[Federal Register Volume 73, Number 204 (Tuesday, October 21, 2008)]
[Notices]
[Pages 62560-62574]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-24896]
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NUCLEAR REGULATORY COMMISSION
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 September 25, 2008 to October 8, 2008. The
last biweekly notice was published on October 7, 2008 (73 FR 58669).
[[Page 62561]]
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, 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 delivered to Room 6D44, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
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. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice,
person(s) may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
via electronic submission through the NRC E-Filing system for a hearing
and a petition for leave to intervene. 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 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 supports 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.
A request for hearing or a petition for leave to intervene must be
filed in
[[Page 62562]]
accordance with the NRC E-Filing rule, which the NRC promulgated on
August 28, 2007 (72 FR 49139). The E-Filing process requires
participants to submit and serve 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 technical help line,
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time,
Monday through Friday. The help line number is (800) 397-4209 or
locally, (301) 415-4737.
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). To be timely, filings must be submitted no later
than 11:59 p.m. Eastern Time on the due date.
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, 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. 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].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: August 27, 2008.
Description of amendments request: The amendment would change the
containment buffering agent from trisodium phosphate (TSP) to sodium
tetraborate in order to minimize the potential for sump screen blockage
due to potential adverse chemical interactions between TSP and certain
insulation materials used in containment under post loss-of-coolant
accident conditions. This amendment is one of the remaining
modifications required for Calvert Cliffs Nuclear Power Plant, Unit
Nos. 1 and 2 to achieve full compliance with the requirements of
Generic Letter 2004-02, ``Potential Impact of Debris Blockage on
Emergency Recirculation During Design Basis Accidents at Pressurized-
Water Reactors'' (Agencywide Documents Access and Management System
(ADAMS) Accession Number ML042360586).
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?
[[Page 62563]]
Response-No.
The proposed amendment does not involve a significant increase
in the probability of an accident previously evaluated because the
containment buffering agent is not an initiator of any analyzed
accident. The proposed change does not impact any failure modes that
could lead to an accident. The proposed amendment does not involve a
significant increase in the consequences of an accident previously
evaluated. The buffering agent in Containment is designed to buffer
the acids expected to be produced after a loss-of-coolant accident
(LOCA) and is credited in the radiological analysis for iodine
retention. Utilizing the required quantity of sodium tetraborate
decahydrate (STB) as a buffering agent ensures the post-LOCA
containment sump mixture will have a pH >= 7.0. The proposed change
of replacing trisodium phosphate (TSP) with STB results in the
radiological consequences remaining within the limits of 10 CFR
50.67. There is no dose change with the pH >= 7.0.
Therefore, operation of the facility in accordance with the
proposed amendment would 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 amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The STB is a passive component that is proposed to be
used as a buffering agent to increase the pH of the initially acidic
post-LOCA containment water to a more neutral pH. Changing the
proposed buffering agent from TSP to STB does not constitute an
accident initiator or create a new or different kind of accident
than previously analyzed. The proposed amendment does not involve
operation of any required systems, structures, or components in a
manner or configuration different from those previously recognized
or evaluated. No new failure mechanisms will be introduced by the
changes being requested. Therefore, the proposed amendment does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response-No.
The proposed amendment does not involve a significant reduction
in a margin of safety. The proposed amendment of changing the
buffering agent from TSP to STB results in equivalent control of
maintaining sump pH at >= 7.0, thereby controlling containment
atmosphere iodine and ensuring the radiological consequences of a
LOCA are within regulatory limits. The change of buffering agent
from TSP to STB also reduces the amount of calcium phosphate
precipitate generated thereby reducing the overall amount of
precipitate that may be formed in a postulated LOCA. The buffer
change would minimize the potential chemical effects and should
enhance the ability of the Emergency Core Cooling System to perform
the post-LOCA mitigating functions.
Therefore, the proposed amendment does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group LLC, 750 East Pratt Street,
17th Floor, Baltimore, MD 21202.
NRC Branch Chief: Mark G. Kowal.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: July 21, 2008.
Description of amendment request: The amendment proposes a change
to the Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specifications
(TSs) to support adoption of Technical Specification Task Force (TSTF)
359, ``Increased Flexibility in Mode Restraints.'' The NRC approved
adoption of TSTF-359 for ANO-1 in TS Amendment 232. The overall intent
of TSTF-359 was to eliminate exceptions to Limiting Condition for
Operation (LCO) 3.0.4 within individual specifications and provide
requirements within LCO 3.0.4 to control mode changes when TS-required
equipment is inoperable. Following implementation of TS Amendment 232,
Entergy discovered that one of the marked-up TS pages which contained
an LCO 3.0.4 exception was not provided to the NRC for review in the
original submittal.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 2, 2002 (67 FR 50475), as part of the
Consolidated Line Item Improvement Process (CLIIP), on possible
amendments to revise the plant-specific TS to modify requirements for
model change limitations in LCO 3.0.4 and SR 3.0.4.
The NRC staff subsequently issued a notice of availability of the
models for Safety Evaluation and No Significant Hazards Consideration
Determination for referencing in license amendment applications in the
Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed
the applicability of the CLIIP, including the model No Significant
Hazards Consideration Determination, in its application dated October
22, 2007.
The proposed TS changes are consistent with NRC-approved Industry
TSTF STS change, TSTF-359, Revision 8, as modified by 68 FR 16579.
TSTF-359, Revision 8, was subsequently revised to incorporate the
modifications discussed in the April 4, 2003, Federal Register notice
and other minor changes. TSTF-359, Revision 9, was subsequently
submitted to the NRC on April 28, 2003, and was approved by the NRC on
May 9, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the NRC staff analysis
of the issue of no significant hazards consideration is presented
below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Response: No.
