[Federal Register: November 4, 2008 (Volume 73, Number 214)]
[Notices]
[Page 65685-65705]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr04no08-81]
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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 October 9, 2008 to October 22, 2008. The
last biweekly notice was published on October 21, 2008 (73 FR 370501).
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
[[Page 65686]]
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 support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding, subject
to any limitations in the order granting leave to intervene, and have
the opportunity to participate fully in the conduct of the hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for hearing or a petition for leave to intervene must be
filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve 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
hearingdocket@nrc.gov, 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).
[[Page 65687]]
Each petitioner/requestor will need to download the Workplace Forms
Viewer\TM\ to access the Electronic Information Exchange (EIE), a
component of the E-Filing system. The Workplace Forms Viewer\TM\ is
free and is available at http://www.nrc.gov/site-help/e-submittals/
install-viewer.html. Information about applying for a digital ID
certificate is available on NRC's public Web site at http://
www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC 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 pdr@nrc.gov.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit No.1, DeWitt County, Illinois
Date of amendment request: September 2, 2008.
Description of amendment request: The proposed amendment would
relocated surveillance requirement (SR) 3.8.3.6 from the technical
specifications (TSs) to a licensee-controlled document. SR 3.8.3.6
requires the emergency diesel generator fuel oil storage tanks to be
drained, sediment removed, and cleaned on a 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 FOSTs [fuel oil storage tanks] provide the storage for the
DG [diesel generator] DG fuel oil, assuring an adequate volume is
available for each DG to operate for seven days in the event of a
loss of offsite power concurrent with a loss of coolant accident.
The relocation of the SR to drain and clean the FOSTs to a licensee-
controlled document will not impact any of the previously analyzed
accidents. Sediment in the tank, or failure to perform this SR, does
not necessarily result in an inoperable storage tank. Fuel oil
quantity and quality are assured by other TS SRs that remain
unchanged.
These SRs help ensure tank sediment is minimized and ensure that
any degradation of the tank wall surface that results in a fuel oil
volume reduction is detected and corrected in a timely manner.
Future changes to the licensee-controlled document will be evaluated
pursuant to the requirements of 10 CFR 50.59, ``Changes, tests, and
experiments,'' to ensure that such changes do not result in more
than a minimal 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, and configuration or the manner in which the plant is
operated and maintained. The proposed change does not adversely
affect the ability of structures, systems or components (SSCs) to
perform their intended function to mitigate the consequences of an
initiating event within the assumed acceptance limits.
The proposed change does not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
Further, the proposed change does not increase the types and amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposure.
[[Page 65688]]
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed TS change does not involve the addition or
modification of any plant equipment. Also, the proposed change will
not alter the design configuration, or method of operation of plant
equipment beyond its normal functional capabilities. The
requirements retained in the TS continue to require testing of the
diesel fuel oil to ensure the proper functioning of the DGs. The
proposed TS change does not create any new credible failure
mechanisms, malfunctions or accident initiators.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change does not alter or exceed a design basis or
safety limit. The requirements retained in the TS continue to
require testing of the diesel fuel oil to ensure the DGs are able to
perform their intended function.
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. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and TN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: August 29, 2008.
Description of amendment request: The amendments would modify
Technical Specification (TS) 5.6.5, Core Operating Limits Report
(COLR), by updating TS 5.6.5b to reflect the current analytical methods
used to determine the core operating limits in Palo Verde Nuclear
Generating Station (PVNGS), Units 1, 2, and 3. The proposed amendment
is an administrative change and all of the analytical methods have been
previously reviewed and approved by the Nuclear Regulatory Commission
(NRC).
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 to the list of methodologies used at PVNGS
[PVNGS, Units 1, 2, and 3] to determine the various COLR limits is
an administrative change which updates the list in the TS to include
NRC reviewed and approved COLR methodologies for PVNGS. It does not
add or modify any previously used methodologies; it updates the list
to include those already approved for use. This change does not make
any physical changes to any structure, system or component, and it
does not affect any design basis accident evaluation.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change to the list of methodologies used at PVNGS
to determine the various COLR limits is an administrative change
which updates the list in the TS to include all of the NRC reviewed
and approved COLR methodologies for PVNGS. This change does not
create any new failure modes or affect the interaction between any
structure, system or component.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to the list of methodologies used at PVNGS
to determine the various COLR limits is an administrative change
which updates the list in the TS to include all of the NRC reviewed
and approved COLR methodologies for PVNGS. This change does not
modify any margin of safety.
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
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: Michael T. Markley.
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 29, 2008.
Description of amendments request: The amendment would revise
Calvert Cliffs Nuclear Power Plant (CCNPP) Operating License Nos. DPR-
53 and DPR-69 and Technical Specifications (TSs) by increasing the
licensed core power of CCNPP, Unit Nos. 1 and 2 by 1.38 percent to 2737
MWt. The power uprate amendment request is based on the use of the
Caldon Leading Edge Flow Measurement (LEFM) CheckPlus system for more
accurate determination of main feedwater flow and the associated
determination of reactor power through the performance of the power
calorimetric calculation currently required by CCNPP TSs.
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. 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.
In support of this measurement uncertainty recapture (MUR) power
uprate, a comprehensive evaluation was performed for Nuclear Steam
Supply System (NSSS), balance of plant systems and components, and
analyses that could be affected by this change. A power calorimetric
uncertainty calculation was performed, and the impact of increasing
plant power by 1.38 percent on the plant's design and licensing
basis was evaluated. The result of these evaluations is that
structures, systems, and components required to mitigate transients
will continue to be capable of performing their design function at
an uprated core power of 2737 MWt. In addition, an evaluation of the
accident analyses demonstrates that applicable analysis acceptance
criteria continue to be met. No accident initiators are affected by
this uprate and no challenges to any plant safety barriers are
created by this change. Therefore, operation of the facility in
accordance with the proposed change will not involve a significant
increase in the probability of an accident previously evaluated.
The proposed change does not affect the radiological release
paths, the frequency of release, or the source-term for release for
any accidents previously evaluated in the Updated Final Safety
Analysis Report. Structures, systems, and components required to
mitigate transients remain capable of performing their design
functions, and thus were found acceptable. The reduced uncertainty
in the feedwater flow input to the power calorimetric measurement
ensures that
[[Page 65689]]
applicable accident analyses acceptance criteria continue to be met
in support of operation at a core power of 2737 MWt. Analyses
performed to assess the effects of mass and energy remain valid. The
source-terms used to assess radiological consequences have been
reviewed and determined to bound operation at the uprated condition.
Therefore, operation of the facility in accordance with the proposed
change will not involve a significant increase in the consequences
of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
No new accident scenarios, failure mechanisms, or single-
failures are introduced as a result of the proposed changes. The
installation of the Caldon LEFM CheckPlus feedwater flow
instrumentation system has been analyzed, and failures of this
system will have no adverse effect on any safety-related system or
any structures, systems, and components required for transient
mitigation. All structures, systems and components previously
required for the mitigation of a transient remain capable of
fulfilling their intended design functions. The proposed changes
have no adverse effects on any safety-related system or component
and do not challenge the performance or integrity of any safety-
related system.
