[Federal Register Volume 73, Number 223 (Tuesday, November 18, 2008)]
[Notices]
[Pages 68451-68460]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-27110]
[[Page 68451]]
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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 23, 2008, to November 5, 2008. The
last biweekly notice was published on November 4, 2008 (73 FR 65685).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example, in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, 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
[[Page 68452]]
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
[email protected], or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer
TM to access the Electronic Information Exchange (EIE), a
component of the E-Filing system. The Workplace Forms Viewer
TM is free and is available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a
digital ID certificate is available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC technical help line,
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time,
Monday through Friday. The help line number is (800) 397-4209 or
locally, (301) 415-4737.
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, in accordance
with 10 CFR 2.302(g), with their initial paper filing requesting
authorization to continue to submit documents in paper format. Such
filings must be submitted by: (1) First class mail addressed to the
Office of the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff; or (2) courier, express mail, or expedited
delivery service to the Office of the Secretary, Sixteenth Floor, One
White Flint North, 11555 Rockville, Pike, Rockville, Maryland 20852,
Attention: Rulemaking and Adjudications Staff. Participants filing a
document in this manner are responsible for serving the document on all
other participants. Filing is considered complete by first-class mail
as of the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the petition
and/or request should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later
than 11:59 p.m. Eastern Time on the due date.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings. With respect to copyrighted
works, except for limited excerpts that serve the purpose of the
adjudicatory filings and would constitute a Fair Use application,
participants are requested not to include copyrighted materials in
their submission.
For further details with respect to this amendment action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: October 1, 2008.
[[Page 68453]]
Description of amendment request: The amendment would modify
Technical Specification (TS) 5.5.16, Containment Leakage Rate Testing
Program, by adding exceptions to Regulatory Guide (RG) 1.163,
``Performance-Based Containment Leak-Test Program,'' that would allow
the next integrated leak rate test (ILRT) (Type A test) to be performed
at a 15-year interval at Palo Verde Nuclear Generating Station (PVNGS),
Units 1, 2, and 3. The proposed amendment is risk-informed and follows
the guidance in RG 1.174, ``An Approach for Using Probabilistic Risk
Assessment in Risk-Informed Decisions on Plant-Specific Changes to the
Licensing Basis.''
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 extend the next ILRT interval from 10 to
15 years one time does not involve a physical change to PVNGS[,]
Units 1, 2, and 3, or a change in the manner in which the plant is
operated or controlled. The containment vessel is designed to
provide an essentially leak-tight barrier against the uncontrolled
release of radioactivity to the environment for any postulated
accidents. As such, the reactor containment itself and the testing
guidelines invoked to periodically demonstrate the integrity of the
containment exist to ensure the containment can mitigate the
consequences of any accident and do not involve the prevention or
identification of any precursors of any accidents. There is no
design basis accident that is initiated by a failure of the
containment leakage mitigation function. The extension of the ILRT
will not create any adverse interactions with other systems that
could result in initiation of a design basis accident. Therefore,
the probability of occurrence of an accident previously evaluated is
not significantly increased.
Based on a completed probability risk assessment of the affects
of this change to the ILRT interval there is a slight increase in
risk dose. This increase in risk in terms of person-rem year within
50 miles of the plant resulting from design basis accidents is
significantly less than one percent and of a magnitude that NUREG-
1493 indicates is imperceptible. The risk assessment also analyzed
the increase in risk in terms of the frequency of large early
releases from accidents. The increase in the large early release
frequency resulting from the proposed extension was determined to be
within the guidelines published in Regulatory Guide 1.174.
Additionally, the proposed change maintains defense-in-depth by
preserving a reasonable balance among prevention of core damage,
prevention of containment failure, and consequence mitigation. The
increase in the conditional containment failure probability from
reducing the ILRT frequency from one test per 10 years to one test
per 15 years is less than one percent and considered insignificant.
Continued containment integrity is assured by the history of
successful ILRTs, and the established programs for local leakage
rate testing and in-service inspections which are not affected by
the proposed change. Therefore, the consequences of an accident
previously analyzed are not significantly increased.