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 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.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Response: No.
The proposed change does not involve a 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. Therefore, this
change does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
Response: No.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
[[Page 62564]]
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 net
change to the margin of safety is insignificant. Therefore, this
change does not involve a significant reduction in a margin of
safety.
The NRC staff proposes to determine that the request for amendment
involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: December 13, 2007.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) Section 4.3.1,
``Criticality,'' to add a new requirement to use a blocking device in
spent fuel storage rack cells that cannot maintain the effective
neutron multiplication factor, Keff, requirements specified
in TS Section 4.3.1.1.a. In addition, the proposed change revises TS
Section 4.3.3 to reflect that the LaSalle County Station, Unit 2 spent
fuel storage capacity is limited to no more than a combination of 4078
fuel assemblies and blocking devices.
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 adds an additional requirement to the TS to
ensure that the effective neutron multiplication factor
Keff, is less than or equal to 0.95, if fully flooded
with borated water. The additional requirement is to insert a
blocking device into unusable storage rack cell locations. Since the
proposed change pertains only to the spent fuel pool (SFP), only
those accidents that are related to movement and storage of fuel
assemblies in the SFP could be potentially affected by the proposed
change.
The probability that a misplaced fuel assembly would result in
an inadvertent criticality is unchanged since the process and
procedural controls governing fuel cell movement in the SFP will not
be changed. The current criticality analysis for the LSCS Unit 2 SFP
credits the neutron absorbing properties of the Boraflex neutron
poison material in the spent fuel storage racks. The current
analysis demonstrates: (1) Adequate margin to criticality for all
spent fuel storage cells, (2) adequate margin for fuel assemblies
inadvertently placed into locations adjacent to the spent fuel
racks, and (3) adequate margin for assemblies accidentally dropped
onto the spent fuel racks. The dose consequences of the most
limiting drop of a fuel assembly in the spent fuel pool is limited
by the number of the fuel rods damaged and other engineered features
unaffected by the proposed change, including the fuel design, fuel
decay time, water level in the spent fuel pool, water temperature of
the spent fuel pool, and the engineering features of the Reactor
Building Ventilation System.
The revised analysis does not result in a significant increase
in the probability of an accident previously analyzed. The revised
analysis takes no credit for the Boraflex material. The use of a
blocking device prevents an inadvertent action to insert a spent
fuel assembly, and prevents an assembly that is accidentally dropped
to penetrate into the empty spent fuel cell. In addition to this
blocking device, administrative controls will be implemented to
prevent insertion of a bundle into a cell that is blocked. The
probability that a fuel assembly would be inadvertently placed into
a location adjacent to the racks is unchanged, and the probability
that a fuel assembly would be dropped is unchanged by the revised
analysis. These events involve failures of administrative controls,
human performance, and equipment failures that are unaffected by the
presence or absence of Boraflex and the blocking devices.
The revised analysis does not result in a significant increase
in the consequence of an accident previously analyzed. The revised
analysis demonstrates adequate margin to criticality for unblocked
cells in the LSCS Unit 2 SFP, adequate margin for assemblies
inadvertently placed into locations adjacent to the spent fuel
racks, and adequate margin for assemblies accidentally dropped onto
the spent fuel racks. Placing a spent fuel assembly into a location
containing a blocking device is not a credible event since there are
diverse and redundant administrative and physical barriers to
prevent that.
The revised analysis does not affect the consequences of a
dropped fuel assembly. The consequences of dropping a fuel assembly
onto any other fuel assembly or other structure, other than a
blocking device, are unaffected by the change. The consequences of
dropping a fuel assembly onto a blocking device are bounded by the
event of dropping an assembly onto another assembly, both for
criticality and for radiological consequences. For criticality, the
blocking device prevents the dropped assembly from entering the
blocked cell. For radiological consequences, the number of rods
damaged when a fuel assembly is accidentally dropped onto a blocking
device is bounded the by the number of rods damaged by an assembly
dropped onto another assembly. The change does not affect the
effectiveness of the other engineered design features to limit the
offsite dose consequences of the limiting fuel assembly drop
accident.
The proposed change to clarify that the capacity of the Unit 2
SFP is limited to no more than a combination of 4078 fuel assemblies
and blocking devices does not affect the probability or consequences
of an accident previously analyzed because no physical modifications
to the storage racks are proposed. The proposed change will reduce
the number of allowable fuel assembly storage locations.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Onsite storage of spent fuel assemblies in the SFP is a normal
activity for which LSCS has been designed and licensed. As part of
assuring that this normal activity can be performed without
endangering public health and safety, the ability to safely
accommodate different possible accidents in the SFP, such as
dropping a fuel assembly or misloading a fuel assembly, have been
analyzed. The proposed fuel storage configuration does not change
the methods of fuel movement or fuel storage. No structural or
mechanical change to the racks or fuel handling equipment is being
proposed. The proposed change allows for partial use of storage rack
locations that have been determined unusable based on the existing
criticality analysis.
The blocking devices are passive devices. These devices, when
inside a spent fuel storage rack cell, perform the same function of
a spent fuel assembly in that cell. These devices do not add any
limiting structural loads or affect the removal of decay heat from
the other assemblies. The devices are resistant to corrosion and
will maintain their structural integrity over the life of the plant.
These devices are not under any structural load during normal
operations. They are only challenged by an accidental fuel assembly
drop. The existing fuel handling accident, which assumes the drop of
a fuel bundle, bounds the drop of a blocking device.
This change does not create the possibility of a misloaded
assembly into a blocked cell. Placing a spent fuel assembly into a
location containing a blocking device is not a credible event since
there are diverse and redundant administrative and physical barriers
to prevent that.
[[Page 62565]]
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.