This change does not adversely affect any current system
interfaces or create any new interfaces that could result in an
accident or malfunction of a different kind than was previously
evaluated. Operating at a core power level of 2737 MWt does not
create any new accident initiators or precursors. The reduced
uncertainty in the feedwater flow input to the power calorimetric
measurement ensures that applicable accident analyses acceptance
criteria continue to be met to support operation at a core power of
2737 MWt. Credible malfunctions continue to be bounded by the
current accident analysis of record or evaluations that demonstrate
that applicable acceptance criteria continue to be met.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The margins of safety associated with the MUR power uprate are
those pertaining to core power. This includes those associated with
the fuel cladding, Reactor Coolant System pressure boundary, and
containment barriers. A comprehensive engineering review was
performed to evaluate the 1.38 percent increase in the licensed core
power from 2700 MWt to 2737 MWt. The 1.38 percent increase required
that revised NSSS design thermal and hydraulic parameters be
established, which then served as the basis for all of the NSSS
analyses and evaluations. This engineering review concluded that no
design modifications are required to accommodate the revised NSSS
design conditions. The NSSS components were evaluated and it was
concluded that the NSSS components have sufficient margin to
accommodate the 1.38 percent power uprate. The NSSS accident
analyses were evaluated for the 1.38 percent power uprate. In all
cases, the evaluations demonstrate that the applicable analyses
acceptance criteria continue to be met. As a result, the margins of
safety continue to be bounded by the current analyses of record for
this change.
Therefore, the proposed change does not involve a significant
reduction in [a] margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: September 11, 2008.
Description of amendment request: The proposed amendment would
revise the Technical Specifications, extending the 15-year interval
between containment Type A tests specified by Specification 4.4.a,
``Integrated Leak Rate Test,'' by 6 months. The current Type A test
interval expires at the end of April 2009. The proposed amendment would
extend this interval, on a one-time basis, to October 2009 to coincide
with completion of the next scheduled refueling outage.
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 probability or consequences of accidents previously
evaluated in the Updated Safety Analysis Report are unaffected by
this proposed change. There is no change to any equipment response
or accident mitigation scenario, and this change results in no
additional challenges to fission product barrier integrity. The
proposed change does not alter the design, configuration, operation,
or function of any plant system, structure, or component. As a
result, the probabilities of previously evaluated accidents are
unaffected. The proposed extension to the Type A test interval does
not involve a significant increase in consequences because, as
discussed in NUREG-1493, Performance Based Containment Leak Rate
Test Program, Type B and C tests identify the vast majority
(approximately 97 percent) of all potential leakage paths. Further,
Type A tests identify only a few potential leakage paths that cannot
be identified through Type B and C testing, and leaks found by Type
A testing have been only marginally greater than existing
requirements. The frequency and methods of performance of Type B and
Type C testing are unaffected by this proposed change. In addition,
periodic inspections of containment required by the ASME [American
Society of Mechanical Engineers] code and the maintenance rule,
which are capable of detecting any significant degradation, are
unaffected by the proposed change.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed change.
The proposed change does not challenge the performance or integrity
of any safety-related system. The proposed change does not install
or remove any plant equipment. The proposed change does not alter
the design, physical configuration, or mode of operation of any
plant structure, system, or component. No physical changes are being
made to the plant, so no new accident causal mechanisms are being
introduced. The proposed change only changes the frequency of
performing the next Type A test; the Type A test implementation and
acceptance criteria are unchanged. Type B and Type C testing
frequency and method of performance are not affected by this
proposed change.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed change will have no affect
on the availability, operability, or performance of the safety-
related systems and components. The proposed change does not alter
the design, configuration, operation, or function of any plant
system, structure, or component. The ability of operable structures,
systems, and components to perform their designated safety function
is unaffected by this proposed change. NUREG-
[[Page 65690]]
1493 concluded that reducing the frequency of Type A tests to one-
in-20 years resulted in an imperceptible increase in risk. Type B
and Type C testing frequency and method of performance are
unaffected by this proposed change. Also, [other] inspections of
containment required by the ASME code and the maintenance rule
[will] provide reasonable assurance that containment will not
degrade in a manner that is only detectable by Type A testing. In
addition, the inherent risk of an additional plant shutdown would be
eliminated by the proposed amendment, further ensuring no
significant reduction in safety margin.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc.,
120 Tredegar Street, Richmond, VA 23219.
NRC Branch Chief: Lois M. James.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: July 28, 2008.
Description of amendment request: The proposed amendment would: (1)
Delete Technical Specification (TS) surveillance requirement (SR)
3.1.3.2 and revise SR 3.1.3.3, (2) remove reference to SR 3.1.3.2 from
Required Action A.2 of TS 3.1.3, ``Control Rod OPERABILITY,'' (3)
clarify the requirement to fully insert all insertable rods for the
limiting condition for operation (LCO) in TS 3.3.1.2, required Action
E.2, ``Source Range Monitoring Instrumentation,'' and (4) revise
Example 1.4-3 in Section 1.4, ``Frequency,'' to clarify the
applicability of the 1.25 surveillance test interval extension.
The NRC staff issued a notice of opportunity to comment in the
Federal Register on August 16, 2007 (72 FR 46103), on possible
amendments to revise the plant-specific TSs, modify TS control rod SR
testing frequency, clarify TS control insertion requirements, and
clarify SR frequency discussions, including a model safety evaluation
and model no significant hazards consideration (NSHC) determination,
using the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on November 13, 2007 (72 FR 63935). The licensee affirmed the
applicability of the model NSHC determination in its application dated
July 28, 2008.
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 generically implements TSTF-475, Revision 1,
``Control Rod Notch Testing Frequency and SRM [Source Range Monitor]
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 the 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 [General Electric]
Nuclear Energy Report, ``CRD [Control Rod Drive] 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 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 amendment request 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.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: July 21, 2008.
Description of amendment request: The proposed amendment would
support a proposed change to the in-service inspection program that is
based on topical report WCAP-16168-NP-A, Revision 2, ``Risk-Informed
Extension of the Reactor Vessel In-Service Inspection Interval.'' In
the referenced safety evaluation of the topical report, the NRC
required licensees to amend their licenses to require that the
information and analyses requested in Section (e) of the final 10 CFR
50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to
issuance of the final 10 CFR 50.61a) be submitted for NRC staff review
and approval within one year of completing the required reactor vessel
weld inspection. Entergy Nuclear Operations, Inc., proposes to add a
new license condition to provide this information.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment changes the renewed facility operating
license by adding a license condition to require that the
[[Page 65691]]
information and analyses requested in Section (e) of the final 10
CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275
prior to issuance of the final 10 CFR 50.61a) will be submitted for
NRC staff review and approval within one year of completing the
required reactor vessel weld inspection. The proposed amendment does
not involve operation of the required structures, systems or
components (SSCs) in a manner or configuration different from those
previously recognized or evaluated.
The proposed changes are administrative and have no impact on
plant operation or equipment.
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 amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed license amendment does not involve a physical
alteration of any SSC or change the way any SSC is operated. The
proposed license amendment does not involve operation of any
required SSCs in a manner or configuration different from those
previously recognized or evaluated.
The proposed changes are administrative and have no impact on
plant operation or equipment.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes are administrative and have no impact on
plant operation or equipment or on any margin of safety.
Therefore, the proposed amendment would 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. William Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White
Plains, NY 10601.
NRC Branch Chief: Lois M. James.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: August 28, 2008.
Description of amendment request: The proposed amendment would
change Technical Specifications (TS) Administrative Controls section 5
to incorporate NRC-approved Technical Specification Task Force (TSTF)
Improved Technical Specification (ITS) TSTF-363, ``Revise Topical
Report references in ITS 5.6.5, [Core Operating Limits Report] COLR,''
revision 0. ENO also proposes to make an administrative change to the
plant staff qualifications section.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed changes are administrative or provide
clarification only.
