[Federal Register Volume 74, Number 35 (Tuesday, February 24, 2009)]
[Notices]
[Pages 8281-8294]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-3515]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0062]
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 staff) 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 January 29, 2009, to February 11, 2009. The
last biweekly notice was published on February 10, 2009 (74 FR 6662).
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 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, 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. Copies of written comments
received may be examined at the Commission'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 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
[[Page 8282]]
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 within 60 days, 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 set forth the specific contentions which the petitioner/requestor
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
petitioner/requestor 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 petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor 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 petitioner/requestor to relief. A petitioner/
requestor 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, and the Commission has not made a final
determination on the issue of no significant hazards consideration, 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, which the NRC
promulgated in August 28, 2007 (72 FR 49139). 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 a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
[email protected], or by calling (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/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer
TM to access the Electronic Information Exchange (EIE), a
component of the E-Filing system. The Workplace Forms Viewer
TM is free and is available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. 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.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it 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 filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE 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
EIE 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 may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC electronic filing
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time,
Monday through Friday, excluding government holidays. The electronic
filing Help Desk can be contacted by telephone at 1-866-672-7640 or by
e-mail at [email protected].
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, in accordance
with 10 CFR 2.302(g), with their initial paper filing requesting
authorization to continue to
[[Page 8283]]
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.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the petition
and/or request should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii).
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://www.ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded
pursuant to an order of the Commission, an Atomic Safety and Licensing
Board, or a 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.
For further details with respect to this amendment action, 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 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
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. 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].
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: September 29, 2008, as supplemented by
letter dated January 16, 2009.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) Sections 5.6.1.3.a and 5.6.1.3.b to
incorporate the results of a new criticality analysis. Specifically the
TSs would be revised to add new requirements for the Boiling Water
Reactor (BWR) spent fuel storage racks containing Boraflex in Spent
Fuel Pools A and B. The requirements for the BWR spent fuel racks as
currently contained in TS 5.6.1.3 would be revised to specify
applicability to the spent fuel storage racks containing Boral in Spent
Fuel Pool B.
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 activity changes the design basis of the BWR Boraflex
storage racks, but does not make physical changes to the facility. The
change to TS Section 5.6.1.3 (BWR Storage Racks in Pools A and B),
which is an update to the administrative controls for maintaining the
required boron concentration in the Boraflex BWR spent fuel storage
racks located in Pools A and B, does not modify the facility.
The accidents currently analyzed in the FSAR [Final Safety Analysis
Report] applicable to the proposed activity are fuel handling
accidents. These accidents include dropping a fuel assembly onto the
top of a fuel rack or in the space between a rack and the pool wall.
These events are caused either by personnel error or equipment
malfunction.
Based on the new criticality analysis, revised acceptance criteria
are needed to ensure the criticality safety of fuel storage in BWR
Boraflex racks in Pools A and B. Similar administrative controls were
previously placed on fuel stored in the PWR [Pressurized Water Reactor]
Boraflex racks in Pools A and B. These changes will eliminate the
dependence on the Boraflex absorber in the BWR storage racks. These
changes do not impact the probability of having a fuel handling
accident and do not impact the consequences of a fuel handling
accident.
Therefore, this amendment 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.
These revised acceptance criteria applicable to the irradiated fuel
stored in the BWR Boraflex racks in Pools A and B are being added to TS
Section 5.6.1.3.a.
The proposed change does not result in any credible new failure
mechanisms, malfunctions or accident initiators not considered in the
original design and licensing bases.
Detailed analyses have been performed to ensure a criticality
accident in Pools A and B is not a credible event. The events that
could lead to a criticality accident are not new. These events include
a fuel mispositioning event, a fuel drop event, and a boron dilution
event. The proposed changes do not impact the probability of any of
these events.
The detailed criticality analyses performed demonstrates that
criticality would not occur following any of these events. Even in a
more likely event, such as a fuel mispositioning event, the acceptance
criteria for keff [the effective multiplication factor]
remains less than or equal to 0.95. In the unlikely event that the
spent fuel storage pool boron concentration were reduced to zero,
keff remains less than 1.0. A criticality accident is
considered ``not credible'' and the proposed action does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Therefore, the proposed change will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Incorporation of the revised criteria for fuel stored in the BWR
Boraflex racks in Pools A and B do not involve a reduction in the
margin of safety. The updated fuel storage condition continues to meet
keff <0.95 with credit for soluble boron and keff
< 1.0 when flooded with unborated water.