LSCS TS 4.3.1 .1 requires the spent fuel storage racks to
maintain the effective neutron multiplication factor,
Keff, less than or equal to 0.95 when fully flooded with
unborated water, which includes an allowance for uncertainties.
Therefore, for criticality, the required safety margin is 5%
including a conservative margin to account for engineering
uncertainties.
The proposed change adds a requirement to use a blocking device
to ensure that Keff continues to be less than or equal to
0.95; thus, the required safety margin of 5% is preserved. The
proposed change also clarifies that the capacity of the Unit 2 SFP
is limited to no more than a combination of 4078 fuel assemblies and
blocking devices. This clarification does not impact the required
safety margin of 5%.
The current analysis assumes an infinite array of fuel with all
fuel at the peak reactivity (i.e., the highest combination of
initial enrichment, gadolinium, and fuel burnup that maximizes the
reactivity of the fuel). The revised analysis demonstrates the same
margin to criticality of 5%, including a conservative margin to
account for engineering uncertainties, is maintained assuming an
infinite array of fuel with all fuel at the peak reactivity. In
addition, the margin of safety for radiological consequences of a
dropped fuel assembly are unchanged because the event involving a
dropped fuel assembly onto a blocking device is bounded by the
consequences of a dropped fuel assembly onto another fuel assembly.
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
requested amendments involve 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: Russell Gibbs.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: May 30, 2008, as supplemented on July 17
and September 10, 2008.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TS) Table 3.3.8.1-1, ``Loss of Power
Instrumentation,'' specifically to change the maximum allowable voltage
of the 4.16-kV Emergency Bus Undervoltage function from less-than-or-
equal to 3899 V to less-than-or-equal-to 3822 V.
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 TS change to the maximum allowable voltage for the
4160 volt Emergency Bus Undervoltage relays affects when an
Emergency Bus that is experiencing degraded voltage will disconnect
from offsite power and transfer to an emergency diesel generator.
While the maximum allowed voltage that initiates this action will be
lowered, the function remains the same. The maximum allowed voltage
has been analyzed to ensure spurious trips will be avoided. The
proposed change will not affect any accident initiators or
precursors. As a result, the probability of any accident previously
evaluated is not significantly increased.
The consequences of any accident previously evaluated are not
increased since the 4160 volt Emergency Bus Undervoltage relays will
continue to meet their required function to transfer the 4160 volt
Emergency Buses to the emergency diesel generators in the event of a
degraded voltage condition on the offsite power supply. This
transfer will ensure that the electrical equipment is capable of
performing its function to meet the requirements of the accident
analyses.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The proposed TS change to the maximum allowable voltage for
the 4160 volt Emergency Bus Undervoltage relays does not affect
existing or introduce any new accident precursors or modes of
operation. The relays will continue to detect undervoltage
conditions and transfer the Emergency Buses to the emergency diesel
generators at a voltage adequate to ensure proper safety equipment
performance and to prevent equipment damage. The function of the
relays remains the same.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed TS change to the maximum allowable voltage for the
4160 volt Emergency Bus Undervoltage relays will allow all safety
loads to have sufficient voltage to perform their intended safety
functions while ensuring spurious trips are avoided. Thus, the
results of the accident analyses will not be affected as the input
assumptions are protected.
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: Mr. R. E. Helfrich, Florida Power & Light
Company, P. O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Lois M. James.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: August 19, 2008.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) requirements for mode change
limitations in accordance with NRC-approved TS Task Force (TSTF)
traveler TSTF-359, Revision 9, ``Increase Flexibility in MODE
Restraints,'' and revise TS Section 1.4, ``Frequency,'' in accordance
with NRC-approved traveler TSTF-485, Revision 0, ``Correct Example 1.4-
1.''
The NRC staff issued a ``Notice of Availability of Model
Application Concerning Technical Specification Improvement To Modify
Requirements Regarding Mode Change Limitations Using the Consolidated
Line Item Improvement Process'' in the Federal Register on April 4,
2003 (68 FR 16579). The notice referenced a model safety evaluation and
a model no significant hazards consideration (NSHC) determination
published in the Federal Register on August 2, 2002 (67 FR 50475). In
its application dated August 19, 2008, 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 adopted by the licensee regarding TSTF-359 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
[[Page 62566]]
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 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 a 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 net
change to the margin of safety is insignificant. Therefore, this
change does not involve a significant reduction in a margin of
safety.
In its application dated August 19, 2008, the licensee also
affirmed the applicability of the NSHC approved by the NRC in TSTF-485,
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 revises Section 1.4, Frequency, Example 1.4-
1, to be consistent with Surveillance Requirement (SR) 3.0.4 and
Limiting Condition for Operation (LCO) 3.0.4. This change is
considered administrative in that it modifies the example to
demonstrate the proper application of SR 3.0.4 and LCO 3.0.4. The
requirements of SR 3.0.4 and LCO 3.0.4 are clear and are clearly
explained in the associated Bases. As a result, modifying the
example will not result in a change in usage of the Technical
Specifications (TS). The proposed change does not adversely affect
accident initiators or precursors, 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, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. Therefore, this change is considered administrative and
will have no effect on the probability or consequences of any
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.
No new or different accidents result from utilizing the proposed
change. The change 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. In addition, the change does not impose any new or
different requirements or eliminate any existing requirements. The
change does not alter assumptions made in the safety analysis. The
proposed change is 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.
The proposed change is administrative and will have no effect on
the application of the Technical Specification requirements.
Therefore, the margin of safety provided by the Technical
Specification requirements is unchanged.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based upon this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendment involves NSHC.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Michael T. Markley.
Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-220, Nine
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York
Date of amendment request: August 15, 2008.
