESFAS instrumentation is accident mitigation
equipment and does not affect the probability of any accident being
initiated. In addition, none of the abovementioned proposed
Technical
[[Page 4385]]
Specification changes affect the probability of any accident being
initiated.
The proposed changes to TS Allowable Values are based on
methodology that is consistent with the intent of ISA [Instrument
Society of America] Standard RP67.04-1994, Part II, ``Methodologies
for the Determination of Setpoints for Nuclear Safety Related
Instrumentation,'' and will preserve assumptions in the applicable
accident analyses. None of the proposed changes alter any assumption
previously made in the radiological consequences evaluations, nor do
they affect mitigation of the radiological consequences of an
accident previously evaluated.
In summary, the proposed changes will 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.
No new accident scenarios, failure mechanisms, or single
failures are introduced as a result of any of the proposed changes.
The RTS and ESFAS are not capable by itself of initiating any
accident. No physical changes to the overall plant are being
proposed. No changes to the overall manner in which the plant is
operated are being proposed. The proposed changes do not introduce
any new failure modes.
Therefore, none of the proposed changes will create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their intended functions.
These barriers include the fuel cladding, the reactor coolant system
pressure boundary, and the containment barriers. The proposed
changes will not have any impact on these barriers. Plant actuation
features and Nominal Trip Setpoints will be unchanged and will
actuate prior to exceeding any analytical limits. No accident
mitigating equipment will be adversely impacted.
Therefore, existing safety margins will be preserved. None of
the proposed changes will involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC
28202.
NRC Branch Chief: Gloria Kulesa.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: October 28, 2010.
Description of amendment request: The proposed amendment would
modify Clinton Power Station Technical Specifications (TS) Section
3.8.1, ``AC Sources Operating,'' by revising certain Surveillance
Requirements (SR) related to the Division 3 alternating current (AC)
Sources. The Division 3 AC Sources are independent sources of offsite
and onsite AC power primarily dedicated to the High-Pressure Core Spray
(HPCS) system. The TS currently prohibit performing the testing
required by SR 3.8.1.8 and SR 3.8.1.12 in Modes 1 or 2, and prohibit
performing the testing required by SR 3.8.1.11, SR 3.8.1.16, and SR
3.8.1.19 in Modes 1, 2, or 3. The proposed amendment would remove these
Mode restrictions and allow all five of the identified SRs to be
performed in any operating Mode for the Division 3 AC Sources.
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: EGC has evaluated whether or
not a significant hazards consideration is involved with the proposed
amendment by focusing on the three standards set forth in 10 CFR 50.92,
``Issuance of Amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Division 3 (i.e., HPCS) diesel generator (DG) and its
associated emergency loads are accident mitigating features, not
accident initiators. Therefore, the proposed TS changes to allow the
performance of certain Division 3 AC Sources surveillance testing in
any plant operating Mode will not significantly impact the
probability of any previously evaluated accident.
The design of plant equipment is not being modified by the
proposed changes. As such, the ability of the Division 3 AC Sources
to respond to a design basis accident will not be adversely impacted
by the proposed changes. Testing procedures include steps to ensure
that injection into the reactor vessel is precluded. The proposed
changes to the TS surveillance testing requirements for the Division
3 AC Sources do not affect the operability requirements for the AC
Sources, as verification of such operability will continue to be
performed as required. Continued verification of operability
supports the capability of the Division 3 AC Sources to perform
their required functions of providing emergency power to HPCS system
equipment, consistent with the plant safety analyses. Limiting
testing to only one AC Source at a time ensures that design basis
requirements are met. Should a fault occur while testing the
Division 3 AC Sources, there would be no significant impact on any
accident consequences since the other two divisional AC Sources and
associated emergency loads would be available to provide the minimum
safety functions necessary to shut down the unit and maintain it in
a safe shutdown condition.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
No changes are being made to the plant that would introduce any
new accident causal mechanisms. Equipment will be operated in the
same configuration with the exception of the plant operating mode in
which the Division 3 AC Sources surveillance testing is conducted.