The proposed changes for storage of irradiated fuel in BWR Boraflex
racks in
[[Page 8284]]
Pools A and B continues to provide the controls necessary to ensure a
criticality event could not occur in the spent fuel storage pool. The
acceptance criteria are consistent with the acceptance criteria
specified in 10 CFR 50.68, which provide an acceptable margin of safety
with regard to the potential for a criticality event.
Therefore, this 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 T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: August 28, 2008, as supplemented by
letter dated January 19, 2009.
Description of amendment request: The proposed amendment would
implement the Technical Specification Task Force Standard Technical
Specification Change Traveler 449, Revision 4 inspection requirements
for the replacement once through steam generators (OTSGs) that are
being installed during the Crystal River Unit 3 Nuclear Generating
Plant fall 2009 refueling outage. The replacement OTSGs differ from the
existing OTSGs in that the tube material is Alloy 690 thermally treated
in the replacements versus Alloy 600 in the existing OTSGs.
Additionally, this amendment would remove inspection requirements that
are designated for specific damage conditions in the existing OTSGs,
remove tube repair techniques approved by the license amendment No.
233, dated May 16, 2007, for the existing OTSGs, and remove inspection
and reporting requirements specific to those repair techniques.
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. The Proposed Change Does Not Involve a Significant Increase in
the Probability or Consequences of an Accident Previously Evaluated.
The proposed change for replacement OTSGs continues to implement
the current OTSG Program that includes performance criteria which
provide reasonable assurance that the replacement OTSG tubing will
retain integrity over the full range of operating conditions (including
startup, operation in the power range, hot standby, cooldown and all
anticipated transients included in the design specifications). This
change removes repair criteria from the OTSG Program that were approved
by previous License Amendments for the existing Steam Generators which
are not applicable to the replacement OTSGs. It removes references to
use of repairs and reporting of repair results in other Technical
Specification sections. This change removes inspection requirements
that are designated for specific damage conditions in the existing
OTSGs.
The change also revises the inspection interval for 100%
inspections of OTSG tubes and the maximum interval for inspection of a
single OTSG consistent with Technical Specification Task Force item 449
for the Alloy 690 tube material in the replacement OTSGs. The revised
inspection requirements are based on properties and experience with the
improved Alloy 690 tube material. The revised inspection requirements
will result in the same outcome that OTSG tube integrity will continue
to be maintained.
This change continues to implement steam generator performance
criteria for tube structural integrity, accident induced leakage, and
operational leakage for the replacement OTSGs. Meeting the performance
criteria provides reasonable assurance that the replacement OTSG tubing
will remain capable of fulfilling its specific safety function of
maintaining reactor coolant pressure boundary integrity throughout each
operating cycle and in the unlikely event of a design basis accident.
The performance criteria are only a part of the OTSG program required
by the existing ITS [Improved Technical Specification]. The program,
defined by NEI [Nuclear Energy Institute] 97-06, Steam Generator
Program Guidelines, includes a framework that incorporates a balance of
prevention, inspection, evaluation, repair, and leakage monitoring.
These features will continue to be implemented as they are currently
approved. The proposed changes do not, therefore, significantly
increase the probability of an accident previously evaluated.
The consequences of design basis accidents are, in part, functions
of the DOSE EQUIVALENT I-131 in the primary coolant and the primary to
secondary LEAKAGE rates resulting from an accident. Therefore, limits
are included in the plant technical specifications for operational
leakage and for DOSE EQUIVALENT I-131 in the primary coolant to ensure
the plant is operated within its analyzed condition. The analysis of
the limiting design basis accident assumes that the primary to
secondary leak rate, after the accident, is 1 gallon per minute with no
more than 150 gallons per day in any one SG [steam generator], and that
the reactor coolant activity levels of DOSE EQUIVALENT I-131 are at the
TS [technical specification] values before the accident. The proposed
change to the OTSG inspection program does not affect the design of the
OTSGs, their method of operation, operational leakage limits, or
primary coolant chemistry controls. The proposed change does not
adversely impact any other previously evaluated design basis accident.