Description of amendment request: The proposed amendment would
revise NMP1 Technical Specification (TS) 6.5.7, ``10 CFR 50 [Part 50 of
Title 10 of the Code of Federal Regulations] Appendix J Testing Program
Plan,'' to allow a one-time extension of the Integrated Leak Rate Test
(ILRT) interval for no more than five (5) years. The proposed amendment
would allow the next ILRT for NMP1 to be performed within 15 years from
the last ILRT as opposed to the current 10-year interval.
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 change involves a one-time extension of the primary
containment ILRT interval from 10 to 15 years. The proposed change
does not involve a physical change to the plant or a change in the
manner in which the plant is operated or controlled. The primary
containment function is to provide an essentially leak tight barrier
against the uncontrolled release of radioactivity to the environment
for postulated accidents. As such, the containment itself and the
testing requirements to periodically demonstrate the integrity of
the containment exist to ensure the plant's ability to mitigate the
consequences of an accident, and do not involve any accident
precursors or initiators. Therefore, the probability of occurrence
of an accident previously evaluated is not significantly increased
by the proposed change.
Continued containment integrity is assured by the established
programs for local leak rate testing and inservice/containment
[[Page 62567]]
inspections, which are unaffected by the proposed change. As
documented in NUREG-1493, ``Performance-Based Containment Leak-Test
Program,'' dated September 1995, industry experience has shown that
local leak rate tests (Type B and C) have identified the vast
majority of containment leakage paths, and that ILRTs detect only a
small fraction of containment leakage pathways.
The potential consequences of the proposed change have been
quantified by analyzing the changes in risk that would result from
extending the ILRT interval from 10 years to 15 years. The increase
in risk in terms of person-rem per year within 50 miles resulting
from design basis accidents was estimated to be of a magnitude that
NUREG-1493 indicates is imperceptible. NMPNS has also analyzed the
increase in risk in terms of the frequency of large early releases
from accidents. The increase in the large early release frequency
resulting from the proposed change was determined to be within the
guidelines published in NRC Regulatory Guide 1.174. Additionally,
the proposed change maintains defense-in-depth by preserving a
reasonable balance among prevention of core damage, prevention of
containment failure, and consequence mitigation. NMPNS has
determined that the increase in conditional containment failure
probability due to the proposed change would be insignificant.
Therefore, it is concluded that the proposed one-time extension of
the primary containment ILRT interval from 10 years to 15 years does
not significantly increase the consequences of an accident
previously evaluated.
Based on the above discussion, it is concluded that the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change involves a one-time extension of the primary
containment ILRT interval. The containment and the testing
requirements to periodically demonstrate the integrity of the
containment exist to ensure the plant's ability to mitigate the
consequences of an accident, and do not involve any accident
precursors or initiators. The proposed change does not involve a
physical change to the plant (i.e., no new or different type of
equipment will be installed) or a change in the manner in which the
plant is operated or controlled.
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 one-time extension of the primary containment ILRT
interval does not alter the manner in which safety limits, limiting
safety system setpoints, or limiting conditions for operation are
determined. The specific requirements and conditions of the 10 CFR
[Part] 50 Appendix J Testing Program Plan, as defined in the TS,
exist to ensure that the degree of primary containment structural
integrity and leak-tightness that is considered in the plant safety
analyses is maintained. The overall containment leakage rate limit
specified by the TS is maintained, and Type B and C containment
leakage tests will continue to be performed at the frequency
currently required by the TS.
NMP1 and industry experience strongly support the conclusion
that Type B and C testing detects a large percentage of containment
leakage paths and that the percentage of containment leakage paths
that are detected only by the ILRT is small. Containment inspections
performed in accordance with other plant programs serve to provide a
high degree of assurance that the containment will not degrade in a
manner that is detectable only by an ILRT. Additionally, the on-line
containment monitoring capability that is inherent to inerted
boiling[-]water reactor containments allows for the detection of
gross containment leakage that may develop during power operation.
This combination of factors ensures that the margin of safety that
is inherent in plant safety analyses is maintained. Furthermore, a
risk assessment using the current NMP1 Probabilistic Risk Assessment
interval events model concluded that extending the ILRT test
interval from 10 to 15 years results in a very small change to the
NMP1 risk profile.
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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Mark G. Kowal.
Nine Mile Point Nuclear Station, LLC (NMPNS), Docket No. 50-410, Nine
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York
Date of amendment request: August 14, 2008.
Description of amendment request: The proposed amendment would (1)
revise the NMP2 Technical Specification (TS) Surveillance Requirement
(SR) frequency in TS 3.1.3, ``Control Rod Operability,'' and (2) revise
Example 1.4-3 in TS Section 1.4, ``Frequency,'' to clarify the
applicability of the 1.25 surveillance test interval extension. The
proposed changes are consistent with Nuclear Regulatory Commission
(NRC)-approved Revision 1 to TS Task Force (TSTF) Change Traveler,
TSTF-475, ``Control Rod Notch Testing Frequency and SRM [Source Range
Monitor] Insert Control Rod Action.'' The availability of this TS
improvement was announced in the Federal Register on November 13, 2007
(72 FR 63943) as part of the consolidated line item improvement
process. The licensee affirmed the applicability of the model no
significant hazards consideration determination in its application.
