[Federal Register Volume 75, Number 192 (Tuesday, October 5, 2010)]
[Notices]
[Pages 61521-61530]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2010-24815]
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NUCLEAR REGULATORY COMMISSION
[NRC-2010-0309]
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
[[Page 61522]]
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 September 9, 2010, to September 22, 2010.
The last biweekly notice was published on September 21, 2010 (75
FR57521).
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), Section 50.92, this means that operation of the
facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
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 File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
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, Public File
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
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
[[Page 61523]]
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://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 (866) 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
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, Public File Area O1F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the ADAMS Public Electronic Reading
[[Page 61524]]
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, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: May 28, 2010.
Description of amendment request: The amendments would revise the
Technical Specifications (TS) to allow manual operation of the
containment spray system (CSS) and to change the setpoints for the
refueling water storage tank (RWST).
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:
Criterion 1: Does the proposed amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No.
The CSS and RWST are accident mitigation equipment. As such,
changes in operation of these systems cannot have an impact on the
probability of an accident.
The RWST will continue to comply with all applicable regulatory
requirements and design criteria following approval of the proposed
changes (e.g., train separation, redundancy, and single failure).
The water level on the containment floor will be higher at the start
of transfer to the containment sump but will remain below the
maximum design level analyzed for equipment submergence. The change
in the sump pH will not result in a significant increase in
radiological consequences of a LOCA [loss-of-coolant accident].
Therefore, the design functions performed by the equipment are not
changed.
The proposed change alters the method of controlling the safety
system following a design basis event so that manual actions are
substituted for automatic actions. Calculations and simulator
exercises confirm these actions will be taken within the appropriate
scenario sequence timing to provide containment cooling and source
term reduction.
The delay in CS [containment spray] operation will result in an
increase in containment temperature, containment pressure, offsite
dose, and control room dose during a LOCA or high energy line break
inside containment. Containment analyses have been performed to
demonstrate that containment pressure and temperature remain within
the design limits and there is no significant impact on the
environmental qualification for equipment inside containment. The
reduction in fission product removal due to delayed CS operation
does not result in exceeding the offsite dose and control room dose
limits in 10 CFR 50.67. The analysis of the change in containment
conditions due to a single failure of an operating spray pump and
the suspension of CS determined that the pressure remained below the
design limits.
The proposed change to adopt [Technical Specification Task
Force] TSTF-493, Rev. 4, on a limited basis clarifies requirements
for instrumentation to ensure the instrumentation will actuate as
assumed in the safety analysis. Instruments are not an assumed
initiator of any accident previously evaluated. As a result, the
proposed change will not increase the probability of an accident
previously evaluated. The proposed change will ensure that the
instruments actuate as assumed to mitigate the accidents previously
evaluated. As a result, the proposed change will not increase the
consequences of an accident previously evaluated.
Based on this discussion, the proposed amendment does not
significantly increase the probability or consequences of an
accident previously evaluated.
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 modification to the low level setpoint will not install any
new plant equipment. The setpoint will continue to be included
within the engineered safeguards features instrumentation and
monitored according to the applicable surveillance requirements. The
evaluation of the new level setpoint and the change in the
switchover sequence concluded that the equipment aligned to the sump
will continue to have sufficient suction pressure prior to
containment sump suction switchover. The design of the RWST low
level instrumentation complies with all applicable regulatory
requirements and design criteria.
The overall function of the CSS is not changed by this proposed
amendment. The proposed change alters the method of controlling the
safety system following a design basis event so that manual actions
are substituted for automatic actions. Calculations confirm that
these actions will be taken within the appropriate scenario sequence
timing to provide containment cooling and source term reduction with
no significant increase in radiological consequences and without
exceeding containment design limits.
The proposed change to adopt TSTF-493, Rev. 4 on a limited basis
does not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operation. The change does not alter
assumptions made in the safety analysis but ensures that the
instruments behave as assumed in the accident analysis. The proposed
change is consistent with the safety analysis assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
Criterion 3: Does the proposed amendment involve a significant
reduction in a margin of safety?
Response: No.
The proposed change will increase the calculated radiological
dose at the site boundary and in the control room. However, the
calculations demonstrate that the dose consequences at the site
boundary, low population zone, and control room remain within
regulatory acceptance limits of 10 CFR 50.67.