In addition, the proposed changes do not affect the consequences of a
Main Steam Line Break, rod ejection, or a reactor coolant pump locked
rotor event, or other previously evaluated accident. Therefore, the
proposed change does not affect the consequences of a Steam Generator
Tube Rupture accident and the probability of such an accident is
unchanged.
2. The Proposed Change Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed license amendment does not affect the design of the
OTSGs, their method of operation, or primary or secondary coolant
chemistry controls. In addition, the proposed amendment does not impact
any other plant system or component. The change modifies existing OTSG
inspection requirements for 100% inspection intervals, but establishes
inspection requirements that are considered equivalent based on
properties and experience with improved materials. Therefore, the
proposed change does not create the possibility of a new or different
type of accident from any accident previously evaluated.
3. The Proposed Change Does Not Involve a Significant Reduction in
the Margin of Safety.
The steam generator tubes in pressurized water reactors are an
integral part of the reactor coolant pressure boundary and, as such,
are relied upon to maintain the primary system's pressure and
inventory. As part of the reactor coolant pressure boundary, the steam
generator tubes are
[[Page 8285]]
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat can
be removed from the primary system. In addition, the steam generator
tubes isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a steam
generator is maintained by ensuring the integrity of its tubes. Steam
generator tube integrity is a function of the design, environment, and
the physical condition of the tube. The proposed change to the OTSG
inspection program does not affect tube design or operating
environment. The existing OTSG Program is maintained in this change.
The repair criteria that are being removed are specific to the existing
OTSGs and are not applicable to the replacement OTSGs. In the case of
the roll repair that is being removed, it potentially leads to
additional cracking over subsequent operating cycles due to tube cold
working during the re-roll. If tube defects are detected that exceed
limits in the new generators, then the tube will be removed from
service. This is considered a more effective means for removing defects
than repairs. For the above reasons, the margin of safety is not
changed and overall plant safety will be enhanced by the proposed
change to the ITS. Based upon the reasoning presented above and the
previous discussion of the amendment request, the requested 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: November 6, 2008.
Description of amendments request: The proposed change would revise
the Crystal River Unit 3 (CR-3) Improved Technical Specifications
Surveillance Requirements (SRs); SR 3.8.1.2, SR 3.8.1.6, and SR
3.8.1.10 to restrict the voltage and frequency limits for all Emergency
Diesel Generator (EDG) starts. The steady state voltage limits would be
revised to be more restrictive (plus or minus 2 percent of the nominal
voltage) to accurately reflect the appropriate calculation and the way
the plant is operated and tested. The steady state frequency limits
would be revised to be more restrictive (plus or minus 1 percent for
all EDG starts) to ensure compliance with the plant design bases and
the way the plant is operated. These changes would ensure that the EDGs
are capable of supplying power, with the correct voltage and frequency,
to the required electrical loads.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The LAR [license amendment request] proposes to provide more
restrictive steady state voltage and frequency limits for the Emergency
Diesel Generators (EDGs). The voltage band is going from a range of
greater than or equal to 3933 V [volts] but less than or equal to 4400
V, to greater than or equal to 4077 V but less than or equal to 4243 V.
The proposed limits are +/-2% [percent] around the nominal safety-
related bus voltage of 4160 V. The Frequency Limits are going from a 2%
tolerance band to a 1% tolerance band around the nominal frequency of
60 Hz [hertz] (59.4 Hz to 60.6 Hz) for all starts of the EDGs.
The EDGs are a safety-related system that functions to mitigate the
impact of an accident with a concurrent loss of offsite power. A loss
of offsite power is typically a significant contributor to postulated
plant risk and, as such, onsite AC [alternating current] generators
have to be maintained available and reliable in the event of a loss of
offsite power event. The EDGs are not initiators for any analyzed
accident, therefore; the probability for an accident that was
previously evaluated is not increased by this change. The revised,
voltage and frequency limits will ensure the EDGs will remain capable
of performing their design function.
The consequences of an accident refer to the impact on both plant
personnel and the public from any radiological release associated with
the accident. The EDG supports equipment that is supposed to preclude
any radiological release. More restrictive voltage and frequency limits
for the output of the EDG restores design margin, and provides
assurance that the equipment supplied by the EDG will operate correctly
and within the assumed timeframe to perform their mitigating functions.