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--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change generically implements TSTF-475, Revision 1,
``Control Rod Notch Testing Frequency and SRM Insert Control Rod
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and
NUREG-1434 (BWR/6) STS. The changes: (1) Revise TS testing frequency
for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod
OPERABILITY,'' (2) clarify the requirement to fully insert all
insertable control rods for the limiting condition for operation
(LCO) in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring
Instrumentation'' (NUREG-1434 only), and (3) revise Example 1.4-3 in
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25
surveillance test interval extension. The consequences of an
accident after adopting TSTF-475, Revision 1 are no different than
the consequences of an accident prior to adoption. 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 Accident Previously
Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
proposed change 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 analyzed. Thus, 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 [a] Margin of Safety
TSTF-475, Revision 1 will: (1) [revise the TS SR 3.1.3.2
frequency in TS 3.1.3, ``Control Rod OPERABILITY,'' (2) clarify the
requirement to fully insert all insertable control rods for the
limiting condition for operation (LCO) in TS 3.3.1.2, ``Source Range
Monitoring Instrumentation,'' and (3)] revise
[[Page 62568]]
Example 1.4-3 in Section 1.4 ``Frequency'' to clarify the
applicability of the 1.25 surveillance test interval extension. [The
GE Nuclear Energy Report, ``CRD Notching Surveillance Testing for
Limerick Generating Station,'' dated November 2006, concludes that
extending the control rod notch test interval from weekly to monthly
is not expected to impact the reliability of the scram system and
that the analysis supports the decision to change the surveillance
frequency.] Therefore, the proposed changes in TSTF-475, Revision 1
are acceptable and 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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Mark G. Kowal.
Nine Mile Point Nuclear Station, LLC (NMPNS) Docket No. 50-220, Nine
Mile Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York
Date of amendment request: August 18, 2008.
Description of amendment request: The proposed amendment would
revise the NMP1 Technical Specification (TS) Section 3/4.1.1, ``Control
Rod System,'' to increase the Surveillance Requirement (SR) frequency
associated with control rod exercising. The proposed change would
revise the required SR frequency from once each week to once every 31
days. The proposed change is consistent with Nuclear Regulatory
Commission (NRC)-approved Revision 1 to TS Task Force (TSTF) Change
Traveler, TSTF-475, ``Control Rod Notch Testing Frequency and SRM
[Source Range Monitor] Insert Control Rod Action,'' and NUREG-1433,
``Standard Technical Specifications General Electric Plants, BWR/4,''
Revision 3.1. The availability of the TS improvement was announced in
the Federal Register on November 13, 2007 (72 FR 63943) as part of the
consolidated line item improvement process. The licensee affirmed the
applicability of the model no significant hazards consideration
determination in its application.
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--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change generically implements TSTF-475, Revision 1,
``Control Rod Notch Testing Frequency and SRM Insert Control Rod
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and
NUREG-1434 (BWR/6) STS. The changes: (1) revise TS testing frequency
for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod
OPERABILITY,'' (2) clarify the requirement to fully insert all
insertable control rods for the limiting condition for operation
(LCO) in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring
Instrumentation'' (NUREG-1434 only), and (3) revise Example 1.4-3 in
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25
surveillance test interval extension. The consequences of an
accident after adopting TSTF-475, Revision 1 are no different than
the consequences of an accident prior to adoption. 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 Accident Previously
Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
proposed change 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 analyzed. Thus, 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 [a] Margin of Safety
TSTF-475, Revision 1 will: (1) [revise the TS SR 3.1.3.2
frequency in TS 3.1.3, ``Control Rod OPERABILITY,'' (2) clarify the
requirement to fully insert all insertable control rods for the
limiting condition for operation (LCO) in TS 3.3.1.2, ``Source Range
Monitoring Instrumentation,'' and (3)] revise Example 1.4-3 in
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25
surveillance test interval extension. [The GE Nuclear Energy Report,
``CRD Notching Surveillance Testing for Limerick Generating
Station,'' dated November 2006, concludes that extending the control
rod notch test interval from weekly to monthly is not expected to
impact the reliability of the scram system and that the analysis
supports the decision to change the surveillance frequency.]
Therefore, the proposed changes in TSTF-475, Revision 1 are
acceptable and 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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Mark G. Kowal.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: July 11, 2008.
Description of amendment request: The proposed amendments would
establish Conditions, Required Actions, and Completion Times in the
Prairie Island Nuclear Generating Plant, Units 1 and 2, Technical
Specifications (TSs) for the condition where one steam supply to the
turbine-driven auxiliary feedwater (AFW) pump is inoperable concurrent
with an inoperable motor-driven AFW train. The proposed amendments
would also make changes to the TSs that establish specific Actions for
when the turbine-driven AFW train is inoperable either (a) due solely
to one inoperable steam supply, or (b) due to reasons other than the
one inoperable steam supply.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 19, 2007 (72 FR 12845), on possible
amendments concerning the consolidated line item improvement process
(CLIIP), including a model safety evaluation and a model no significant
hazards consideration (NSHC) determination. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on July 17, 2007
(72 FR 39089), as part of the CLIIP. In its application dated July 11,
2008, the licensee affirmed the applicability of the following
determination.
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:
1. Does the proposed change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The Auxiliary/Emergency Feedwater (AFW/EFW) System is not an
initiator of any design basis accident or event, and therefore the
proposed changes do not increase the probability of any accident
previously evaluated. The proposed changes to address the condition
of one or two motor driven AFW/EFW trains inoperable and the turbine
[[Page 62569]]
driven AFW/EFW train inoperable due to one steam supply inoperable
do not change the response of the plant to any accidents.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not adversely
affect the ability of structures, systems, and components (SSCs) to
perform their intended safety function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of any accident previously evaluated.
Further, the proposed changes do not increase the types and amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures.
Therefore, the changes do not involve a significant increase in
the probability or consequences of any 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 changes do not result in a change in the manner in
which the AFW/EFW System provides plant protection. The AFW/EFW
System will continue to supply water to the steam generators to
remove decay heat and other residual heat by delivering at least the
minimum required flow rate to the steam generators. There are no
design changes associated with the proposed changes. The changes to
the Conditions and Required Actions do not change any existing
accident scenarios, nor create any new or different accident
scenarios.