Performance of these surveillances tests while online will continue
to verify operability of the Division 3 AC Sources. The proposed
amendment does not impact any plant systems that are accident
initiators and does not adversely impact any accident mitigating
systems.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is related to confidence in the ability of the
fission product barriers (fuel cladding, reactor coolant system, and
primary containment) to perform their design functions during and
following postulated accidents. The proposed changes to the TS
surveillance testing requirements for the Division 3 AC Sources do
not affect the operability requirements for the AC Sources, as
verification of such operability will continue to be performed as
required. Continued verification of operability supports the
capability of the Division 3 AC Sources to perform their required
function of providing emergency power to HPCS system equipment,
consistent with the plant safety analyses. Consequently, the
performance of the fission product barriers will not be adversely
impacted by implementation of the proposed amendment. In addition,
the proposed changes do not alter setpoints or limits established or
assumed by the accident analysis. Further, performing Division 3 AC
Sources surveillance activities online increases the Division 3 DG
and HPCS system availability during refueling outages and allows the
testing of the Division 3 systems to be conducted when both Division
1 and 2 systems are required to be OPERABLE.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
[[Page 4386]]
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. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Robert D. Carlson.
Florida Power and Light Company (FPL), Docket Nos. 50-250 and 50-251,
Turkey Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: July 16, 2010.
Description of amendment request: The amendments would revise the
Technical Specifications (TSs) to adopt Nuclear Regulatory Commission
(NRC)-approved Revision 3 to Technical Specification Task Force (TSTF)
Improved Standard Technical Specification Change Traveler, TSTF-448,
``Control Room Envelope Habitability.'' The proposed amendments include
changes to the TS requirements related to control room envelope (CRE)
habitability in TS 3/4.7.5, ``Control Room Emergency Ventilation System
(CREVS),'' and TS Section 6.8, ``Administrative Controls--Procedures
and Programs.'' This submittal satisfies the commitment identified in
FPL's letter dated August 10, 2007, to adopt the applicable portions of
TSTF-448. Additionally, this application updates the original submittal
of license amendment request 194 dated September 26, 2008, in response
to an NRC request for additional information to remove any reference of
unapproved TSTF-508, which has been done.
The NRC staff published a notice of opportunity for comment in the
Federal Register on October 17, 2006 (71 FR 61075), on possible
amendments adopting TSTF-448, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line-item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on January 17, 2007 (72 FR 2022). The licensee affirmed the
applicability of the following NSHC determination in its application
dated July 16, 2010.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1: The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2: The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Accident Previously
Evaluated.
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Criterion 3: The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. The
proposed change does not adversely affect systems that respond to
safely shut down the plant and to maintain the plant in a safe
shutdown condition.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Douglas A. Broaddus.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: October 29, 2010.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.8.4, ``DC [Direct Current]
Sources--Operating,'' and TS 3.8.6, ``Battery Cell Parameters.''
Specifically, the proposed changes would replace non-conservative
minimum voltages in Surveillance Requirement 3.8.4.1 for the 125 volt
direct current (V DC) and 250 V DC essential batteries, and the non-
conservative battery specific gravity values listed in TS Table 3.8.6-
1, ``Battery Cell Parameter Requirements.''
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.
Performing surveillances that verify terminal voltage and
specific gravity of batteries is not a precursor of any accident
previously evaluated. Restoring battery limits to conservative
values does not significantly affect the method of performing the
surveillances, such that the probability of an accident would be
affected. Therefore, the proposed changes do not result in a
significant increase in the probability of an accident previously
evaluated.
Restoring battery limits to conservative values so that
batteries are maintained in
[[Page 4387]]
accordance with plant design basis ensures they provide the power
assumed in design basis accident mitigation calculations. Therefore,
the change does not involve a significant increase in the
consequences of an accident previously evaluated.
NPPD [Nebraska Public Power District] concludes that the
proposed changes do 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.
The proposed change does not involve any modification to the
plant or equipment or how they are operated. Therefore, NPPD
concludes that these proposed changes do 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 proposed change will continue to ensure station batteries
are able to perform their design function as assumed in calculations
that evaluate their function during design basis accidents. The
proposed change actually increases the margin of safety by restoring
conservatisms inherent in battery design and manufacturer's
recommendations. Based on this, the ability of CNS [Cooper Nuclear
Station] to mitigate the design basis accidents that rely on
operation of the station batteries is not adversely impacted.
Therefore, NPPD concludes that these 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 No. 1, Washington County, Nebraska
Date of amendment request: July 12, 2010.