Until the proposed CR-3 ITS [Improved Technical Specifications] EDG
voltage and frequency limits are approved by the NRC, administratively
controlled limits have been established in accordance with NRC
Administrative Letter 98-10 to ensure all EDG mitigation functions will
be performed, per design, in the event of a loss of offsite power.
These administrative limits have been determined as acceptable and have
been incorporated into the surveillance test procedures under the
provisions of 10 CFR 50.59. Periodic testing has been performed with
acceptable results. Since EDGs are mitigating components and are not
initiators for any analyzed accident, no increased probability of an
accident can occur. Since administrative limits will ensure the EDGs
will perform as designed, consequences will not be significantly
affected.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Administrative voltage limits were established using verified
design calculations and the guidance of NRC Administrative Letter 98-
10. These administrative limits will ensure the EDGs will perform as
designed. No new configuration is established by this change. The
administrative limits for the EDG frequency were determined to be
sufficient to account for measurement and other uncertainties.
The proposed amendment will place the administrative limits into
the CR-3 ITS. The more restrictive voltage and frequency limits will
provide additional assurance that the EDG can provide the necessary
power to supply the required safety-related loads during an analyzed
accident.
The proposed ITS voltage and frequency limits restore the EDG
capability to those analyzed by engineering calculation. No new
configuration is established. Therefore, no new or different kind of
accident from any previously evaluated can be created.
3. Does not involve a significant reduction in a margin of safety.
The LAR proposes to provide more restrictive steady state voltage
and frequency limits for the EDGs. The change in the acceptance
criteria for specific surveillance testing provides assurance that the
EDGs will be capable of performing their design function. Previous test
history has shown that the new limits are well within the
[[Page 8286]]
capability of the EDGs and are repeatable. The ``as-left'' settings for
voltage and frequency will be adjusted such that they remain within a
tight band and this ensures that the ``as-found'' settings will be in
an acceptable tolerance band.
The proposed ITS limits on voltage and frequency will ensure that
the EDG will be able to perform all design functions assumed in the
accident analyses. Administrative limits are in place to ensure these
parameters remain within analyzed limits. As such, the proposed change
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Thomas H. Boyce.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: September 26, 2008.
Description of amendment request: The amendments would revise the
Technical Specifications 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.'' In addition, the improvements to TSTF-448, Revision 3
as recommended in TSTF-508, Revision 0, ``Revise Control Room Envelope
Habitability Actions to Address Lessons Learned from TSTF-448
Implementation,'' have been incorporated as appropriate.
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 September 26, 2008.
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: Thomas H. Boyce.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: January 5, 2009.
Description of amendment request: The proposed amendment would
modify Technical Specifications (TS) requirements for mode change
limitations in accordance with Revision 9 of Nuclear Regulatory
Commission (NRC)-approved TS Task Force (TSTF) change TSTF-359,
``Increase Flexibility in Mode Restraints.''
In a Federal Register notice dated August 2, 2002 (67 FR 50475),
the NRC staff issued a notice of opportunity to comment on a model
safety evaluation and model no significant hazards consideration (NSHC)
determination for proposed license amendments adopting TSTF-359 using
the consolidated line item improvement process (CLIIP).
In a Federal Register notice dated April 4, 2003 (68 FR 16579), the
NRC staff issued a notice of availability of a model application for
proposed license amendments adopting TSTF-359 using the CLIIP. The
notice also included a revised model safety evaluation and a
[[Page 8287]]
model NSHC determination. In its application dated January 5, 2009, the
licensee affirmed the applicability of the model NSHC determination
which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC 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 allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO [Limiting Condition for Operation] 3.0.4,
are no different than the consequences of an accident while entering
and relying on the required actions while starting in a condition of
applicability of the TS. Therefore, the consequences of an accident
previously evaluated are not significantly affected by this change.
The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve the physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS Limiting
Conditions for Operation (LCO). The risk associated with this
allowance is managed by the imposition of required actions that must
be performed within the prescribed completion times. The net effect
of being in a TS condition on the margin of safety is not considered
significant. The proposed change does not alter the required actions
or completion times of the TS. The proposed change allows TS
conditions to be entered, and the associated required actions and
completion times to be used in new circumstances. This use is
predicated upon the licensee's performance of a risk assessment and
the management of plant risk. The change also eliminates current
allowances for utilizing required actions and completion times in
similar circumstances, without assessing and managing risk. The new
change to the margin of safety is insignificant. Therefore, this
change does not involve a significant reduction in a margin of
safety.