Additional analysis concluded:
Peak containment pressure for analyzed design basis
accidents will not be significantly increased and containment design
limits will not be exceeded.
Assumptions used in the environmental qualification of
equipment exposed to the containment atmosphere remain bounding.
Pumps aligned to the RWST and to the containment sump
will have adequate suction pressure.
The CSS will retain its ability to undergo all
appropriate testing requirements following implementation of the
proposed amendment. These testing requirements are conducted in
accordance with the McGuire Inservice Testing Program and TS 3.6.6.
It is estimated that the implementation of this license
amendment request will result in an approximate 22% reduction in
core damage frequency. This amendment request is based on the
Nuclear Energy Institute (NEI) and the Pressurized Water Reactor
(PWR) Owners Group initiative to extend the post-Loss of Coolant
Accident (LOCA) injection phase and delay the onset of the
containment sump recirculation phase.
The proposed change to adopt TSTF-493, Rev. 4 on a limited basis
clarifies the requirements for instrumentation to ensure the
instrumentation will actuate as assumed in the accident analysis. No
change is made to the accident analysis assumptions and no margin of
safety is reduced as part of this change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC
28202.
NRC Branch Chief: Gloria Kulesa.
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.
Description of amendment request: The proposed amendment would
revise
[[Page 61525]]
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 ([deg]F) as a consequence of inservice leak and
hydrostatic testing, and as a consequence of scram time testing
initiated in conjunction with an inservice leak or hydrostatic test,
while considering operational conditions to be in Mode 4. The proposed
change is consistent with NRC-approved Technical Specification Task
Force (TSTF) Improved Standard Technical Specification Traveller, TSTF-
484, ``Use of TS 3.10.1 for Scram Time Testing Activities,'' that was
announced in the Federal Register on October 27, 2001 (71 FR 63050), as
part of the consolidated Line Item Improvement Process (CCIIP).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Technical Specifications currently allow for operation at > 200
[deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. Extending the
activities that can apply this allowance will not adversely impact
the probability or consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Technical Specifications currently allow for operation at > 200
[deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. No new
operational conditions beyond those currently allowed by LCO 3.10.1
are introduced. The extended allowances would result from operations
that commence at reduced temperatures, but approach the normal MODE
4 limit of 200 [deg]F prior to completion of the inspections or
testing. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements or eliminate any existing requirements. The
changes do not alter assumptions made in the safety analysis. The
proposed changes are consistent with the safety analysis assumptions
and current plant operating practice.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Technical Specifications currently allow for operation at > 200
[deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. Extending the
activities that can apply this allowance will not adversely impact
any margin of safety. Allowing completion of inspections and testing
and supporting completion of scram time testing initiated in
conjunction with an inservice leak or hydrostatic test prior to
power operation, results in enhanced safe operations by eliminating
unnecessary maneuvers to control reactor temperature and pressure.
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: Joseph A. Aluise, Assistant General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: August 19, 2010.
Description of amendment request: The proposed amendment would
revise Technical Specifications to be consistent with Standard
Technical Specifications 3.6.1.8 ``Suppression Chamber-to-Drywell
Vacuum Breakers'' and 3.6.2.5 ``Drywell-to-Suppression Chamber
Differential Pressure,'' along with the associated Bases, of NUREG-
1433, Revision 3, ``Standard Technical Specifications General Electric
Plants, BWR/4,'' modified to account for plant specific design details.
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 does not significantly increase the
probability or consequences of an accident since it does not involve
a modification to any plant equipment or affect how plant systems or
components are operated. No design functions or design parameters
are affected by the proposed amendment. The proposed amendment
involves the operation and testing of Primary Containment systems
but does not impact containment design or performance requirements.
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.
The proposed change does not involve any physical alteration of
plant equipment and does not change the method by which any safety-
related system performs its function. No new or different types of
equipment will be installed and the basic operation of installed
equipment is unchanged. The methods governing plant operation and
testing remain consistent with current safety analysis assumptions.
The proposed amendment involves the operation and testing of Primary
Containment systems but does not alter the way that the systems are
operated or how the tests are performed. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change ensures that the safety functions of the
pressure suppression chamber-drywell vacuum breakers and drywell-
suppression chamber differential pressure are fulfilled by
incorporating the guidance of NUREG-1433. The proposed amendment
does not involve a physical modification of the plant and does not
change the design or function of any component or system. Therefore,
the proposed amendment will not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Nancy Salgado.