Description of amendment request: This notice is being reissued in
its entirety due to missing statements from the description of the
amendment request in the notice published in the Federal Register on
December 28, 2010 (75 FR 81671). The proposed amendment would modify
Item 1 of Table 2-5, ``Instrumentation Operating Requirements for Other
Safety Feature Functions,'' of Technical Specification (TS) 2.15,
``Instrumentation and Control Systems,'' to provide new Note (e), and
Surveillance Requirement (SR) Items 1 and 2 of Table 3-3, ``Minimum
Frequencies for Checks, Calibrations and Testing of Miscellaneous
Instrumentation and Controls,'' of TS 3.1, ``Instrumentation and
Control,'' which pertain to operability of the primary and secondary
control element assembly (CEA) position indication system (CEAPIS)
channels. A new SR is proposed for Item 4 of Table 3-3 of TS 3.1, which
will verify the position of CEAs each shift. The proposed amendment
will ensure that CEA alignment is maintained during power operations so
that the power distribution and reactivity limits defined by the design
power peaking and shutdown margin (SDM) limits are preserved. The
proposed amendment would also revise TS 2.10.2(7)c regarding actions to
be taken when the regulating CEA groups are inserted below the Long
Term Insertion Limit. The TS would be revised to require actions to be
taken when either time interval is exceeded, which would also make TS
2.10.2(7)c more consistent with Combustion Engineering (CE) Standard
Technical Specifications (STS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment will allow plant operation to continue
when a CEAPIS channel is inoperable by requiring prompt verification
of CEA positions following CEA movement. CEAs are most likely to
become misaligned during movement and therefore, this change will
cause CEA alignment errors to be promptly detected and corrected. It
is appropriate to clarify that CEAPIS channels are not subject to
the requirements of TS 2.15(1), (2), and (3) as they are not
designed to be placed in trip or bypass, nor are they engineered
safety feature (ESF) or isolation logic subsystems.
The proposed amendment does not alter the requirements of TS
2.15(4) regarding the rod block function of the secondary CEAPIS
channel. Should the secondary CEAPIS channel or its rod block
function be inoperable, several additional CEA deviation events are
possible. However, this situation is already addressed by TS
2.15(4), which requires the CEAs (rods) to be maintained fully
withdrawn with the control rod drive system mode switch in the off
position except when manual motion of CEA Group 4 is required to
control axial power distribution. This is the same position that the
CEAs must be in (fully withdrawn) when the plant is at power (Mode
1) in order to utilize distributed control system (DCS) core mimic
to CHANNEL CHECK the CEAPIS channels.
If it was not possible to use DCS core mimic to verify the
primary CEAPIS channel as would be the case if CEA Group 4 was
inserted to control axial power distribution, then the primary
CEAPIS channel would be declared inoperable when the CHANNEL CHECK
could not be accomplished. The plant would then be placed in hot
shutdown (Mode 3) within 12 hours in accordance with TS 2.15(4).
Therefore, although the proposed amendment will allow a CEAPIS
channel to be inoperable indefinitely, there is no significant
increase in the probability or consequences of an accident as the
requirements of TS 2.15(4) will continue to be met. This serves to
prevent the type of CEA deviation events that the rod block function
was designed for.
Replacing the current method of verifying CEAPIS data with the
defined term CHANNEL CHECK is an improvement that provides
additional flexibility without weakening the intent of the
surveillance. As a result, when it is feasible to obtain CEA
position indication from DCS core mimic (i.e., when the CEAs are
either fully inserted or fully withdrawn), the primary and secondary
CEAPIS channels will be compared with DCS core mimic indication as
well as each other.
As an additional means of verifying CEA positions, DCS core
mimic indication provides added confidence that the CEAs are in the
indicated positions. Should the primary or secondary CEAPIS channel
become inoperable, the accuracy and reliability of DCS core mimic
indication is assured by its previous comparison with both OPERABLE
channels. Comparison of the OPERABLE CEAPIS channel with DCS core
mimic will satisfy the required CHANNEL CHECK and allow continued
operation while the inoperable channel is repaired. The proposed
amendment ensures that the CEA alignment required by TS 2.10.2(4) is
met each shift by requiring all full length (shutdown and
regulating) CEAs to be positioned within 12 inches of all other CEAs
in the group.