In summary, the probability of occurrence and the consequences
of an accident previously evaluated are not significantly increased.
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 extend the ILRT interval from 10 to 15
years does not create any new or different accident initiators or
precursors. The length of the ILRT interval does not affect the
manner in which any accident begins. The proposed change does not
physically change the plant, does not create any new failure modes
for the containment and does not affect the interaction between the
containment and any other system. Thus, the proposed changes do 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 risk-based margins of safety associated with the containment
ILRT are those associated with the estimated person-rem per year,
the large early release frequency, and the conditional containment
failure probability. The potential effect of the proposed change on
the parameters have been quantified and it has been determined that
the effect is considered insignificant. The non-risk-based margins
of safety associated with the containment ILRT are those involved
with its structural integrity and leak tightness. The proposed
change to extend the ILRT interval from 10 to 15 years does not
adversely affect either of these attributes. The proposed change
only affects the frequency at which these attributes are verified.
Therefore, the proposed change does not involve a significant
reduction in 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.
Entergy Nuclear Operations, Inc., Docket No. 50-155, Big Rock Point
Plant, Charleviox County, Michigan
Date of amendment request: September 22, 2008.
Description of amendment request: The proposed amendment would
amend the facility operating license by changing the names of the
licensees from Entergy Nuclear Operations, Inc., and Entergy Nuclear
Palisades, LLC to EquaGen Nuclear LLC and Enexus Nuclear Palisades,
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 associated order requirements. The proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated. The proposed
changes do not create the possibility of a new or different kind of
accident from an accident previously evaluated. The proposed changes
do not involve a significant reduction in 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 Nos. 50-003, 50-247, and 50-
286, Indian Point Nuclear Generating Unit Nos. 1, 2 and 3, Westchester
County, New York
Date of amendment request: September 30, 2008 (2 letters).
Description of amendment request: This is an administrative change
which would reflect the creation of new companies as approved by the
NRC Order dated July 28, 2008. The amendments would not be implemented
until the restructuring transactions have been completed. The
amendments would revise the names on the plant licenses to match the
names of the new companies. Entergy Nuclear Indian Point 2, LLC would
be changed to Enexus Nuclear Indian Point 2, LLC. Entergy Nuclear
Indian Point 3, LLC
[[Page 68454]]
would be changed to Enexus Nuclear Indian Point 3, LLC. Entergy Nuclear
Operations, Inc. would be changed to 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 associated order requirements. The proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated. The proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated. 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: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: September 22, 2008.
Description of amendment request: The proposed amendment would
amend the renewed facility operating license and Technical
Specifications Design Features, Section 4, by changing the names of the
licensees from Entergy Nuclear Operations, Inc. and Entergy Nuclear
Palisades, LLC to EquaGen Nuclear LLC and Enexus Nuclear Palisades,
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 associated order requirements. The proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated. The proposed
changes do not create the possibility of a new or different kind of
accident from an accident previously evaluated. The proposed changes
do not involve a significant reduction in 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 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
relocate the contents of the Vermont Yankee (VY) Technical
Specification (TS) relating to the Reactor Building crane to the VY
Technical Requirements Manual (TRM).
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 relocates the VY TS and associated Bases
related to the Reactor Building crane to the VY TRM. The proposed
amendment does not impact the operability of any structure, system
or component that affects the probability of an accident or that
supports mitigation of an accident previously evaluated. The
proposed amendment does not affect reactor operations or accident
analysis and has no radiological consequences. The operability
requirements for accident mitigation systems remain consistent with
the licensing and design basis. Therefore, the proposed amendment
does 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.
This proposed change relocates the VY TS and associated Bases
related to the Reactor Building crane to the VY TRM. 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.
This proposed change relocates the VY TS and associated Bases
related to the Reactor Building crane to the VY TRM. The proposed
amendment does not change the design or function of any component or
system. The proposed amendment does not involve any safety limits,
safety settings or safety margins. The ability of the Reactor
Building crane to perform its intended functions will continue to be
required in accordance with the VY TRM.