Based upon the reasoning presented above 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: January 5, 2009.
Description of amendment request: The proposed amendments would
delete Section 2.F of the Facility Operating License (FOL) for Hope
Creek Generating Station (Hope Creek) and Section 2.I of the FOL for
Salem Nuclear Generating Station (Salem) Unit No. 2. The FOL sections
being deleted require reporting of violations of the requirements in
Section 2.C of the respective FOLs. The proposed amendments would also
delete Technical Specification (TS) 6.9.3 for Hope Creek, Salem Unit
No. 1 and Salem Unit No. 2. These TSs contain a reporting requirement
that is duplicative of Nuclear Regulatory Commission (NRC) regulations.
The NRC staff issued a ``Notice of Opportunity to Comment on Model
Safety Evaluation on Elimination of Typical License Condition Requiring
Reporting of Violations of Section 2.C of Operating Licensing Using the
Consolidated Line Item Improvement Process,'' in the Federal Register
on August 29, 2005 (70 FR 51098). The notice included a model safety
evaluation (SE) and a model no significant hazards consideration (NSHC)
determination. On November 4, 2005, the NRC staff issued a notice in
the Federal Register (70 FR 67202) announcing that the model SE and
model NSHC determination may be referenced in plant-specific
applications to adopt the changes. In its application dated January 5,
2009, the licensee affirmed the applicability of the model NSHC
determination which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves the deletion of a reporting
requirement. The change does not affect plant equipment or operating
practices and therefore does not significantly increase the probability
or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed change is administrative in that it deletes a
reporting requirement. The change does not add new plant equipment,
change existing plant equipment, or affect the operating practices of
the facility. Therefore, the 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 deletes a reporting requirement. The change
does not affect plant equipment or operating practices and therefore
does not involve a significant reduction in a margin of safety.
Based on the above, the NRC staff proposes that the change presents
no significant hazards consideration under the standards set forth in
10 CFR 50.92(c).
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
[[Page 8288]]
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant (BFN), Units 1, 2 and 3, Limestone County,
Alabama
Date of amendment request: October 30 and November 20, 2008 (TS-
463-T).
Description of amendment request: The BFN requests adoption of an
approved change to the Standard Technical Specifications (TSs) for
General Electric Plants (NUREG-1433, BWR/4) and plant-specific TSs,
that allows: (1) Revising the frequency of Surveillance Requirement
(SR) 3.1.3.2, notch testing of fully withdrawn control rod, from ``7
days after the control rod is withdrawn and THERMAL POWER is greater
than the low-power set point (LPSP) of rod worth minimizer (RWM)'' to
``31 days after the control rod is withdrawn and THERMAL POWER is
greater than the LPSP of the RWM,'' (2) adding the word ``fully'' to
Limiting Condition for Operation LCO 3.3.1.2, Required Action E.2 to
clarify the requirement to fully insert all insertable control rods in
core cells containing one or more fuel assemblies when the associated
source range monitor instrument is inoperable, and (3) revising Example
1.4-3 in Section 1.4 ``Frequency'' to clarify that the 1.25
surveillance test interval extension in SR 3.0.2 is applicable to time
periods discussed in NOTES in the ``SURVEILLANCE'' column in addition
to the time periods in the ``FREQUENCY'' column.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No
This change does not affect either the design or operation of the
Control Rod Drive Mechanism (CRDM). The affected surveillance and
Required Action is not considered to be an initiator of any analyzed
event. Revising the frequency for notch testing fully withdrawn control
rods will not affect the ability of the control rods to shutdown the
reactor if required. Given the extremely reliable nature of the CRDM,
as demonstrated through industry operating experience, the proposed
monthly notch testing of all withdrawn control rods continues to
provide a high level of confidence in control rod operability. Hence,
the overall intent of the notch testing surveillances, which is to
detect either random stuck control rods or identify generic concerns
affecting control rod operability, is not significantly affected by the
proposed change. Requiring control rods to be fully inserted when the
associated SRM is inoperable is consistent with other similar
requirements and will increase the shutdown margin. The clarification
of Example 1.4-3 in Section 1.4 ``Frequency'' is an editorial change
made to provide consistency with other TSTF-475, Rev. 1 discussions in
Section 1.4. Therefore, the proposed changes do 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
Revising the frequency for notch testing fully withdrawn control
rods does not involve physical modification to the plant and does not
introduce a new mode of operation. Requiring control rods to be fully
inserted will make this action consistent with other similar actions.