The proposed changes do not have any impact on the integrity of
any plant system, structure, or component (SSC) that initiates an
analyzed event. The proposed changes will not alter the operation
of, or otherwise increase the failure probability of any plant
equipment that initiates an analyzed accident. Thus, the probability
of any accident previously evaluated is not significantly increased.
The proposed changes do not affect the ability to mitigate
previously evaluated accidents, and do not affect radiological
assumptions used in the evaluations. The proposed changes do not
change or alter the design criteria for the systems or components
used to mitigate the consequences of any design-basis accident. The
proposed amendment does not involve operation of the required SSCs
in a manner or configuration different from those previously
recognized or evaluated. Thus, the radiological consequences of any
accident previously evaluated are not increased.
Therefore, operation of the facility in accordance with the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed amendment does not involve a physical
alteration of any SSC or a change in the way any SSC is operated.
The proposed amendment does not involve operation of any required
SSCs 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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The amendment does not involve a significant reduction in a
margin of safety. The proposed amendment does not affect any margin
of safety. The proposed amendment does not involve any physical
changes to the plant or manner in which the plant is operated.
Therefore, the proposed amendment would 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. William Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White
Plains, NY 10601.
NRC Branch Chief: Lois M. James.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: September 30, 2008.
Description of amendment request: The proposed amendment would
revise the Facility Operating License and Technical Specification
Section 4.0 by changing the names of the licensees to Enexus Nuclear
Pilgrim LLC and EquaGen Nuclear LLC.
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:
The proposed amendment would only change the names of the
licensees and reflect the referenced NRC Order requirements.
Principal management and operational staffing for the restructured
organization remain largely unchanged. The proposed changes do not:
(a) Involve a significant increase in the probability or
consequences of an accident previously evaluated; (b) create the
possibility of a new or different kind of accident from any accident
previously evaluated; or (c) 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. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400
[[Page 65692]]
Hamilton Avenue, White Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc. Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: September 4, 2008.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) Section 5.1, ``Site,'' to
remove the restriction on the sale and lease of site property and
replace the restriction with a requirement to retain complete authority
to determine and maintain sufficient control of all activities,
including the authority to exclude or remove personnel and property,
within the minimum exclusion area as described in 10 CFR 100.3.
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 change does not impact the function of any
structure, system or component that affects the probability of an
accident or that supports mitigation or consequences of an accident
previously evaluated. The proposed change establishes requirements
for sale or lease of property within the exclusion area.
Additionally, ENO [Entergy Nuclear Operations, Inc.] will retain
authority to determine all activities within the exclusion area and
to remove personnel and property from the area as necessary to
ensure the regulatory exposure limits are met.
The proposed change does not affect reactor operations or
accident analysis and there is no change to the radiological
consequences of a previously analyzed accident. The operability
requirements for accident mitigation systems remain consistent with
the licensing and design basis. Therefore, the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The proposed change does not involve any physical
alteration of plant equipment and does not change the method by
which any safety-related system performs its function. The proposed
change establishes requirements for sale or lease of property within
the exclusion area. Any additional activities performed within the
exclusion area will be reviewed by ENO and verified to not represent
a new hazard or that they have been accommodated in the plant
licensing and design basis. As such, no new or different types of
equipment will be installed or operated without additional review
and approval by ENO. Operation of existing installed equipment is
unchanged. The methods governing plant operation and testing remain
consistent with current safety analysis assumptions. 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. These changes do not change any existing design or
operational requirements, and do not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analysis. As such, there are no changes being
made to safety analysis assumptions, safety limits or safety system
settings that would adversely affect plant safety as a result of the
proposed change. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc. Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: September 22, 2008.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) to remove the requirement to
perform quarterly closure time testing of the Main Steam Isolation
Valves (MSIVs) by deleting TS Surveillance Requirement 4.7.D.1.c.
Operability testing of the MSIVs will continue to be required by the
Vermont Yankee Inservice Test Program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The operation of Vermont Yankee Nuclear Power Station (VY) in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
This proposed change deletes the specific surveillance
requirement to exercise the MSIVs once per quarter from the TS.
Following implementation of the proposed change, the VY TS still
will require operability testing of the MSIVs by reference to the VY
IST program. The quarterly exercise involves a timed full stroke
closure of each individual MSIV and subsequent reopening to the full
open position. Details of MSIV testing requirements will continue to
be contained in the VY IST program. The MSIV closure time setpoint
values related to the safety functions of the MSIVs will continue to
be contained in the VY UFSAR [Updated Final Safety Analysis Report]
and the VY TRM [Technical Requirements Manual]. Changes to the VY
UFSAR and TRM are evaluated per the requirements of 10 CFR 50.59.
These controls are adequate to ensure the required inservice testing
is performed to verify the MSIVs are operable and capable of
performing their safety functions. The proposed amendment introduces
no new equipment or changes to how equipment is operated. Therefore,
the proposed amendment will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station (VY) in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed amendment deletes the specific surveillance
requirement to exercise the MSIVs once per quarter from the TS.
Following implementation of the proposed change, the VY TS still
will require operability testing of the MSIVs by reference to the VY
IST program. The quarterly exercise involves a timed full stroke
closure of each individual MSIV and subsequent reopening to the full
open position. The proposed amendment does not change the design or
function of any component or system. No new modes of failure or
initiating events are being introduced. Therefore, operation of VY
in accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The operation of Vermont Yankee Nuclear Power Station (VY) in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
The proposed amendment deletes the specific surveillance
requirement to exercise the MSIVs once per quarter from the TS.
Following implementation of the proposed change, the VY TS still
will require operability testing of the MSIVs by reference to the VY
IST program. The quarterly exercise involves a timed full stroke
closure of each individual MSIV and subsequent reopening to the full
open position. The proposed amendment does not change the design or
function of any component or system. The proposed amendment does not
involve any safety limits or safety settings. The ability of the
MSIVs to perform their safety function will continue to be tested in
[[Page 65693]]
accordance with the IST Program, through TS SR 4.7.D.1.b.
Since the proposed controls are adequate to ensure the required
inservice testing is performed, there will still be high assurance
that the components are operable and capable of performing their
respective safety functions, and that the systems will respond as
designed to mitigate the subject events. Therefore, operation of VY
in accordance with the proposed amendment will not involve a
significant reduction in [a] margin to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc. Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: September 30, 2008.
Description of amendment request: The proposed amendment would
revise the Facility Operating License and Technical Specification
Section 5.0 by changing the names of the licensees to EquaGen Nuclear
LLC and Enexus Nuclear Vermont Yankee LLC, respectively.
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:
The proposed amendment would only change the names of the
licensees and reflect the referenced NRC Order requirements;
principal management and operational staffing for the restructured
organization remain largely unchanged. The proposed changes do not:
(a) Involve a significant increase in the probability or
consequences of an accident previously evaluated; (b) create the
possibility of a new or different kind of accident from any accident
previously evaluated; or (c) 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. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment requests: July 21, 2008.
Description of amendment request: The proposed change allows a
delay time for entering a supported system Technical Specification (TS)
when the inoperability is due solely to an inoperable snubber, if risk
is assessed and managed consistent with the program in place for
complying with the requirements of 10 CFR 50.65(a)(4). Limiting
Condition for Operation (LCO) 3.0.8 is added to the TS to provide this
allowance and define the requirements and limitations for its use.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff
issued a notice of opportunity for comment in the Federal Register on
November 24, 2004 (69 FR 68412), on possible amendments concerning
TSTF-372, including a model safety evaluation and model no significant
hazards consideration (NSHC) determination, using the consolidated line
item improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on May 4, 2005 (70 FR 23252).