Since the proposed controls are adequate to ensure the
operability of the Reactor Building crane, there will still be high
assurance that the components are operable and capable of performing
their respective functions. 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 22, 2008.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) to change requirements related
to Battery Systems specified in TS Section 3.10 resulting in removing
the Limiting Condition for Operation pertaining to 345 kV switchyard
batteries, chargers and associated direct current distribution panel.
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 (SSC)
[[Page 68455]]
that affects the probability of an accident or that supports
mitigation or consequences of an accident previously evaluated. The
proposed change removes unnecessary information from the Technical
Specifications that is not required in accordance with 10 CFR 50.36.
The proposed change does not affect any plant equipment operation or
accident analysis and has no radiological consequences. 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 their function. The proposed change removes
unnecessary information from the Technical Specifications that is
not required in accordance with 10 CFR 50.36. As such, no new or
different types of equipment will be installed or removed from the
facility. Operation of existing installed equipment is unchanged.
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.
This change does not change any existing design or operational
requirements and does 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 operation 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.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: June 3, 2008.
Description of amendment request: The proposed amendment would
revise the analysis methodology in the Final Safety Analysis Report,
Section 5.4.3, ``Structural Design Criteria,'' and Section 5.4.5.3,
``Missile Analysis.'' The amendment would allow the licensee to use the
yield line theory methodology to qualify the east wall of the Auxiliary
Building for tornado wind and missile loading.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed LAR [license amendment request] will revise the
methodology used to qualify the east wall of the CR-3 [Crystal River
Unit 3 Nuclear Generating Plant] Auxiliary Building for all expected
and postulated loads including tornado wind and missile loading. The
Yield Line Theory methodology is an industry standard that is used
for the design and analysis of concrete slabs. The Yield Line Theory
methodology is used for investigating the failure mechanisms of flat
reinforced concrete slabs at the ultimate limit (failure point). In
other words, this methodology determines either the moments in a
slab at the point of failure or the load at which the slab will
fail. A change in the methodology of an analysis used to verify
qualification of an existing structure will not have any impact on
the probability of accidents previously evaluated.
The analysis performed demonstrates that the CR-3 Auxiliary
Building east wall will remain structurally intact following the
worst case loadings assumed in the calculation. Therefore, this
proposed change does not involve a significant increase in the
probability or consequences previously evaluated.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The function of the CR-3 Auxiliary Building wall is to house and
protect the equipment that is important to safety from damage during
normal operation, transients and design basis accidents. The use of
the Yield Line Theory methodology for qualifying the east wall of
the CR-3 Auxiliary Building has no impact on the capability of the
structure. A calculation that uses the Yield Line Theory methodology
demonstrated that the structure meets required design criteria. This
ensures that the wall is capable of performing its design function
without alteration or compensatory actions of any kind. No changes
to any plant system, structure, or component (SSC) are proposed. No
changes to any plant operating practices, procedures, computer
firmware/ software will occur.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does not involve a significant reduction in a margin on
safety.
The design basis of the plant requires structures to be capable
of withstanding normal and accident loads including those from a
design basis tornado. The Yield Line Theory methodology, as applied
in an approved plant calculation, has demonstrated that the east
wall of the CR-3 Auxiliary Building is capable of performing its
design function. There is a slight reduction in conservatism between
the method used for the remaining Class 1 structures, American
Concrete Institute (ACI) standard 318-63 and the Yield Line Theory
methodology, but the calculation performed with the Yield Line
Theory methodology validates the requirement that the east wall of
the CR-3 Auxiliary Building will protect the important to safety
SSCs located in proximity to the wall from damage.
ACI 318-63 utilizes conservative methods, and due to the
assumptions and technique, results in a Code defined value for
strength that is not the maximum limit. The Yield Line Theory
methodology uses assumptions and techniques that will define the
failure point. However, the calculation performed for the east wall
of the CR-3 Auxiliary Building demonstrates that there is margin to
this ``failure point,'' and the strength of the wall exceeds the
applied loads, including the tornado wind and pressure drop loads,
and will not fail due to tornado missile impact.
Therefore, the proposed 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: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: July 31, 2008.