[[Page 61526]]
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: August 10, 2010.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.9.3, ``Reactor Building
Penetrations,'' to allow reactor building flow path(s) providing direct
access from the reactor building atmosphere to the outside atmosphere
to be unisolated under administrative control, during movement of
irradiated fuel assemblies. The proposed change is consistent with
Technical Specification Task Force (TSTF) Technical Change Traveler
312, Revision 1, ``Administratively Control Containment Penetrations.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The status of the penetration flow paths during fuel movement in
the reactor building has no affect on the probability of the
occurrence of any accident previously evaluated. The proposed change
does not alter any plant equipment or operating practices in such a
manner that the probability of an accident is increased. Since the
consequences of a fuel handling accident (FHA) inside the reactor
building with open penetrations flow paths is bounded by the current
FHA analyses and the probability of an accident is not affected by
the status of the penetration flow paths, the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The open reactor building penetration flow paths are not
accident initiators. The proposed allowance to open the reactor
building penetrations during fuel movement inside the reactor
building will not adversely affect plant safety functions or
equipment operating practices such that a new or different accident
could be created. Therefore, the proposed change does not create the
possibility of an accident of a different kind than previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Technical Specification (TS) 3.9.3 closure requirements for
reactor building penetrations ensure that the consequences of a
postulated FHA inside the reactor building during irradiated fuel
handling activities are minimized. The Limiting Condition for
Operation establishes reactor building closure requirements, which
limit the potential escape paths for fission products by ensuring
that there is at least one integral barrier to the release of
radioactive material. The proposed change to allow the reactor
building penetration flow paths to be open during refueling
operations under administrative controls does not significantly
affect the expected dose consequences of a FHA because the limiting
FHA does not credit reactor building closure or filtration. The
proposed administrative controls provide assurance that prompt
closure of the penetration flow paths will be accomplished in the
event of a[n] FHA inside the reactor building. The provisions to
promptly isolate open penetration flow paths provide assurance that
the offsite dose consequences of a[n] FHA inside containment will be
minimized. Therefore, this 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: Joseph A. Aluise, Assistant General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois; Docket Nos.
STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle
County, Illinois
Date of amendment request: June 29, 2010, as supplemented on August
24, 2010.
Description of amendment request: The proposed amendments would
revise Technical Specifications (TS) Section 3.4.12, ``Low Temperature
Overpressure Protection (LTOP) System,'' to correct an inconsistency
between the TS, and implementation of procedures and administrative
controls for Safety Injection (SI) pumps required to mitigate a
postulated loss of decay heat removal during mid-loop operation as
discussed in NRC Generic Letter (GL) 88-17, ``Loss of Decay Heat
Removal.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not result in any physical changes to
safety related structures, systems, or components. The proposed
change revises TS 3.4.12 to correct an inconsistency between the TS,
and implementation of procedures and administrative controls for SI
pumps required to mitigate a postulated loss of decay heat removal
during mid-loop operation as discussed in GL 88-17. Specifically,
the proposed change adds a note to TS LCO [limiting condition for
operation] 3.4.12 that states: ``For the purpose of protecting the
decay heat removal function, one or more SI pumps may be capable of
injecting into the RCS in MODE 5 and MODE 6 when the reactor vessel
head is on provided pressurizer level is <= 5 percent.'' The
proposed change corrects an oversight introduced during the
conversion of the Braidwood Station and Byron Station TS to the ITS
[Improved TS].
The probability of occurrence of an accident is not increased
since the proposed change will continue to require that no SI pumps
are capable of injecting into the RCS in Modes 5 and 6 with
pressurizer level greater than 5 percent.