Basis for proposed no significant hazards consideration
determination: Entergy Operations, Inc. (Entergy) has reviewed the
proposed NSHC determination published in the Federal Register as part
of the CLIIP. Entergy has concluded that the proposed NSHC
determination presented in the Federal Register notice is applicable to
Arkansas Nuclear One, Unit 2 and 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 a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. Therefore, the probability
of an accident previously evaluated is not significantly increased,
if at all. The consequences of an accident while relying on
allowance provided by proposed LCO 3.0.8 are no different than the
consequences of an accident while relying on the TS required actions
in effect without the allowance provided by proposed LCO 3.0.8.
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).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, 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 a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in Regulatory Guide 1.177. A bounding risk assessment
was performed to justify the proposed TS changes. The proposed LCO
3.0.8 defines limitations on the use of the provision and includes a
requirement for the licensee to assess and manage the risk
associated with operation with an inoperable snubber. 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 amendment request
involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
[[Page 65694]]
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: July 30, 2008, as supplemented on
October 2, 2008.
Description of amendment request: Entergy Operations Inc. (the
licensee) proposes to modify the technical specifications (TS) 3.6.6,
``Spray Additive System.'' Specifically, this amendment proposes to
revise the Sodium Hydroxide (NaOH) tank concentration stated in TS
3.6.6.3 from between 5.0 percent and 16.5 percent to between 6.0
percent and 8.5 percent.
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.
There are no changes to the design or operation of the plant
that could affect system, component, or accident functions as a
result of changing the sodium hydroxide (NaOH) tank solution
concentration limits. In addition, the dose reduction provided by
maintaining the sump pH above 7.0 is retained, and therefore, dose
consequences resulting from iodine dissolution remain unchanged. The
proposed change simply imposes more restrictive operating conditions
than are within the current TS limits. 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 accident scenarios, failure mechanisms, or single
failures are introduced as a result of the proposed change.
Structures, systems, and components previously required for
mitigation of an accident remain capable of fulfilling their
intended design function with this change to the TS. The proposed
change has no new adverse effects on safety-related systems or
components and does not challenge the performance or integrity of
safety-related systems. The proposed change simply imposes more
restrictive operating conditions that are within the current TS
limits. 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 change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change imposes more restrictive operating
conditions that are within the current TS limits. Revising the NaOH
tank solution concentration limits reduces the amount of chemical
precipitates formed under post-loss-of-coolant accident conditions.
The margin of safety related to ensuring that the sump pH remains
above 7.0 is not reduced. 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: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: August 21, 2008.
Description of amendment request: Entergy Operations Inc. (the
licensee) proposes a one-time amendment for next containment integrated
leakage rate test (ILRT) or Type A test at the Arkansas Nuclear One,
Unit No. 2 (ANO-2). The ILRT is required by Technical Specification
(TS) 6.5.16, ``Containment Leakage Rate Testing Program,'' to be
performed every ten-years. The amendment would permit the existing ILRT
frequency to be extended from 120 months (10 years) to approximately
135 months.
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 exemption involves a one-time extension to the
current interval for Type A containment testing. The current test
interval of 120 months (10 years) would be extended on a one-time
basis to no longer than approximately 135 months from the last Type
A test. The proposed extension does not involve a physical change to
the plant or a change in the manner in which the plant is operated
or controlled. The containment is designed to provide an essentially
leak tight barrier against the uncontrolled release of radioactivity
to the environment for postulated accidents. As such, the reactor
containment itself and the testing requirements invoked to
periodically demonstrate the integrity of the reactor containment
exist to ensure the plant's ability to mitigate the consequences of
an accident, and do not involve the prevention or identification of
any precursors of an accident. Therefore, this proposed extension
does not involve a significant increase in the probability of an
accident previously evaluated nor does it create the possibility of
a new or different kind of accident.
This proposed extension is for the Type A containment leak rate
tests only. The Type B and C containment leak rate tests will
continue to be performed at the frequency currently required by the
ANO-2 TS. As documented in NUREG 1493, Type B and C tests have
identified a very large percentage of containment leakage paths and
that the percentage of containment leakage paths that are detected
only by Type A testing is very small. ANO-2's Type A test history
supports this conclusion.
The integrity of the reactor containment is subject to two types
of failure mechanisms which can be categorized as (1) activity based
and (2) time based. Activity based failure mechanisms are defined as
degradation due to system and/or component modifications or
maintenance. Local leak rate test requirements and administrative
controls such as configuration management and procedural
requirements for system restoration ensure that containment
integrity is not degraded by plant modifications or maintenance
activities. The design and construction requirements of the
containment itself combined with the containment inspections
performed in accordance with ASME, Section XI, the Maintenance Rule,
and Licensing commitments serve to provide a high degree of
assurance that the containment will not degrade in a manner that is
detectable only by a Type A test. Based on the above, the proposed
extension does not involve a significant increase in the
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 revision to the TS involves a one-time extension to
the current interval for Type A containment testing. The reactor
containment and the testing requirements invoked to periodically
demonstrate the integrity of the reactor containment exist to ensure
the plant's ability to mitigate the consequences of an accident and
do not involve the prevention or identification of any precursors of
an accident. The proposed TS change does not involve a physical
change to the plant or the manner in which the plant is operated or
controlled. Therefore, the proposed TS 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 to the TS involves a one-time extension to
the current interval for Type A containment testing. The proposed TS
change does not involve a physical
[[Page 65695]]
change to the plant or a change in the manner in which the plant is
operated or controlled. The specific requirements and conditions of
the Primary Containment leak Rate Testing Program, as defined in the
TS, exist to ensure that the degree of reactor containment
structural integrity and leak-tightness that is considered in the
plant safety analysis is maintained. The overall containment leak
rate limit specified by TS is maintained. The proposed change
involves only the extension of the interval between Type A
containment leak rate tests. The proposed surveillance interval
extension is bounded by the 15 month extension currently authorized
within NEI 94-01, Revision 0. Type B and C containment leak rate
tests will continue to be performed at the frequency currently
required by TS. Industry experience supports 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 Type A testing is small. The containment
inspections performed in accordance with ASME, Section XI and the
Maintenance Rule serve to provide a high degree of assurance that
the containment will not degrade in a manner that is detectable only
by Type A testing. The combination of these factors ensures that the
margin of safety that is in plant safety analysis is maintained. The
design, operation, testing methods and acceptance criteria for Type
A, B, and C containment leakage tests specified in applicable codes
and standards will continue to be met, with the acceptance of this
proposed change, since these are not affected by changes to the Type
A test interval. Therefore, the proposed TS 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: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
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 requests: The proposed amendments would
modify the Technical Specification (TS) by adding Limiting Condition
for Operation (LCO) 3.0.8 on the inoperability of snubbers using the
Consolidated Line Item Improvement Process (CLIIP). The proposed
amendments would also make conforming changes to TS LCO 3.0.1. This
request is consistent with NRC-approved Industry/Technical
Specification Task Force (TSTF) Traveler No. 372, Revision 4,
``Addition of LCO 3.0.8, Inoperability of Snubbers.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on November 24, 2004 (69 FR 68412), on possible
amendments concerning TSTF-372, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on May 4, 2005 (70 FR 23252). Basis for proposed no significant hazards
consideration determination: Entergy Operations, Inc. (Entergy) has
reviewed the proposed NSHC determination published in the Federal
Register as part of the CLIIP. Entergy has affirmed the applicability
of the following NSHC for Arkansas Nuclear One, Unit 1 in its
application and as published in the Federal Register.