The clarification of Example 1.4-3 in Section 1.4 ``Frequency'' is an
editorial change made to provide consistency with other discussions in
Section 1.4. 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 change involve a significant reduction in a
margin of safety?
Response: No
The CRDs and CRDMs are extremely reliable systems and, as such,
reducing the number of control rod notch tests will not significantly
impact the likelihood of detecting a stuck control rod. If a stuck
control rod is detected, existing action requirements will ensure
prompt action is taken to ensure there is not a generic problem. Other
surveillances are routinely performed to ensure that the performance of
the control rods in the event of a DBA [design-basis accident] or
transient meets the assumptions used in the safety analyses. As such,
potential effects of reducing the number of notch tests are far
outweighed by the benefit of reducing undue burden on reactor operators
and reducing the potential for mispositioning events which accompanies
any control rod manipulation. Requiring control rods to be fully
inserted instead of partially inserted when the associated SRM is
inoperable will increase the margin of safety. The clarification of
Example 1.4-3 in Section 1.4 ``Frequency'' is an editorial change made
to provide consistency with other discussions in Section 1.4.
Therefore, the proposed changes do not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Thomas H. Boyce.
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.
Duke Power Company LLC, Docket Nos. 50-414, Catawba Nuclear Station,
Unit 2, York County, South Carolina
Date of application for amendments: January 20, 2009.
Brief description of amendments: The proposed amendment would allow
a one-time limited duration extension of the Technical Specification
(TS) Surveillance (SR) 3.3.1.4 frequency. SR 3.3.1.4 is a Trip
Actuating Device Operational Test (TADOT) of the reactor trip breakers
(RTBs) and reactor trip bypass breakers.
Date of publication of individual notice in Federal Register:
January 28, 2009 (74 FR 4986).
Expiration date of individual notice: 30 days February 27, 2009; 60
days March 30, 2009.
[[Page 8289]]
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 Systems (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].
Carolina Power & Light Company, et. al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: January 4, 2008.
Brief description of amendment: The amendment establishes more
effective and appropriate action, surveillance, and administrative
requirements related to ensuring the habitability of the control room
envelope in accordance with the NRC-approved Technical Specification
Task Force (TSTF) Standard Technical Specification change traveler
TSTF-448, Revision 3, ``Control Room Habitability.'' This technical
specification improvement was initially made available in the Federal
Registerby the NRC on January 17, 2007 (72 FR 2022).
Date of issuance: January 29, 2009.
Effective date: Effective as of the date of issuance and shall be
implemented within 180 days.
Amendment No: 128.
Renewed Facility Operating License No. NPF-63: The amendment
revises the Technical Specifications and Facility Operating License.
Date of initial notice in Federal Register: May 20, 2008 (73 FR
29161).
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated January 29, 2009.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, et. al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: April 3, 2008, as supplemented
by letters dated December 9, 2008, and January 9, 2009.
Brief description of amendment: The amendment revises Technical
Specification Section 5.6.3.b to allow a reconfiguration of the fuel
racks in Spent Fuel Pool (SFP) C and allow the use of Metamic as an
alternate neutron poison material in the new storage racks for SFP C
and D. The amendment: (1) Revises the rack configuration in SFP C to
allow the substitution of four previously approved (13 x 13 cell)
Boiling Water Reactor racks with an equal number of (9 x 9 cell)
Pressurized Water Reactor racks, and (2) authorizes the use of Metamic
as an alternate spent fuel rack poison material.
Date of issuance: January 29, 2009.
Effective date: Effective as of the date of issuance and shall be
implemented within 60 days.
Amendment No: 129.
Renewed Facility Operating License No. NPF-63: The amendment
revises the Technical Specifications and Facility Operating License.
Date of initial notice in Federal Register: June 10, 2008 (73 FR
32744). The supplemental letters provided clarifying information that
was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated January 29, 2009.