The NRC has previously evaluated the allowance for one or more
SI pumps to be capable of injecting into the RCS in Mode 5 or Mode 6
when the reactor vessel head is on provided pressurizer level is <=
5 percent for the Braidwood Station and Byron Station. In a safety
evaluation dated August 31, 1990, related to Braidwood Station,
Units 1 and 2, Amendment 25, and Byron Station, Units 1 and 2,
Amendment 38, the NRC concluded that allowing SI pump capability to
inject into the RCS in Mode 5 or Mode 6 when the reactor vessel head
is on provided pressurizer level is <= 5 percent was acceptable. The
availability of SI pumps under these circumstances does not present
a concern regarding cold overpressure protection since sufficient
air volume exists which allows Operations personnel time to mitigate
the transient. This is in contrast to the analyzed cold overpressure
transients, in which the RCS is assumed to be water solid at the
onset of the event.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises TS 3.4.12 to correct an
inconsistency between the TS, and implementation of procedures and
administrative controls for SI pumps required to mitigate a
postulated loss of decay heat removal during mid-loop operation as
discussed in GL 88-17. Specifically, the proposed change adds a note
to TS LCO 3.4.12 that states: ``For the purpose of protecting the
decay heat removal function, one or more SI pumps may be
[[Page 61527]]
capable of injecting into the RCS in MODE 5 and MODE 6 when the
reactor vessel head is on provided pressurizer level is <= 5
percent.'' The proposed change corrects an oversight introduced
during the conversion of the Braidwood Station and Byron Station TS
to the ITS.
The proposed change is necessary for the purpose of mitigating
the consequences of a loss of decay heat removal during mid-loop
operations. Operation of at least one SI pump is required in some
cases to prevent the core from uncovering. The only new
configuration allowed by the proposed change is the potential of
having one or more SI pumps available in Modes 5 and 6 with
pressurizer level <= 5 percent. The potential overpressurization
accident has been analyzed and accounted for by requiring
pressurizer level to be <= 5 percent if one or more SI pumps are
available.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises TS 3.4.12 to correct an
inconsistency between the TS, and implementation of procedures and
administrative controls for SI pumps required to mitigate a
postulated loss of decay heat removal during mid-loop operation as
discussed in GL 88-17. Specifically, the proposed change adds a note
to TS LCO 3.4.12 that states: ``For the purpose of protecting the
decay heat removal function, one or more SI pumps may be capable of
injecting into the RCS in MODE 5 and MODE 6 when the reactor vessel
head is on provided pressurizer level is <= 5 percent.'' The
proposed change corrects an oversight introduced during the
conversion of the Braidwood Station and Byron Station TS to the ITS.
The proposed note allows one or more SI pumps to be capable of
injecting into the RCS only when pressurizer level is <= 5 percent
in Mode 5 and Mode 6 when the reactor vessel head is on. This
provides protection to limit coolant input capacity during shutdown
in which a pressure fluctuation due to coolant input from the SI
pumps could occur more quickly than an operator could react, while
providing an allowance for one or more SI pumps to be capable of
injecting into the RCS during conditions in which a loss of decay
heat removal could result in rapid core uncovery.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (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 Generation Company, LLC, 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: August 5, 2010.
Description of amendment request: The proposed amendments would
revise technical specification (TS) 5.5.1 Fuel Storage--Criticality, to
include new spent fuel storage patterns that account for both the
increase in fuel maximum enrichment from 4.5 weight percentage (wt%) U-
235 to 5.0 wt% U-235 and the impact on the fuel of higher power
operation proposed under the Extended Power Uprate (EPU) project.
Although the fuel storage has been analyzed at the higher fuel
enrichment in the new criticality analysis, the fuel enrichment limit
of 4.5 wt% U-235 specified in TS 5.5.1 will not be changed under this
license amendment request. The proposed TS changes and a new supporting
criticality analysis are being submitted to revise the current
licensing basis analysis for both new fuel and spent fuel pool storage.
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?
No. The proposed amendments do not change or modify the fuel,
fuel handling processes, fuel storage racks, number of fuel
assemblies that may be stored in the spent fuel pool (SFP), decay
heat generation rate, or the spent fuel pool cooling and cleanup
system. The proposed amendment was evaluated for impact on the
following previously evaluated events and accidents:
a. A fuel handling accident (FHA),
b. A cask drop accident,
c. A fuel mispositioning event,
d. A spent fuel pool boron dilution event,
e. A seismic event, and
f. A loss of spent fuel pool cooling event.
Although the proposed amendment will require increased handling
of the fuel, the probability of a FHA is not significantly increased
because the implementation of the proposed amendment will employ the
same equipment and process to handle fuel assemblies that is
currently used. Also, tests have confirmed that the Metamic inserts
can be installed and removed without damaging the host fuel
assemblies. The FHA radiological dose consequences associated with
fuel enrichment at this level were addressed in LAR [license
amendment request] 196 on Alternative Source Term implementation at
EPU conditions and remain unchanged. Therefore, the proposed
amendments do not significantly increase the probability or
consequences of a FHA.