The change proposed for TS 2.10.2(7)c incorporates more
conservative wording to ensure that the regulating CEA groups are
maintained within the Long Term Insertion Limit. The proposed change
will ensure that corrective actions are taken if either time
interval is exceeded and makes TS 2.10.2(7)c more consistent with CE
STS.
The proposed amendment does not alter the plant configuration,
require new plant equipment to be installed, alter accident analysis
assumptions, add any initiators, or affect the function of plant
systems or the manner in which systems are operated, maintained,
modified, tested, or inspected.
As an additional means of verifying primary and secondary CEAPIS
data, DCS core mimic indication increases confidence in the
reliability of CEAPIS data.
The proposed amendment will help minimize unplanned shutdowns
that can
[[Page 4388]]
cause plant transients yet continues to ensure that power
distribution and reactivity limits are maintained. Therefore, it is
concluded that this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not change the design function or
operation of the primary or secondary CEAPIS channels. If one CEAPIS
channel should become inoperable, the position of CEAs will be
verified within 15 minutes of any CEA movement to quickly detect and
correct CEA alignment errors. Data from each CEAPIS channel will
continue to be compared to the other channel each shift as before.
However, a CHANNEL CHECK will require that CEAPIS channel data also
be compared with DCS core mimic indication when it is available.
Thus, when the CEAPIS channels are required to be OPERABLE, there
will be at least two means of verifying the position of CEAs or else
appropriate actions must be taken. The CEA alignment required by TS
2.10.2(4) is assured by requiring verification each shift that all
full length (shutdown and regulating) CEAs are positioned within 12
inches of all other CEAs in the group.
No changes are proposed to testing and calibration of the CEAPIS
channels and these requirements will continue to ensure that they
are capable of performing their design function. Use of the defined
term CHANNEL CHECK is an appropriate surveillance method as it
requires that the channel be compared with other independent
channels measuring the same variable where feasible. DCS core mimic
is a diverse, accurate and reliable means of verifying CEA positions
when the CEAs are fully inserted or fully withdrawn. The change
proposed for TS 2.10.2(7)c ensures that appropriate corrective
actions are taken when the regulating CEA groups are below the Long
Term Insertion Limit in excess of either of the specified time
intervals.
No new structures, systems, or components (SSCs) are being
installed, and no credible new failure mechanisms, malfunctions, or
accident initiators are created. Therefore, the proposed amendment
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.
When a CEAPIS channel is inoperable, the proposed amendment
allows plant operation to continue but requires more frequent
verification of CEA positions following any CEA movement, which is
when CEAs are most likely to become misaligned. This will enable CEA
alignment errors to be detected and corrected more promptly. As
CEAPIS channels are not designed to be placed in trip or bypass, nor
are they engineered safety feature (ESF) or isolation logic
subsystems, it is appropriate to clarify that TS 2.15(1), (2), and
(3) do not apply. FCS normally operates with the CEAs fully
withdrawn and maintains reactivity control by adjusting reactor
coolant system (RCS) boric acid concentration. When the CEAs are
fully withdrawn (or fully inserted), DCS core mimic indication
provides accurate and reliable indication of CEA positions suitable
for comparison with the primary and secondary CEAPIS channels. Thus,
even with one CEAPIS channel inoperable, a diverse means of
verifying the accuracy of the OPERABLE CEAPIS channel will be
available. The accuracy and reliability of DCS core mimic is assured
by testing conducted each refueling outage with continued assurance
provided by comparison with primary and secondary CEAPIS each shift.
The change also ensures that the CEA alignment required by TS
2.10.2(4) is met each shift by requiring all full length (shutdown
and regulating) CEAs to be positioned within 12 inches of all other
CEAs in the group. The proposed amendment does not alter the TS
2.15(4) requirement to place the reactor in hot shutdown in the
event that both CEAPIS channels are inoperable. The change proposed
for TS 2.10.2(7)c incorporates more conservative wording to ensure
that the regulating CEA groups are maintained within the Long Term
Insertion Limit.