Criterion 1: The Proposed Changes Do Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously Evaluated
The proposed changes allow a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. Therefore, the probability
of an accident previously evaluated is not significantly increased,
if at all. The consequences of an accident while relying on
allowance provided by proposed LCO 3.0.8 are no different than the
consequences of an accident while relying on the TS required actions
in effect without the allowance provided by proposed LCO 3.0.8.
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, these changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2: The Proposed Changes Do Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed).
Allowing delay times for entering a supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, 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, these changes do not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3: The Proposed Changes Do Not Involve a Significant
Reduction in the Margin of Safety
The proposed changes allow a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in NRC Regulatory Guide 1.177. A bounding risk
assessment was performed to justify the proposed TS changes. The
application of LCO 3.0.8 is predicated upon the licensee's
performance of a risk assessment and management of plant risk [which
is required by the proposed TS 3.0.8]. The net change to the margin
of safety is insignificant. Therefore, these changes do not involve
a significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael Markley.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 17, 2008.
Description of amendment request: The proposed change will revise
the Operating License to modify Note 2 of Waterford 3 Technical
Specification Table 4.3-1. The licensee stated that the proposed change
will result in the addition of conservatism to Core Protection
Calculator (CPC) power indications when calibrations are required in
certain conditions.
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 65696]]
Response: No.
The proposed change will redefine the tolerance band allowed for
the Reactor Protection System (RPS) linear power, Core Protection
Calculator (CPC) [Delta]T [Delta Temperature] power, and CPC neutron
flux power signals, and clarify the intent of the calibration
requirements for CPC power indications when at less than 15%
[percent] power, and specify that adjustment limits are percentages
of RATED THERMAL POWER instead of percentages of current power.
Redefining the tolerance band is in conformance with the safety
analysis. The consequences of an accident will be in conformance
with the safety analysis.
Clarifying the intent of there being no calibration requirements
for CPC power indications when at less than 15% power is essentially
editorial. At this low power level, CPC calculations compensate for
any potential de-calibration. Specifying that adjustment limits are
percentages of RATED THERMAL POWER instead of percentages of current
power is essentially editorial. This change is made to avoid
confusion in interpreting the requirements. This amendment request
does not change the design, analysis or operation of any plant
systems or components.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to Technical Specification power calibration
tolerance limits is in conformance with the safety analysis. This
amendment request does not change the design, analysis or operation
of any plant systems or components. CPC's cannot cause an accident,
and this change will not create the possibility of a new or
different type of accident.
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 change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to Technical Specification power calibration
tolerance limits is in conformance with the safety analysis. This
proposed change maintains the margin of safety for design basis
events. Therefore, this change 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
amendment request 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.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 17, 2008.
Description of amendment request: Entergy has proposed to add a
license condition on the extension of the reactor vessel inservice
inspection interval. This proposed license condition is the result of a
condition in the Nuclear Regulatory Commission (NRC) safety evaluation,
issued by letter dated May 8, 2008, on topical report WCAP-16168-NP-A,
Revision 2, ``Risk-Informed Extension of the Reactor Vessel In-Service
Inspection [ISI] Interval,'' dated June 8, 2008. The ISI interval
extension part of a relief request is being separately evaluated by NRC
and independent of this amendment request.
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 will revise the license to require the
submission of information and analyses to the NRC following
completion of each ASME Code, Section XI, Category B-A and B-D
reactor vessel weld inspection. The extension of the ISI interval
from 10 to 20 years is being evaluated as part of the relief request
independent from this license change. Submission of the information
and analyses are administrative in nature and has no impact on any
plant configuration or system performance relied upon to mitigate
the consequences of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
any SSC or change the way any SSC is operated. The proposed addition
of the license condition has no impact on any plant configurations
or on system performance that is relied upon to mitigate the
consequences of an accident. The license condition is administrative
in nature and does not result in a change to the physical plant or
to the modes of operation defined in the facility license. Entergy
has demonstrated that the Limitations and Conditions associated with
the NRC SE will be met.
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 change involve a significant reduction in a
margin of safety?
Response: No.
The addition of the license condition is administrative in
nature and has no impact on plant operation or equipment or on any
margin of safety. The license condition to submit information and
analyses is an administrative tool to assure the NRC has the ability
to independently review information developed by the Licensee.
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: Terence A. Burke, Associate General
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: June 27, 2007, as supplemented on
September 4, 2008.
Description of amendment request: The proposed amendment request
dated June 27, 2008, would revise Technical Specifications (TS)
Surveillance Requirements 3.8.1.2, 8, 12, 13, 16, and 19, changing the
steady state frequency and voltage of all diesel generators (DGs) from
the currently allowed frequency range of 59.4-61.2 Hz to 59.4-60.5 Hz
(i.e., a decrease of the upper limit, resulting in narrowing of the
current range). The licensee stated that the current frequency range is
nonconservative and could result in undesirable effects such as
centrifugal charging pump motor brake horsepower exceeding its
nameplate maximum horsepower, and overloading the DGs. The Commission
previously noticed this proposed amendment request on August 14, 2007
(72 FR 45458).
The scope of the June 27, 2008, proposed amendment request was
expanded as described in a supplemental letter dated September 4, 2008.
The expanded scope would revise (1) TS Surveillance Requirements
3.8.1.8, 13, 16, and 22, changing the minimum voltage and frequency
that
[[Page 65697]]
the DGs must achieve within 10 seconds after starting from >= 3740
Volts (V) to >= 3910 V and >= 58.8 Hz to >= 59.4 Hz, respectively, and
(2) TS Surveillance Requirement 3.8.1.10, changing the maximum DG
frequency allowed to occur within 2 seconds following a load rejection
of the single largest post-accident load from <= 61.2 Hz to <= 60.5 Hz.
The changes proposed by the supplement indirectly affect TS 3.8.2.1
which requires that TS Surveillance Requirements 3.8.1.8, 10, and 16 be
met.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration. The
NRC staff has performed its own analysis, which is presented below:
(1) Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
The more restrictive transient voltage and frequency limits
ensures that the equipment powered from the DGs will function as
designed to mitigate an accident as described in the Update Final
Safety Analysis Report (UFSAR). The DGs and the equipment they power
are part of the systems required to mitigate accidents; no accident
analyzed in the UFSAR is initiated by mitigation equipment.
Therefore, the proposed change to the allowed frequency range of the
DGs will not have any impact on the probability of an accident
previously evaluated. Furthermore, other than requiring more
restrictive transient voltage and frequency limits of DGs, there is
no other design or operational change. Therefore, the proposed
change does not increase the probability of malfunction of the DGs
or the equipment they power.
The more restrictive DG transient voltage and frequency limits
will ensure that the equipment powered by the DGs will perform as
originally designed and analyzed to mitigate the consequences of any
accident described in the UFSAR. Therefore, the proposed change does
not involve a significant increase in the consequences of an
accident previously evaluated in the UFSAR.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
There is no design change associated with the proposed
amendment. Making an existing DG requirement more restrictive alone
will not alter plant configuration because no new or different type
of equipment will be installed, and because no methods governing
plant operation will be changed. The proposed change to transient
voltage and frequency limits will not have any effect on the
assumptions of accident scenarios previously made in the UFSAR.
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 change involve a significant reduction in
a margin of safety?
Response: No.
Despite the proposed change to the DG transient voltage and
frequency limits, the DGs and equipment powered by the DGs will
continue to perform as originally designed, and originally analyzed
in the UFSAR. There is no associated change to the methods and
assumptions used to analyze DG performance. The proposed change will
maintain the required function of the DGs and the equipment powered
by the DGs to ensure that operation of structures, systems, or
components is as currently set forth in the UFSAR. Therefore, the
proposed change does not involve a significant reduction in the
margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on its own analysis, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the proposed amendment involves no
significant hazards consideration.