The proposed amendments do not increase the probability of
dropping a fuel transfer cask because they do not introduce any new
heavy loads to the SFP and do not affect heavy load handling
processes. Also, the insertion of Metamic rack inserts does not
increase the consequences of the cask drop accident because the
radiological source term of that accident is developed from a non-
mechanistically derived quantity of damaged fuel stored in the spent
fuel pool. Therefore, the proposed amendments do not significantly
increase the probability or consequences of a cask drop accident.
Operation in accordance with the proposed amendment will not
change the probability of a fuel mispositioning event because fuel
movement will continue to be controlled by approved fuel handling
procedures. These procedures continue to require identification of
the initial and target locations for each fuel assembly that is
moved. The consequences of a fuel mispositioning event are not
changed because the reactivity analysis demonstrates that the same
subcriticality criteria and requirements continue to be met for the
worst-case fuel mispositioning event.
Operation in accordance with the proposed amendment will not
change the probability of a boron dilution event because the systems
and events that could affect spent fuel pool soluble boron are
unchanged. The consequences of a boron dilution event are unchanged
because the proposed amendment reduces the soluble boron requirement
below the currently required value and the maximum possible water
volume displaced by the inserts is an insignificant fraction of the
total spent fuel pool water volume.
Operation in accordance with the proposed amendment will not
change the probability of a seismic event. The consequences of a
seismic event are not significantly increased because the forcing
functions for seismic excitation are not increased and because the
mass of storage racks with Metamic inserts is not appreciably
increased. Seismic analyses demonstrate adequate stress levels in
the storage racks when inserts are installed.
Operation in accordance with the proposed amendment will not
change the probability of a loss of SFP cooling event because the
systems and events that could affect SFP cooling are unchanged. The
consequences are not significantly increased because there are no
changes in the SFP heat load or SFP cooling systems, structures or
components. Furthermore, conservative analyses indicate that the
current design requirements and criteria continue to be met with the
Metamic inserts installed.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No. The proposed amendments do not change or modify the fuel,
fuel handling processes, fuel racks, number of fuel assemblies that
may be stored in the pool,
[[Page 61528]]
decay heat generation rate, or the spent fuel pool cooling and
cleanup system. The effects of operating with the proposed amendment
are listed below. The proposed amendments were evaluated for the
potential of each effect to create the possibility of a new or
different kind of accident:
a. Addition of inserts to the fuel storage racks,
b. New storage patterns,
c. Additional weight from the inserts,
d. Insert movement above fuel, and
e. Displacement of fuel pool water by the inserts.
Each insert will be placed between a fuel assembly and the
storage cell wall, taking up some of the space available on two
sides of the fuel assembly. Tests confirm that the insert can be
installed and removed without damaging the fuel assembly. Analyses
demonstrate that the presence of the inserts does not adversely
affect spent fuel cooling, seismic capability, or subcriticality.
The aluminum (alloy 6061) and boron carbide materials of
construction have been shown to be compatible with nuclear fuel,
storage racks and spent fuel pool environments, and generate no
adverse material interactions. Therefore, placing the inserts into
the spent fuel pool storage racks cannot cause a new or different
kind of accident.
Operation with the proposed fuel storage patterns will not
create a new or different kind of accident because fuel movement
will continue to be controlled by approved fuel handling procedures.
These procedures continue to require identification of the initial
and target locations for each fuel assembly that is moved. There are
no changes in the criteria or design requirements pertaining to fuel
storage safety, including subcriticality requirements, and analyses
demonstrate that the proposed storage patterns meet these
requirements and criteria with adequate margins. Therefore, the
proposed storage patterns cannot cause a new or different kind of
accident.
Operation with the added weight of the Metamic inserts will not
create a new or different accident. The net effect of the adding the
maximum number of inserts is to add less than one percent to the
weight of the loaded racks. Furthermore, the analyses of the racks
with Metamic inserts installed demonstrate that the stress levels in
the rack modules continue to be considerably less than allowable
stress limits. Therefore, the added weight from the inserts cannot
cause a new or different kind of accident.