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 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 proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by these changes. The proposed changes will not
result in plant operation in a configuration outside the design
basis.
Therefore, it is concluded that the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment requests
involve no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Lois M. James.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendments: September 4, 2008.
Brief description of amendments: The proposed amendment will delete
the Technical specification (TS) requirements related to hydrogen
recombiners and hydrogen monitors. Licensees were generally required to
implement upgrades as described in NUREG-0737, ``Clarification of TMI
[Three Mile Island] Action Plan Requirements,'' and Regulatory Guide
(RG) 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was an outcome of the
lessons learned from the accident that occurred at TMI, Unit 2.
Requirements related to combustible gas control were imposed by
Order for many facilities and were added to or included in the TSs for
nuclear power reactors currently licensed to operate. The revised 10
CFR 50.44, ``Standards for Combustible Gas Control System in Light-
Water-Cooled Power Reactors,'' eliminated the requirements for hydrogen
recombiners and relaxed safety classifications and licensee commitments
to certain design and qualification criteria for hydrogen and oxygen
monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on September
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the
model NSHC determination in its application dated September 4, 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to 17 approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
RG 1.97 Category 1, is intended for key variables that most directly
indicate the accomplishment of a safety function for design-basis
accident events. The hydrogen monitors no longer meet the definition
of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR
50.44, the Commission found that Category 3, as defined in RG 1.97,
is an appropriate categorization or the hydrogen monitors because
the monitors are required to diagnose the course of beyond design-
basis accidents. The regulatory requirements for the hydrogen
monitors can be relaxed without degrading the plant emergency
response. The emergency response, in this sense, refers to the
methodologies used in ascertaining the condition of the reactor
core, mitigating the consequences of an accident, assessing and
projecting offsite releases of radioactivity, and establishing
protective action recommendations to be communicated to offsite
authorities. Classification of the hydrogen monitors as Category 3,
and removal of the hydrogen monitors from TS will not prevent an
accident management strategy through the use of the SAMGs, the
emergency plan (EP), the emergency operating procedures (EOP), and
site survey monitoring that support modification of emergency plan
protective action recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TS, does not involve a
significant increase in the probability or the consequences of any
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 elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner
[[Page 62570]]
and hydrogen monitor equipment are not considered accident
precursors, nor does their existence or elimination have any adverse
impact on the pre-accident state of the reactor core or post
accident confinement of radionuclides within the containment
building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a
neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: L. Raghavan.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: January 14, 2008.
Description of amendment request: The proposed amendment would
modify the Technical Specification (TS) requirements related to control
room envelope habitability in accordance with TS Task Force (TSTF)
traveler TSTF-448-A, ``Control Room Habitability,'' Revision 3.
The NRC staff issued a ``Notice of Availability of Technical
Specification Improvement to Modify Requirements Regarding Control Room
Envelope Habitability Using the Consolidated Line Item Improvement
Process'' in the Federal Register on January 17, 2007 (72 FR 2022). The
notice referenced a model safety evaluation, a model no significant
hazards consideration (NSHC) determination, and a model license
amendment request published in the Federal Register on October 17, 2006
(71 FR 61075). In its application dated January 14, 2008, 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 adopted by the licensee 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 has reviewed the analysis adopted by the licensee
and, based upon this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendment involves NSHC.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the
[[Page 62571]]
action involved exigent circumstances. They are repeated here because
the biweekly notice lists all amendments issued or proposed to be
issued involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: July 30, 2008.
Description of amendment request: This amendment revises the Indian
Point Nuclear Generating Unit No. 2 Technical Specification 3.8.1,
Required Action A.4, to allow a one time extension to the completion
time for the loss of one offsite power circuit from 72 hours to 144
hours. This change will ensure that there is enough time for the failed
oil cooling pump on the station auxiliary transformer to be removed,
and for the new oil cooling pump to be installed and tested.
Date of publication of individual notice in Federal Register:
August 27, 2008.
Expiration date of individual notice: October 27, 2008.
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].
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: March 10, 2008, as supplemented by
letters dated June 30, 2008, and September 29, 2008.
Description of amendment request: The amendment revised the Oyster
Creek Technical Specifications (TSs) 3.3, ``Reactor Coolant.''
Specifically, the amendment relocated the pressure and temperature
limit curves to the licensee controlled document, ``Pressure and
Temperature Limits Report'' (PTLR). Additionally, the amendment
introduced supporting definitions and adds controls regarding the PTLR
to Section 6.0, ``Administrative Controls.''
Date of issuance: September 30, 2008.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 269.
Facility Operating License No. DPR-16: The amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: June 17, 2008 (73 FR
34339). The supplemental letters 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 initial proposed
no significant hazards determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
September 30, 2008.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423,
Millstone Power Station, Unit Nos. 2 and 3, New London County,
Connecticut
Date of application for amendment: August 15, 2007, as supplemented
on May 27, 2008, July 24, 2008, and September 3, 2008.
Brief description of amendment: The proposed amendment modified
Technical Specification (TS) 3.3.3.1, ``Radiation Monitoring,'' TS
3.4.6.1, ``Reactor Coolant System Leakage Detection Systems,'' and
Surveillance Requirements 4.4.6.1, ``Reactor Coolant System Leakage
Detection Systems.'' Specifically, the proposed amendment removed
credit for the gaseous radiation monitor for Reactor Coolant System
leakage detection. Improvements in nuclear fuel reliability over time
have resulted in the reduction of effectiveness of the monitors in
detecting very small leaks and very small changes in the leak rate. The
proposed change also addressed the condition when the remaining
monitoring systems are all inoperable.
Date of issuance: September 30, 2008.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 306 and 244.
Renewed Facility Operating License Nos. DPR-65 and NPF-49:
Amendment revised the License and Technical Specifications.