No significant hazards consideration comments received: No.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: January 22, 2008.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) requirements related to control room
envelope habitability in accordance with TS Task Force (TSTF) traveler
TSTF-448, ``Control Room Habitability,'' Revision 3. This TS
improvement was made available by the Commission on January 17, 2007
(72 FR 2022) as part of the consolidated line item improvement process
(CLIIP).
Date of issuance: January 30, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 249 and 229.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: March 25, 2008 (73 FR
15784).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 30, 2009.
No significant hazards consideration comments received: No
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: September 22, 2008.
Brief description of amendment: The amendment revised the Technical
Specification (TS) to change requirements related to Battery Systems
specified in TS Section 3.10 resulting in
[[Page 8290]]
removing the Limiting Condition for Operation pertaining to 345 kV
switchyard batteries, chargers and associated direct current
distribution panel.
Date of Issuance: February 11, 2009.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 234.
Facility Operating License No. DPR-28: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: November 18, 2008 (73
FR 68454).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 30, 2009.
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: January 2, 2008, as supplemented by
letter dated January 22, 2009.
Brief description of amendment: The amendment revised the actions
for inoperable containment isolation valves (CIVs) in Technical
Specification 3/4.6.3, ``Containment Isolation Valves,'' to increase
the allowed outage time from 4 hours to 72 hours for inoperable CIVs
for penetrations with closed systems inside containment.
Date of issuance: January 30, 2009.
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 217.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 29, 2008 (73 FR
5219). The supplemental letter dated January 22, 2009, 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 January 30, 2009.
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: February 21, 2008.
Brief description of amendment: The amendments approved revisions
to the current licensing basis for Braidwood and Byron associated with
the application of an alternative source term (AST) methodology,
previously approved by the Nuclear Regulatory Commission staff.
Specifically, the amendments approved removing credit for the control
room ventilation system recirculation prefilters and reducing the
assumed control room unfiltered inleakage in the AST analyses.
Date of issuance: February 5, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: Braidwood Unit 1-155; Braidwood Unit 2-155; Byron
Unit No. 1-160; and Byron Unit No. 2-160.
Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and NPF-66:
The amendments revised the current licensing basis for Braidwood and
Byron.
Date of initial notice in Federal Register: June 3, 2008 (73 FR
31720).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 5, 2009.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: January 23, 2008.
Brief description of amendment: The proposed amendment would extend
the pressure temperature (PT) limit curves and the low temperature
overpressure protection (LTOP) setpoints for operation to 55 Effective
Full Power Years (EFPYs). The current PT limit curves (and the LTOP
setpoints) are applicable to 21.7 EFPYs. The new PT limits and LTOP
settings will be applicable to 60 calendar years, which includes the
period until the end of the renewed operating license.
Date of Issuance: January 29, 2009.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 154.
Renewed Facility Operating License No. NPF-16: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: September 9, 2008 (73
FR 52418).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 29, 2009.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: February 6, 2008, as
supplemented on September 16 and November 6, 2008.
Brief description of amendment: The amendment approved the
installation and use of the General Electric--Hitachi nuclear
measurement analysis and control digital Power Range Neutron Monitoring
System (PRNMS), and approved changes in the Technical Specifications to
reflect use of the PRNMS at Monticello Nuclear Generating Plant.
Date of issuance: January 30, 2009.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 159.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications and Facility Operating License.
Date of initial notice in Federal Register: March 11, 2008 (73 FR
13025).
The supplemental letters contained clarifying information, did not
change the initial no significant hazards consideration determination,
and did not expand the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 30, 2009.
No significant hazards consideration comments received: No.
Southern California Edison Company, et. al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: June 27, 2008.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to adopt Technical Specification Task
Force (TSTF) Change Traveler TSTF-487, Revision 1, ``Relocate DNB
[Departure from Nucleate Boiling] Parameters to the COLR [Core
Operating Limits Report].'' Specifically, the amendments revised TS
3.4.1 and its associated bases and TS
[[Page 8291]]
5.7.1.5 to replace the DNB numeric limits in TSs with references to the
COLR.
Date of issuance: February 3, 2009.
Effective date: As of its date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 2-219; Unit 3-212.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: September 23, 2008 (73
FR 54868).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 3, 2009.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: January 23, 2008.