Operation with insert movement above stored fuel will not create
a new or different kind of accident. The insert with its handling
tool weighs considerably less than the weight of a single fuel
assembly. Single fuel assemblies are routinely moved safely over
fuel assemblies and the same level of safety in design and operation
will be maintained when moving the inserts. Furthermore, the effect
of a dropped insert to block the top of a storage cell has been
evaluated in thermal-hydraulic analyses. Therefore, the movement of
inserts cannot cause a new or different kind of accident.
Whereas the installed rack inserts will displace a very small
fraction of the fuel pool water volume and impose a very small
reduction in operator response time to previously-evaluated SFP
accidents, the reduction will not promote a new or different kind of
accident. Also, displacement of water along two sides of a stored
fuel assembly may have some local reduction in the peripheral
cooling flow; however, this effect would be small compared to the
flow induced through the fuel assembly and would in no way promote a
new or different kind of accident.
The accidents and events previously analyzed and presented in
the Boraflex Remedy and Alternative Source Term LARs remain
bounding.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
No. The proposed change was evaluated for its effect on current
margins of safety as they relate to criticality, structural
integrity, and spent fuel heat removal capability.
The margin of safety for subcriticality required by 10 CFR
50.68(b)(4) is unchanged. New criticality analysis confirms that
operation in accordance with the proposed amendment continues to
meet the required subcriticality margins.
The structural evaluations for the racks and spent fuel pool
with Metamic inserts installed show that the rack and spent fuel
pool are unimpaired by loading combinations during seismic motion,
and there is no adverse seismic-induced interaction between the rack
and Metamic inserts.
The proposed change does not affect spent fuel heat generation
or the spent fuel pool cooling systems. A conservative analysis
indicates that the design basis requirements and criteria for spent
fuel cooling continue to be met with the Metamic inserts in place,
and displacing coolant. Thermal hydraulic analysis of the local
effects of an installed rack insert blocking peripheral flow show a
small increase in local water and fuel clad temperatures, but will
remain within acceptable limits including no departure from nucleate
boiling.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
Based on the above discussion, FPL has determined that the
proposed change does not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Douglas A. Broaddus.
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.
Virginia Electric and Power Company: Docket Nos. 50-338 and 50-339,
North Anna Power Station, Unit Nos. 1 and 2, Located in Louisa County,
Virginia; and 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and
2, Located in Surry County, Virginia
Date of amendment request: May 6 and February 10, 2010.
Brief description of amendment request: The proposed amendments
will add Optimized ZIRLO as an acceptable fuel rod cladding material
and in addition, propose adding the Westinghouse topical report for
Optimized ZIRLO to the analytical methods used to determine the core
operating limits listed in the Technical Specifications.
Date of publication of individual notice in Federal Register:
August 27, 2010 (75 FR 52781).
Expiration date of individual notice: Comments, September 27, 2010;
Hearing, October 26, 2010.
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
[[Page 61529]]
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].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendment: October 30, 2009, as
supplemented by letters dated April 29 and August 24, 2010.
Brief description of amendment: The amendments consisted of
administrative changes to update the licenses and the technical
specifications as a result of changes that were approved in previously
issued amendments. The amendments removed requirements that are no
longer applicable due to the completion of power uprates, the
replacement of steam generators, the removal of part-length control
element assemblies, the completion of the core protection calculator
upgrade, and made a minor administrative change to the nomenclature of
the containment sump trash racks and screens.
Date of issuance: September 10, 2010.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: Unit 1-179; Unit 2-179; Unit 3-179.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendment revised the Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: January 26, 2010 (75 FR
4113). The supplemental letters dated April 29 and August 24, 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 September 10, 2010.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: November 19, 2009.
Brief description of amendment: The amendment revises the charcoal
testing criteria in Technical Specification 5.5.9, ``Ventilation Filter
Testing Program.''
Date of issuance: September 13, 2010.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 265.
Facility Operating License No. DPR-26: The amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: January 26, 2010 (75 FR
4115).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 13, 2010.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 9, 2009.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3/4 .9.7, ``Crane Travel--Fuel Handling Building,''
to permit certain operations needed for dry cask storage of spent
nuclear fuel. Specifically, the proposed change to this TS, while
continuing to prohibit travel of a heavy load over irradiated fuel
assemblies in the spent fuel pool, would permit travel of loads in
excess of 2,000 pounds over a transfer cask containing irradiated fuel
assemblies, provided a single-failure-proof handling system is used.