Date of initial notice in Federal Register: June 17, 2008 (73 FR
34341). The supplements dated May 27, 2008, July 24, 2008, and
September 3, 2008, clarified the application, did not expand the scope
of the application as originally noticed, 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 September 30, 2008.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: May 8, 2008, as supplemented by
letter dated August 14, 2008.
Brief description of amendment: This amendment request contains
sensitive unclassified non-safeguards
[[Page 62572]]
information. The changes allow for interim alternate steam generator
tube repair criterion, as specified in the Millstone Power Station,
Unit 3 (MPS3) technical specifications. The interim alternate repair
criterion is for the upcoming refueling outage and the subsequent
operating cycle. The amendment also adds three reporting criteria to
the MPS3 technical specifications for steam generator tube inspections.
Date of issuance: September 30, 2008.
Effective date: As of the date of issuance and shall be implemented
prior to Mode 5 startup.
Amendment No.: 245.
Renewed Facility Operating License No. NPF-49: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: July 8, 2008 (73 FR
39054). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 30, 2008.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413, Catawba Nuclear
Station, Unit 1, York County, South Carolina
Date of application for amendment: December 20, 2007.
Brief description of amendment: The amendment reflects the direct
transfer of the undivided ownership interest of the Saluda River
Electric Cooperation, Inc., in Catawba Nuclear Station, Unit 1, to Duke
Energy Carolinas, LLC, a current owner and operator, and the North
Carolina Electric Membership Corporation, a current owner.
Date of issuance: September 30, 2008.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 245.
Facility Operating License Nos. NPF-35: Amendment revised the
license.
Date of initial notice in Federal Register: July 21, 2008 (73 FR
42375). The supplement dated May 29, 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.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 25, 2008.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: July 26, 2007, as superseded by
application dated August 8, 2007, and as supplemented by letters dated
November 19, 2007, and June 5 and July 21, 2008.
Brief description of amendment: The amendment revises the
requirements of Technical Specification (TS) 3.3.5.2, ``Reactor Core
Isolation Cooling (RCIC) System Instrumentation,'' and TS 3.5.2, ``ECCS
[Emergency Core Cooling System]-Shutdown,'' to increase the Condensate
Storage Tank level.
Date of issuance: September 30, 2008.
Effective date: As of its date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 210.
Facility Operating License No. NPF-21: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 28, 2007 (72 FR
49572).
The supplements dated November 19, 2007, and June 5 and July 21,
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 September 30, 2008.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: April 22, 2008, as supplemented
by letters date July 2, July 22, and September 24, 2008.
Brief description of amendment: The amendment modified Technical
Specification (TS) 1.0, ``Definitions,'' Limiting Conditions for
Operation and Surveillance Requirement Applicability Section 3.4.9,
``RCS [Reactor Coolant System] Pressure and Temperature (P-T) Limits,''
and Section 5.0, ``Administrative Controls,'' to delete reference to
the pressure and temperature curves, and include reference to the
Pressure and Temperature Limits Report (PTLR). This change adopted the
methodology of SIR-05-044-A, ``Pressure-Temperature Limits Report
Methodology for Boiling Water Reactors,'' for preparation of the
pressure and temperature curves, and incorporated the guidance of TSTF-
419-A, ``Revise PTLR Definition and References in ISTS [Improved
Standard Technical Specifications] 5.6.6, RCS PTLR.''
Date of issuance: October 3, 2008.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 292.
Facility Operating License No. DPR-59: The amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: July 1, 2008 (73 FR
37503). The supplemental submissions dated July 2, July 22, and
September 24, 2008, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the NRC staff's original proposed no
significant hazards consideration determination as published in the
Federal Register. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated October 3, 2008.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station (Byron), Unit Nos. 1 and 2, Ogle County, Illinois
Date of application for amendment: June 17, 2008.
Brief description of amendment: The amendments revise Technical
Specification (TS) 5.5.9, ``Steam Generator (SG) Program,'' and TS
5.6.9, ``Steam Generator (SG) Tube Inspection Report.'' For TS 5.5.9,
the amendments incorporate a one-cycle interim alternate repair
criteria in the provisions for SG tube repair criteria during Byron,
Unit No. 2, refueling outage 14 and the subsequent operating cycle. For
TS 5.6.9, the amendments revise the current reporting requirements.
These changes only affect Byron, Unit No. 2; however, this action is
docketed for both Byron units because the TS are common to both units.
Date of issuance: October 1, 2008.
Effective date: As of the date of issuance and shall be implemented
prior to the return to service from Byron, Unit No. 2, fall 2008
Refueling Outage 14.
Amendment Nos.: Unit 1--158; Unit 2--158.
Facility Operating License Nos. NPF-37 and NPF-66: The amendment
revised the TSs and License.
[[Page 62573]]
Date of initial notice in Federal Register: August 5, 2008 (73 FR
45485).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 1, 2008.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: July 16, 2007, as supplemented
May 20 and August 26, 2008.
Brief description of amendments: Amendments modified the technical
specification requirements related to control room envelope
habitability in accordance with Technical Specification Task Force
(TSTF) Traveler TSTF-448, Revision 3, ``Control Room Habitability.''
Date of Issuance: September 30, 2008.
Effective Date: Unit 1--Amendment is effective as of the date of
its issuance and shall be implemented following implementation of the
Amendment No. 152, regarding Alternative Source Term and with the
completion of the installation and testing of the plant modifications
described in the licensee's application, including letters dated July
16, 2007, February 14, March 18, April 14, June 2, July 11, and August
13, 2008. Unit 2--This license amendment is effective as of the date of
its issuance and shall be implemented following implementation of
License Amendment No. 152.
Amendment Nos.: 205 and 153.