Attorney for licensee: James M. Petro, Jr., One Cook Place,
Bridgman, MI 49106.
NRC Branch Chief: Lois M. James.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: July 2, 2008.
Description of amendment request: The proposed amendment would
correct several typographical errors and make administrative
clarifications to the Technical Specifications (TS).
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 changes correct typographical and administrative
errors, or make clarifications that more accurately reflect TS
requirements. Administrative and editorial changes such as these are
not an initiator of any accident previously evaluated. As a result,
the probability of an accident previously evaluated is not affected.
The consequences of an accident with the incorporation of these
administrative and editorial changes are no different than the
consequences of the same accident without these changes. As a
result, the consequences of an accident previously evaluated are not
affected by these changes.
The proposed changes do not alter or prevent the ability of
structures, systems, and components from performing their intended
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 an
accident previously evaluated.
Further, the proposed changes do not increase the types or
amounts of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures. The proposed changes are consistent with the
safety analysis assumptions and resultant consequences. Therefore,
the proposed changes do not involve an increase in the probability
or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different accident from any accident previously evaluated?
Response: No.
The proposed 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. The proposed changes do not alter any assumptions made in
the safety analysis. Therefore, the proposed changes do not create
the possibility of a new or different 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 consist of administrative and editorial
changes to correct typographical or administrative errors and
oversights or clarify the meaning of the TS. The changes do not
alter the manner in which safety limits, limiting safety system
settings or limiting conditions for operation are determined. The
safety analysis acceptance criteria are not affected by these
changes. The proposed changes will not result in plant operation in
a configuration outside of the design basis. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: September 18, 2008.
Description of amendment request: The proposed amendment would
revise technical specification (TS) 3.8.7,
[[Page 65698]]
``Inverters--Operating.'' The current TS requires one inverter for each
of the four channels. The proposed amendment would revise TS 3.8.7 to
require two inverters for each of the four channels.
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 revisions to WBN's Vital AC [alternating current]
Power System do not alter the safety functions of the Vital
Inverters or the Unit 1 and Unit 2 120V [volt] AC Vital Instrument
Power Boards. The initial conditions for the DBAs [design-basis
accidents] defined in the WBN UFSAR [Updated Final Safety Analysis
Report] assume the ESF [engineered safety feature] systems are
operable. The vital inverters are designed to provide the required
capacity, capability, redundancy, and reliability to ensure the
availability of necessary power to vital instrumentation so that the
fuel, reactor coolant system, and containment design limits are not
exceeded. Separating the Unit 2 loads from the Unit 1 inverters does
not alter the accident analyses. Design calculations document that
the inverters have adequate capacity to support the loads required
for Unit 1 operation and no changes are proposed that will impact
the separation of the Vital AC Power System.
The inverters and the associated 120V AC Vital Instrument Power
Boards are utilized to support instrumentation that monitor critical
plant parameters to aid in the detection of accidents and to support
the mitigation of accidents, but are not considered to be an
initiator of design basis accidents. Based on this and the preceding
information, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
When implemented, the proposed TS amendment will allow the Unit
2 Vital Instrument Power Boards to receive their UPS
[uninterruptible power supply] power from new Unit 2 inverters.
Calculations have verified that the loads will not affect the
ability of the inverters to perform their intended safety functions.
In addition, the inverters and the 120V AC Vital Instrument Power
Boards are not considered to be an initiator of a DBA. These
components provide power to instrumentation that supports the
identification and mitigation of accidents as well as system control
functions during normal plant operations. The functions of the
inverters are not altered by the proposed TS change and will not
create the possibility of a new or different accident. Further, the
separation of the Unit 2 loads from the Unit 1 inverters is the
principal change to the inverter system, and this change is bounded
by previously evaluated accident analyses. 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 plant setpoints and limits that are utilized to ensure safe
operation and detect accident conditions are not impacted by the
proposed TS amendment. The inverters and the 120V Vital Instrument
Power Boards will continue to provide reliable power to safety-
related instrumentation for the identification and mitigation of
accidents and to support plant operation. Therefore, the margin of
safety is not reduced.
Based on the above, TVA concludes that the proposed amendment
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: September 18, 2008.
Description of amendment request: The proposed amendment would
revise technical specification (TS) Table 3.3.2-1, ``Engineered Safety
Feature Actuation System Instrumentation,'' to modify Mode 1 and 2
Applicability for Function 6.e, and would revise limiting condition for
operation (LCO) 3.3.2, Condition J.
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 design basis events which impose [auxiliary feedwater] AFW
safety function requirements are loss of normal main feedwater, main
feed line or main steam line break, loss of offsite power (LOOP),
and small break loss of coolant accident. These design bases event
evaluations assume actuation of the AFW due to LOOP signal, low-low
steam generator level or a safety injection signal. The anticipatory
AFW auto-start signals from the turbine driven main feedwater
(TDMFW) pumps are not credited in any design basis accidents and
are, therefore, not part of the primary success path for postulated
accident mitigation as defined by 10 CFR 50.36(c)(2)(ii), Criterion
3. Modifying Mode 1 and 2 Applicability for this function will not
impact any previously evaluated design basis accidents. 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.
This TS change allows for an operational allowance during Mode 1
and 2 for placing TDMFW pumps in service or securing TDMFW pumps.
This change involves an anticipatory AFW auto-start function that is
not credited in the accident analysis. Since this change only
affects the conditions at which this auto-start function needs to be
operable and does not affect the function that actuates AFW due to
loss of offsite power, low-low steam generator level or a safety
injection signal, it will not be an initiator to 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.
This TS change involves the automatic start of the AFW pumps due
to trip of both TDMFW pumps, which is not an assumed start signal
for design basis events. This change does not modify any values or
limits involved in a safety related function or accident analysis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, TVA concludes that the proposed amendment
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: September 19, 2008.
[[Page 65699]]
Description of amendment request: The proposed amendment would
modify the WBN Final Safety Analysis Report (FSAR) by requiring an
inspection of the ice condenser within 24 hours of experiencing a
seismic event greater than or equal to an Operating Basis Earthquake
(OBE) within the five week period after ice basket replenishment has
been completed.
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 analyzed accidents of consideration in regard to changes
potentially affecting the ice condenser are a loss of coolant
accident and a steam or feedwater line break inside Containment. The
ice condenser is an accident mitigator and is not postulated as
being the initiator of a LOCA [loss-of-coolant accident] or HELB
[high energy line break]. The ice condenser is structurally designed
to withstand a Safe Shutdown Earthquake plus a Design Basis Accident
and does not interconnect or interact with any systems that
interconnect or interact with the Reactor Coolant, Main Steam, or
Feedwater systems. Because the proposed changes do not result in, or
require any physical change to the ice condenser that could
introduce an interaction with the Reactor Coolant, Main Steam, or
Feedwater systems, there can be no change in the probability of an
accident previously evaluated.
Under the proposed change, there is some finite probability
that, within 24 hours following a seismic disturbance, a LOCA or
HELB in Containment could occur within five weeks of the completion
of ice basket replenishment. However, several factors provide
defense-in-depth and tend to mitigate the potential consequences of
the proposed change.
Design basis accidents are not assumed to occur simultaneously
with a seismic event. Therefore, the coincident occurrence of a LOCA
or HELB with a seismic event is strictly a function of the combined
probability of the occurrence of independent events, which in this
case is very low. Based on the Probabilistic Risk Assessment model
and seismic hazard analysis, the combined probability of occurrence
of a seismic disturbance greater than or equal to an OBE during the
5 week period following ice replenishment coincident with or
subsequently followed by a LOCA or HELB during the time required to
perform the proposed inspection (24 hours) and if required by
Technical Specifications, complete Unit shutdown (37 hours), is less
than 3.7E-09 for WBN. This probability is well below the threshold
that is typically considered credible.