Brief description of amendments: The amendments revised the actions
specified in Technical Specification (TS) 3.6.1.3, ``Containment Air
Locks,'' when limiting condition for operation (LCO) 3.6.1.3 is not
met. The amendments allow plant personnel to repair containment air
lock components while the plant remains at power and ensure that the
containment air locks will continue to meet the requirements of the
design basis.
Date of issuance: January 30, 2009.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1-190; Unit 2-178.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: March 25, 2008 (73 FR
15788).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 30, 2009.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: July 10, 2008, as supplemented by letter
dated August 26, 2008.
Brief description of amendment: The amendment modified Technical
Specification (TS) 5.5.6 consistent with the Technical Specification
Task Force (TSTF) Standard Technical Specification Change Traveler,
TSTF-419, Revision 0, ``Revise PTLR [Pressure and Temperature Limits
Report] Definition and References in ISTS [Improved Standard TS] 5.6.6,
RCS [Reactor Coolant System] PTLR.'' The revised TS 5.6.6 references
only the Topical Report (TR) number and title in TS 5.6.6, ``Reactor
Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR).''
This allows the use of the currently approved TRs to determine the
pressure and temperature limits in the PTLR without having to submit an
amendment to the Operating License. The change does not alter (1) the
U.S. Nuclear Regulatory Commission (NRC) reviewed and approved
analytical methods used to determine the pressure and temperature
limits or Low Temperature Overpressure Protection System setpoints, or
(2) the requirement to use NRC-approved analytical methods to determine
the limits or setpoints.
Date of issuance: January 27, 2009.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 180.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Renewed Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 26, 2008 (73 FR
50362). The supplemental letter dated August 26, 2008, 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 January 27, 2009.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
[[Page 8292]]
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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 51.12(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
application for amendment, (2) the amendment to Facility Operating
License, 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's (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].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. 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 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland, and electronically on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. 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 petitioner/requestor
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
petitioner/requestor 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 petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the 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.\1\ 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 petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
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\1\ To the extent that the application contains attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel to discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
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
[[Page 8293]]
under 10 CFR 2.315(c), must be filed in accordance with the NRC E-
Filing rule, which the NRC promulgated in August 28, 2007 (72 FR
49139). The E-Filing process requires participants to submit and serve
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 a waiver in accordance with
the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
[email protected], or by calling (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/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM
to access the Electronic Information Exchange (EIE), a component of the
E-Filing system. The Workplace Forms ViewerTM is free and is
available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. 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.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it 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 filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE 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
EIE 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 may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC electronic filing
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time,
Monday through Friday, excluding government holidays. The electronic
filing Help Desk can be contacted by telephone at 1-866-672-7640 or by
e-mail at [email protected].
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, 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.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the petition
and/or request should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii).
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://www.ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded
pursuant to an order of the Commission, an Atomic Safety and Licensing
Board, or a 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.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station (KPS), Kewaunee County, Wisconsin
Date of amendment request: January 23, 2009, as supplemented by
letters of January 26, January 30 and February 5, 2009.
Description of amendment request: The amendment revised the KPS
facility operating license by modifying the Technical Specifications in
Section 3.7.a.7 from ``The two underground storage tanks combine to
supply at least 35,000 gallons of fuel oil for either diesel generator
and the day tanks for each diesel generator contain at least 1,000
gallons of fuel oil'' to require each diesel generator's underground
storage tank and corresponding day tanks to contain a minimum useable
volume of 32,888 gallons.
Date of issuance: February 6, 2009.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 203.
Facility Operating License No. DPR-43: Amendment revised Facility
Operating License No. DPR-43 and Appendix A of the Technical
Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. The Nuclear Regulatory Commission (NRC)
staff published a public notice of the proposed amendment, issued a
proposed finding of NSHC, and requested that any comments on the
proposed NSHC be provided to the NRC staff no later than close of
business on February 5, 2009. The notice was published in the ``Herald
Times Reporter'' of Manitowoc, Wisconsin, on January 29, 2009. No
comments have been received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated February 6, 2009.
[[Page 8294]]
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc.,
120 Tredegar Street, Richmond, VA 23219.
NRC Branch Chief: Lois M. James.
Dated at Rockville, Maryland, this 12th day of February 2009.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation. 57
[FR Doc. E9-3515 Filed 2-23-09; 8:45 am]
BILLING CODE 7590-01-P