[Federal Register Volume 76, Number 16 (Tuesday, January 25, 2011)]
[Notices]
[Pages 4381-4390]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-1480]
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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, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
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
[email protected], 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 http://
[[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 http://www.nrc.gov/site-help/e-submittals.html. Participants may attempt to use other software not
listed on the Web site, but should note that the NRC's E-Filing system
does not support unlisted software, and the NRC Meta System Help Desk
will not be able to offer assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through 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 http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an 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 http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at 1-866-672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852-0238, 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
http://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, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have
access to ADAMS or who encounter problems in accessing the documents
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to [email protected].
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
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 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, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
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,
[[Page 4390]]
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-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 licensee-controlled 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
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]
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