Even if ice were to fall from ice baskets during a seismic event
occurring coincident with or subsequently followed by an accident,
the ice condenser would be expected to perform its intended safety
function. Due to the ice servicing methodology utilized by WBN, the
relatively small amount of ice that may potentially fallout from the
ice baskets to the floor behind the lower inlet doors during the
seismic event is such that complete blockage of flow into the ice
condenser is not credible during a LOCA or HELB.
Based on the above, the proposed changes do not involve a
significant increase in the probability or consequences. The ice
condenser is expected to perform its intended safety function under
all circumstances following a LOCA or HELB in Containment.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change provides an alternate methodology to confirm
the ice condenser lower inlet doors are capable of opening if a
seismic event occurs within five weeks of ice basket replenishment.
As previously discussed, the ice condenser is not postulated as an
initiator of any design basis accident. The proposed change does not
impact any plant system, structure, or component that is an accident
initiator. The proposed change does not involve any hardware changes
to the ice condenser or other changes that could create new accident
mechanisms. Therefore, there can be no new or different accidents
created from those previously identified and evaluated.
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 change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the Reactor Coolant system, and the Containment
system. The performance of the fuel cladding and the Reactor Coolant
system will not be impacted by the proposed change.
The requirement to inspect the ice condensers within 24 hours of
experiencing seismic activity greater than or equal to an OBE during
the five (5) week period following the completion of ice basket
replenishment will confirm whether the ice condenser lower inlet
doors are capable of opening. This inspection will either confirm
that the ice condenser doors remain fully capable of performing
their intended safety function under credible circumstances or that
a Unit shutdown is required.
The ice condenser has reasonable assurance of performing its
intended function during the highly unlikely scenario in which a
postulated accident (LOCA or HELB) occurs coincident with or
subsequently following a seismic event.
The proposed change affects the assumed timing of a postulated
seismic and design basis accident applied to the ice condenser and
provides an alternate methodology in confirming the ice condenser
lower inlet doors are capable of opening. As previously discussed,
the combined probability of occurrence of a LOCA or HELB and a
seismic disturbance greater than or equal to an OBE during the
``period of potential exposure'' is less than 3.7E-09 for WBN. This
probability is well below the threshold that is considered credible.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety. The WBN ice condenser will
perform its intended safety function under credible circumstances.
The changes proposed in this LAR [license amendment request] do
not make any physical alteration to the ice condensers, nor does it
affect the required functional capability of the ice condenser in
any way. The intent of the proposed change to the FSAR is to
eliminate an overly restrictive waiting period prior to Unit ascent
to power operations following the completion of ice basket
replenishment. The required inspection of the ice condenser
following a seismic event greater than or equal to an OBE will
confirm whether the ice condenser lower inlet doors will continue to
fully perform their safety function as assumed in the WBN safety
analyses.
Thus, it can be concluded that the proposed change does not
involve a significant reduction in the margin of safety.
Based on the above, TVA concludes that the proposed amendment
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: April 2, 2008.
Description of amendment request: The proposed change revises
Technical Specification (TS) Section 5.0, ``Design Features,'' to
delete certain design details and descriptions included in TS 5.0 that
are already contained in the Updated Final Safety Analysis Report
[[Page 65700]]
(UFSAR), or are redundant to existing TS requirements, and are not
required to be included in the TSs pursuant to Title 10 of the Code of
Federal Regulations (10 CFR), Part 50, Section 50.36(d)(4). The
proposed change also revises the format of, and incorporates design
descriptions into, TS 5.0 consistent with Nuclear Regulatory Commission
(NRC) policy and NUREG-1431, ``Standard Technical Specifications
Westinghouse Plants, Revision 3.0,'' to the extent practical. An
editorial change is also proposed to address a minor TS discrepancy
introduced by a previous license amendment. More specifically, the
proposed change includes removing Section 5.2, ``Containment,'' from
the TSs in its entirety. This section contains the minimum spray flows
for the Containment Spray (CS) and Recirculation Spray (RS) Subsystems.
The proposed change also removes the statement describing how draining
of the spent fuel pool is prevented, and includes a statement in the TS
that would limit draining the spent fuel pool below the elevation of 41
feet, 2 inches mean sea level. Additionally, the licensee proposes to
incorporate the spent fuel pool storage capacity of 1044 assemblies
into the TSs. This limit was previously established by Amendment Nos.
37 and 36 to Surry Power Station, Unit Nos. 1 and 2, respectively.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration. The
NRC staff has performed its own analysis, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to Section 5.0, ``Design Features,''
removes certain details from the TSs that are not required to be
maintained in the TSs by 10 CFR 50.36(d)(4), or are adequately
controlled by other existing TSs, incorporates previously approved
TS limits that meet the 10 CFR 50.36(d)(4) inclusion criteria, and
revises the TSs for consistency with NUREG-1431. An additional
change addresses a minor editorial discrepancy introduced by a
previous amendment. The minimum spray flow values for the CS and RS
Subsystems are removed, but operability and performance of both
subsystems are adequately controlled by existing TSs ensuring they
will continue to perform their design functions. The proposed
changes remove the statement describing how draining of the spent
fuel pool is prevented (does not meet the criteria of 10 CFR
50.36(d)(4)for inclusion in the TSs) and includes a statement in the
TS that would limit draining the spent fuel pool below the elevation
of 41 feet, 2 inches mean sea level (as analyzed in the UFSAR and
consistent with the content and format of NUREG-1431). The proposed
change incorporates the spent fuel pool storage capacity of 1044
assemblies into the TSs. This limit was evaluated in previously
approved Amendment Nos. 37 and 36 to Surry Power Station, Unit Nos.
1 and 2, respectively. The proposed changes are considered
administrative in nature and do not affect initiators of previously
analyzed events or assumed mitigation of accident or transient
events. Therefore, the proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
There is no physical alteration of the plant (no new or
different type of equipment will be installed) associated with the
proposed amendment. The proposed changes will not have any effect on
the assumptions of accident scenarios previously made in the UFSAR.
The proposed changes do not alter or prevent the ability of
structures, systems, and components to perform their intended
function to mitigate the consequences of an initiating event. The
proposed changes are considered administrative in nature. Therefore,
the proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
Response: No.
The spent fuel pool and the CS and RS Subsystems will continue
to perform as designed and analyzed in the UFSAR. There is no
associated change to the methods and assumptions used to analyze
their performance. Their required function will be maintained as
currently set forth in the UFSAR and existing TSs. The proposed
changes do not result in plant operation in a configuration outside
the design basis. The proposed changes do not adversely affect
systems that respond to safely shutdown the plant and to maintain
the plant in a safe shutdown condition. The dose analysis is also
not affected. The proposed changes are considered administrative in
nature and do not alter the manner in which safety limits, limiting
safety system settings or limiting conditions for operation are
determined. Therefore, the proposed amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
its own analysis, 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: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., 120 Tredegar Street, RS-2 Richmond,
VA 23219.
NRC Branch Chief: Melanie C. Wong.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Nine Mile Point Nuclear Station, LLC, (NMPNS) Docket No. 50-410, Nine
Mile Point Nuclear Station Unit No. 2, Oswego County, New York
Date of amendment request: July 30, 2007, as supplemented on April
7 and September 8, 2008.