Renewed Facility Operating License Nos. DPR-67 and NPF-16:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 28, 2007 (72 FR
49578). The supplements dated May 20 and 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 amendments is contained
in a Safety Evaluation dated September 30, 2008.
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: July 16, 2007, as supplemented
by letters dated February 14, March 18, April 14, June 2, July 11, and
August 13, 2008.
Brief description of amendment: The amendment modifies the
facility's operating licensing bases to adopt the alternative source
term as allowed in 10 CFR 50.67, and as described in Regulatory Guide
1.183. The licensee revised the plant licensing basis through
reanalysis of the radiological consequences of the following Updated
Final Safety Analysis Report Chapter 15 accidents: Loss-of-Coolant
Accident, Fuel-Handling Accident, Main Steam Line Break, Steam
Generator Tube Rupture, Reactor Coolant Pump Shaft Seizure, Control
Element Assembly Ejection, Letdown Line Break, and Feedwater Line
Break.
Date of issuance: September 29, 2008.
Effective date: Effective as of the date of issuance and shall be
implemented within 180 days.
Amendment No.: 152.
Renewed Facility Operating License No. NPF-16: The amendment
revises the Technical Specifications and the Renewed Facility Operating
License.
Date of initial notice in Federal Register: June 12, 2008 (73 FR
33460). The supplements dated February 14, March 18, April 14, June 2,
July 11, and August 13, 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.
Public comments received as to proposed no significant hazards
consideration (NSHC): No.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 29, 2008.
Attorney for licensee: M. S. Ross, Managing Attorney, Florida Power
and Light Company, P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Thomas H. Boyce.
Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220 and 50-410,
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2 (NMP1 and NMP2),
Oswego County, New York
Date of application for amendment: December 20, 2007.
Brief description of amendments: The amendments revise NMP1
Technical Specification (TS) Section 6.3, ``Unit Staff
Qualifications,'' and NMP2 TS Section 5.3, ``Unit Staff
Qualifications,'' to update requirements that have been superseded due
to the accreditation of the NMPNS licensed operator training program
and due to promulgation of the revised Title 10 of the Code of Federal
Regulations (10 CFR), Part 55, ``Operators' Licenses,'' which became
effective on May 26, 1987 (52 FR 9453). Additionally, the amendment for
NMP1 revises the TSs by eliminating the qualification requirement
exceptions listed for the position of Manager Operations which were
previously approved by the NRC staff. The position of Manager
Operations would meet the minimum qualification requirements as
required in American National Standard Institute Standard NI8.1-1971,
``American National Standard for Selection and Training of Nuclear
Power Plant Personnel.''
Date of issuance: September 29, 2008.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment Nos.: 198 and 127.
Renewed Facility Operating License No. DPR-63 and NPF-069:
Amendments revise the License and TSs.
Date of initial notice in Federal Register: January 28, 2008 (73 FR
5225).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 29, 2008.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: October 3, 2007.
Brief description of amendments: The amendments revised a footnote
in Technical Specifications Table 3.3.2.1-1, ``Control Rod Block
Instrumentation,'' such that a new banked position withdrawal sequence
shutdown sequence could be utilized. Associated changes are made to the
TS Bases. This operating license improvement was made available by the
NRC staff on May 23, 2007, as part of the consolidated line item
improvement process.
Date of issuance: October 1, 2008.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1--258, Unit 2--202.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the licenses and the technical specifications.
[[Page 62574]]
Date of initial notice in Federal Register: November 6, 2007 (72 FR
62691).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 1, 2008.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: October 5, 2007.
Brief description of amendments: The amendments revise the TSs
completion times (CTs) for TS Limiting Condition of Operation (LCO)
3.8.1, Conditions B and C, by specifying when maintenance restrictions
need to be met and by adding a 72-hour CT for the swing DG 1B.
Date of issuance: October 2, 2008.
Effective date: As of the date of issuance and shall be implemented
within 45 days from the date of issuance.
Amendment Nos.: Unit 1--259, Unit 2--203.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: November 6, 2007, (72
FR 62691).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 2, 2008.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama Southern Nuclear Operating Company, Inc., Georgia Power
Company, Oglethorpe Power Corporation, Municipal Electric Authority of
Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin
I. Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia Southern
Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle
Electric Generating Plant, Units 1 and 2, Burke County, Georgia
Date of application for amendments: June 12, 2008.
Brief description of amendments: The amendments revised the
Technical Specifications requirement for the Plant Manager or the
Operations Manager regarding the holding of a Senior Reactor Operator
license.
Date of issuance: October 7, 2008.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Farley Unit 1--179; Unit 2--171; Hatch Unit 1--260;
Unit 2--204; Vogtle Unit 1--153; Unit 2--134.
Facility Operating License Nos. NPF-2 and NPF-8; DPR-57 and NPF-5;
NPF-68 and NPF-81: Amendments revised the licenses and the technical
specifications.
Date of initial notice in Federal Register: July 1, 2008, 73 FR
37505.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 7, 2008.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of application for amendment: April 14, 2008.
Brief description of amendment: The amendment revises the list of
topical reports referenced in Technical Specification Section
6.9.1.14.a for use in preparing the core operating limits report by
adding EMF-2103P-A, ``Realistic Large Break LOCA Methodology for
Pressurized Water Reactors.'' The change will be utilized in core
loading designs for Unit 1 fuel-load configurations in future operating
cycles.
Date of issuance: September 24, 2008.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 320.
Facility Operating License No. DPR-77: Amendment revises the
technical specifications.
Date of initial notice in Federal Register: June 10, 2008 (73 FR
32746). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 24, 2008.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 10th day of October 2008.
For the Nuclear Regulatory Commission.
Joseph Gitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E8-24896 Filed 10-20-08; 8:45 am]
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