Date of issuance: September 13, 2010.
Effective date: As of the date of issuance and shall be implemented
prior to the start of the dry cask storage operations.
Amendment No.: 227.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 17, 2009 (74
FR 59261). The supplemental letters dated June 8 and July 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 as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 13, 2010.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 22, 2009.
Brief description of amendment: The amendment modified the
Technical Specifications (TS) Table 2.2-1 and Table 3.3-1.
Specifically, the TS changes clarify TS Table 2.2-1 Notes (1) and (5),
TS Table 3.3-1 Notes (a) and (c), and TS Table 3.3-1 Actions 2 and 3,
which have resulted in Plant Protection System redundancy issues with
respect to verbatim compliance. While the changes modified the table
notations for the 10-4 percent Bistable in the Tables, they
still maintain the safety function associated with the Core Protection
Calculators and High Logarithmic Power trip functions, and with the
small hysteresis for the 10-4 percent Bistable, there is a
negligible impact on the Control Element Assembly withdrawal analysis.
Additionally, the calculated peak power and heat flux are not
significantly changed.
Date of issuance: September 13, 2010.
[[Page 61530]]
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 228.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 15, 2009 (74
FR 66384).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 13, 2010.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2 (Braidwood), Will County, Illinois;
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2
(Byron), Ogle County, Illinois
Date of application for amendment: March 29, 2010.
Brief description of amendment: The amendments revise Technical
Specification (TS) 5.5.7, ``Reactor Coolant Pump Flywheel Inspection
Program,'' to extend the reactor coolant pump (RCP) motor flywheel
examination frequency from the currently-approved 10-year inspection
interval to an interval not to exceed 20 years for certain Braidwood
and Byron RCPs. These changes are consistent with TS Task Force (TSTF)
traveler TSTF-421, ``Revision to RCP Flywheel Inspection Program (WCAP-
15666),'' Revision 0, that has been approved generically for the
Westinghouse Standard Technical Specifications, NUREG-1431. A notice
announcing the availability of this proposed TS change using the
Consolidated Line Item Improvement Process was published in the Federal
Register on October 22, 2003 (68 FR 60422).
Date of issuance: September 16, 2010.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: Braidwood Unit 1-163; Braidwood Unit 2-163; Byron
Unit No. 1-169; and Byron Unit No. 2-169.
Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66:
The amendments revise the TSs and Licenses.
Date of initial notice in Federal Register: May 18, 2010 (75 FR
27827).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 16, 2010.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Date of application for amendment: April 2, 2010, as supplemented
by letters dated June 19, 2009, and March 31, 2010.
Brief description of amendment: The amendment revises the Exelon
Nuclear Radiological Emergency Plan Annex for Clinton Station, Table B-
1, ``Minimum Staffing Requirements for the On-Shift Clinton Station
ERO,'' to increase the Non-Licensed Operator staffing from two to four,
allow in-plant protective actions to be performed by personnel assigned
to other functions, and replace a Mechanical Maintenance person with a
Non-Licensed Operator.
Date of issuance: September 21, 2010.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 191.
Facility Operating License No. NPF-62: The amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: June 1, 2010 (75 FR
30445).
The June 19, 2009, and March 31, 2010, supplement, contained
clarifying information and did not change the NRC staff's initial
proposed finding of no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 21, 2010.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: January 27, 2010, as
supplemented by letters dated May 12, and May 13, 2010.
Brief description of amendments: The amendments would revise the
Operating License and technical Specifications to implement an increase
of approximately 1.65 percent in rated thermal power from the current
licensed thermal power of 3489 megawatts thermal (MWt) to 3546 MWt.
Date of issuance: September 16, 2010.
Effective date: As of the date of issuance and shall be implemented
within 90 days for Unit 1 and within 90 days of completion of refueling
outage L2R13, which is currently scheduled for March 2011, for Unit 2.
Amendment Nos.: 198/185.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications and License.
Date of initial notice in Federal Register: May 11, 2010 (75 FR
26289).
The May 12, and May 13, 2010, supplements, contained clarifying
information and did not change the NRC staff's initial proposed finding
of no significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 16, 2010.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 23rd day of September 2010.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2010-24815 Filed 10-4-10; 8:45 am]
BILLING CODE 7590-01-P