Description of amendment request: This amendment would modify
Technical Specification 3.7.3, ``Control Room Envelope Air Conditioning
(AC) System,'' by adding an Action Statement to the Limiting Conditions
for Operation. The new Action Statement allows a finite time to restore
one control room envelope AC subsystem to operable status and requires
verification that the control room temperature remains <90 [deg]F every
4 hours.
Date of publication of individual notice in Federal Register: (73
FR 55166) September 24, 2008.
Expiration date of individual notice: November 23, 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
[[Page 65701]]
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 pdr@nrc.gov.
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: October 18, 2007, as supplemented by
letter dated July 3, 2008.
Description of amendment request: The amendment changed the Oyster
Creek Technical Specifications Section 4.5.M.1.e.1 regarding the
mechanical snubber functional test acceptance test acceptance criteria.
Specifically, the change replaced the snubber breakaway test with the
drag force test.
Date of issuance: October 10, 2008.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 270.
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 supplement dated July 3, 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 determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 10, 2008.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: November 29, 2007.
Brief description of amendment: The amendment consists of changes
to Technical Specification Section 3.6.8, ``Isolation Valve Seal Water
(IVSW) System.'' The amendment revises Surveillance Requirements (SR)
3.6.8.2 and 3.6.8.6 related to IVSW tank volume and header flow rates.
Specifically, the change clarifies the wording of SR 3.6.8.2, and
revises SR 3.6.8.6 to provide a total flow rate limit from all four
headers in place of the individual header limits.
Date of issuance: October 3, 2008.
Effective date: Effective as of the date of issuance and shall be
implemented within 60 days.
Amendment No. 220.
Renewed Facility Operating License No. DPR-23: The amendment
revises the technical specifications and facility operating license.
Date of initial notice in Federal Register: January 15, 2008 (73 FR
2548). 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.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602-1551.
NRC Branch Chief: Thomas H. Boyce.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: July 17, 2007, as supplemented
by letters dated August 7, 2007, and September 2, 2008.
Brief description of amendment: The amendment added a new license
condition (43) on the control room envelope habitability program,
revised Technical Specification (TS) requirements related to the
control room envelope habitability in TS 3.7.3, ``Control Room Fresh
Air (CRFA) System,'' and added the new TS 5.5.13, ``Control Room
Envelope Habitability Program.''
Date of issuance: October 14, 2008.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No: 178.
Facility Operating License No. NPF-29: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 25, 2007 (72
FR 54473). The supplemental letters dated August 7, 2007, and September
2, 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 October 14, 2008.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: December 5, 2007, as
supplemented by letters dated July 21 and August 28, 2008.
Brief description of amendment: The amendment changed Technical
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' to
add a reference to an analytical method that will be used to determine
core operating limits. The new reference, NEDC-33383P, ``GEXL97
Correlation Applicable to ATRIUM-10 Fuel,'' will allow the licensee to
use a Global Nuclear Fuel method to determine fuel assembly critical
power of AREVA ATRIUM-10 fuel. Additionally, the amendment made an
administrative change to an existing reference in TS 5.6.5.
Date of issuance: October 16, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 179.
[[Page 65702]]
Facility Operating License No. NPF-29: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 31, 2007 (72
FR 74358). The supplements dated July 21 and August 8, 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 October 16, 2008.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of application for amendment: December 27, 2007, as
supplemented by letter dated July 14, 2008.
Brief description of amendment: The amendment revised Technical
Specifications (TS) Section 3.4.1, ``RCS [Reactor Coolant System]
Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)
Limits,'' to increase the minimum RCS flow rate from 341,100 to 354,000
gallons per minute. The increased flow rate supports a new analysis of
a large break loss-of-coolant accident (LOCA). The new analysis is
performed using an NRC-approved methodology set forth in Westinghouse
Topical Report WCAP-16009-P-A, ``Realistic Large-Break LOCA Evaluation
Methodology Using the Automated Statistical Treatment of Uncertainty
Method (ASTRUM).'' This methodology will be endorsed and reflected by a
revision to TS Section 5.6.5, ``Core Operating Limits Report (COLR).''
Date of issuance: October 17, 2008.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 306.
Facility Operating License No. DPR-58: Amendment revised the
Renewed Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 29, 2008 (73 FR
5223). The supplement dated July 14, 2008, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staffs
original proposed no significant hazards consideration determination
published in the Federal Register on January 29, 2008.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 17, 2008.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: January 28, 2008, as supplemented by
letters dated July 28 and September 25, 2008.
Brief description of amendments: The amendments revised (1) Action
5 in Technical Specification (TS) 3.3.1, ``Reactor Trip
Instrumentation,'' for one inoperable channel of extended range neutron
flux instrumentation and (2) Action c in TS 3.4.1.4.2, ``Reactor
Coolant System, Cold Shutdown--Loops Not Filled.'' The amendments do
not complete the Nuclear Regulatory Commission staff's review of the
licensee's proposed TS changes in the application. The remaining
proposed TS changes to Action 5 will be addressed in a future letter to
the licensee.
Date of issuance: October 16, 2008.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1-187; Unit 2-174.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: March 25, 2008 (73 FR
15788). The supplemental letters dated July 28 and September 25, 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 October 16, 2008.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: October 26, 2007.
Brief description of amendment: The amendment revises the Technical
Specifications (TS) to adopt TS Task Force (TSTF) Change Traveler TSTF-
448, Revision 3, ``Control Room Envelope Habitability.''
Date of issuance: October 8, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 70.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications and License.
Date of initial notice in Federal Register: August 29, 2008 (73 FR
51014). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 8, 2008.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendment: October 24, 2007, as
supplemented by letter dated August 7, 2008.
Brief description of amendment: The amendments change Technical
Specifications (TSs) Limiting Condition for Operations (LCO) 3.8.7 and
3.8.9, pertaining to electrical power systems and distribution
associated with the 120 Volt AC vital bus inverters. The TS changes are
intended to support operability of components shared between Unit 1 and
Unit 2. The proposed changes will add new Conditions, Required Action
statements and Completion Times for LCO 3.8.7 and LCO 3.8.9 to address
shared components.
Date of issuance: October 9, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 253, 234.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
change the licenses and the technical specifications.
Date of initial notice in Federal Register: December 18, 2007 (72
FR 71717). The supplement dated August 7, 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 amendments is contained in a
Safety Evaluation dated October 9, 2008.
No significant hazards consideration comments received: No.
[[Page 65703]]
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement Or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/
adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr@nrc.gov.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, 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, and electronically on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to pdr@nrc.gov. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: ( 1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases
[[Page 65704]]
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
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\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
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Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for hearing or a petition for leave to intervene must be
filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve 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
HEARINGDOCKET@NRC.GOV, or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer\TM\ to
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ is free and is available
at http://www.nrc.gov/site-help/e-submittals/install-viewer.html.
Information about applying for a digital ID certificate is available on
NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/
apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC 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
[[Page 65705]]
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.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: October 13, 2008.
Description of amendment request: The amendment revised the
surveillance frequency for Technical Specification Surveillance
Requirement 3.8.1.10 for the endurance test conducted every 2 years on
the diesel generators.
Date of issuance: October 20, 2008.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 255.
Facility Operating License No. DPR-26: Amendment revises the
Technical Specifications and License.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. Public notice of the proposed amendment was
published in The Journal News newspaper, located in Westchester County,
New York on October 17 and October 18, 2008. The notice provided an
opportunity to submit comments on the Commission's proposed NSHC
determination. No comments have been received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated October 20, 2008.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Dated at Rockville, Maryland, this 24th day October 2008.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E8-25882 Filed 11-3-08; 8:45 am]
BILLING CODE 7590-01-P