The proposed amendment will help minimize unplanned shutdowns
that can cause plant transients yet continues to ensure that power
distribution and reactivity limits are maintained. The proposed
amendment does not alter the plant configuration, require new plant
equipment to be installed, alter accident analysis assumptions, add
any initiators, or affect the function of plant systems or the
manner in which systems are operated, maintained, modified, tested,
or inspected. Therefore, the proposed amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: August 16, 2010, as supplemented by
letter dated September 27, 2010.
Description of amendment request: The proposed amendment would
remove the Technical Specification (TS) limiting condition for
operation (LCO) 2.15, ``Instrumentation and Control Systems,'' Table 2-
5, ``Instrumentation Operating Requirements for Other Safety Feature
Functions,'' Items 3, 4, and 5, the associated Notes a, b, c, and d,
and the associated footnote, for power-operated relief valve (PORV) and
pressurizer safety valve (PSV) acoustic position indication and tail
pipe temperature from the Fort Calhoun Station (FCS) TS. The proposed
amendment would also revise the surveillance requirement (SR), TS 3.1,
``Instrumentation and Control,'' Table 3-3, ``Minimum Frequencies for
Checks, Calibrations and Testing of Miscellaneous Instrumentation and
Controls,'' Items 21, 23, and 24 for PORV Operation and Acoustic
Position Indication, Safety Valve Acoustic Position Indication, and
PORV/Safety Valve Tail Pipe Temperature, respectively. Specifically,
Table 3-3, Item 21 will be revised to reflect the performance of the
PORV operation channel functional test on its existing refueling
frequency and deletes the monthly frequency denoted in the TS for the
acoustic position indication which would also be more aligned with
NUREG-1432, ``Standard Technical Specifications, Combustion Engineering
Plants,'' Revision 3, for PORV operation; and Items 21, 23, and 24 will
be revised to relocate the acoustic position indication and tail pipe
temperature indication SRs from the FCS TS. In conjunction with the
proposed TS changes, operability and surveillance requirements for the
acoustic position indication and tail pipe temperature indication
instrumentation would be incorporated into the FCS Updated Safety
Analysis Report (USAR) and associated plant procedures.
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 safety valve acoustic position indication does not affect
the operation of its associated spring-loaded safety valve. As such,
the proposed change does not increase the probability of an
accident. The acoustic monitor and tail pipe temperature indication
are only two of the indications used to identify that a safety valve
is open. Other indications are available to the operators and alarm
in the control room. The acoustic monitor is only one of the
indications that the abnormal and emergency procedures direct
operators to use to diagnose the opening of a safety valve. The
failure of the power operated relief valve (PORV)/safety valve
position instrumentation is not assumed to be an initiator of any
analyzed event in the Updated Safety Analysis Report (USAR). The
proposed changes do not alter
[[Page 4389]]
the physical design of the PORVs/safety valves or any other plant
structure, system or component (SSC). The changes would remove the
PORV/safety valve position indicator operability and surveillance
requirements from the Fort Calhoun Station (FCS) Technical
Specifications (TS), and incorporate the requirements for this
instrumentation into a licensee-controlled document under the
control of 10 CFR 50.59.
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 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 instrumentation is not needed for manual operator actions
necessary for safety systems to accomplish their safety function for
the design basis accident events. The acoustic position indicator
and tail-pipe temperature instrumentation provides only alarm and
PORV/safety valve position indication, and does not provide an input
to any automatic trip function. Diverse means are available to
monitor PORV/safety valve position, and operability and surveillance
requirements will be established in a licensee-controlled document
to ensure the reliability of the PORV/safety valve position
monitoring capability. Changes to these requirements will be subject
to the controls of 10 CFR 50.59, providing the appropriate level of
regulatory control. In addition, the PORV operation is currently
tested on a refueling frequency, which is aligned with the
surveillance 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.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of amendment request: January 27, 2010.
Brief description of amendment request: The amendment revises
Section 2.E. of the Palisades Nuclear Plant (PNP) Renewed Facility
Operating License to remove the name of the former operator of the
plant in the title of the PNP physical security plan and replace it
with Entergy Nuclear. The change also removes the security plan
revision number and the date the plan was submitted to the Nuclear
Regulatory Commission.
Date of publication of individual notice in Federal Register:
November 18, 2010 (75 FR 70708).
Expiration date of individual: January 17, 2011
Notice of Issuance of Amendments to Facility Operating 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.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, 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 envir