Description of amendment request: The proposed amendments would
change the PPL Susquehanna, LLC (PPL) Units 1 and 2 Technical
Specification (TSs) 3.6.1.3 ``Primary Containment Isolation Valves
(PCIVs).'' It proposes to revise the Secondary Containment Bypass
Leakage (SCBL) limit in Surveillance Requirement 3.6.1.3.11 from ``less
than or equal to 9 standard cubic foot/feet per hour (scfh)'' to ``less
than or equal to 15 scfh when pressurized to greater than or equal to
Pa.''
Basis for proposed no significant hazards consideration
determination:
[[Page 68456]]
As required by 10 CFR 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The structures, systems and components affected by the proposed
change act as mitigators to the consequences of accidents. These
components are not initiators of any accident analyzed in the Final
Safety Analysis Report (FSAR). As such, the proposed change does not
increase the probability of an accident previously evaluated. Based
on the revised analysis, the proposed change does revise the
performance requirement; however, the proposed change does not
involve a revision to the parameters or conditions that could
contribute to the initiation of a DBA [design-basis accident]
discussed in Chapter 15 of the FSAR.
Plant-specific radiological analysis has been performed using
the increased Secondary Containment Bypass Leakage (SCBL) limit.
This analysis demonstrates that the CRHE [control room habitability
envelope] dose consequences meet the regulatory guidance provided
for use with the Alternative Source Term (AST), and the offsite
doses are well within acceptable limits (10 CFR 50.67, Regulatory
Guide (RG) 1.183, and Standard Review Plan Section (SRP) 15.0.1).
Therefore, the proposed amendment does not result in a
significant increase in the consequences of any previously evaluated
accident.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
any plant equipment. No new equipment is being introduced, and
installed equipment is not being operated in a new or different
manner. There are no setpoints, at which protective or mitigative
actions are initiated, affected by this change. This change does not
alter the manner in which equipment operation is initiated, nor will
the function demands on credited equipment be changed. No
alterations in the procedures that ensure the plant remains within
analyzed limits are being proposed, and no changes are being made to
the procedures relied upon to respond to an off-normal event as
described in the FSAR. As such, no new failure modes are being
introduced. The change does not alter assumptions made in the safety
analysis and licensing basis.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The results of the revised accident analysis are subject to the
acceptance criteria in 10 CFR 50.67. The revised Secondary
Containment Bypass Leakage rate limit is used in the LOCA [loss-of-
coolant accident] radiological analysis. The analysis has been
performed using conservative methodologies. Safety margins and
analytical conservatisms have been evaluated and have been found
acceptable. The analyzed LOCA event has been carefully selected and
margin has been retained to ensure that the analysis adequately
bounds postulated event scenarios. The dose consequences of the
limiting event is within the acceptance criteria presented in 10 CFR
50.67, RG 1.183, and SRP 15.0.1. The effect of the revision to the
Technical Specification requirements has been analyzed and doses
resulting from the pertinent design basis accident have been found
to remain within regulatory limits. The change continues to ensure
that the doses at the exclusion area and low population zone
boundaries, as well as the control room, are within the
corresponding regulatory limits. 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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief : Mark Kowal.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of amendment request: July 18, 2008.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) 2.1.1.2 to decrease the safety
limit minimum critical power ratio (SLMCPR) from 1.11 to 1.09 for
single recirculation loop operation and from 1.09 to 1.07 for two
recirculation loop operation.
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 Technical Specification change involve a
significant increase in the probability or consequences of an
accident previously evaluated?
No. The proposed amendment establishes a revised SLMCPR value
for single and two recirculation loop operation. The probability of
an evaluated accident is derived from the probabilities of the
individual precursors to that accident. The proposed SLMCPR values
preserve the existing margin to transition boiling and the
probability of fuel damage is not increased. Since the change does
not require any physical plant modifications or physically affect
any plant components, no individual precursors of an accident are
affected and the probability of an evaluated accident is not
increased by revising the SLMCPR values.
The consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. The revised SLMCPR values have been determined using
NRC-approved methods and procedures. The basis of the MCPR Safety
Limit is to ensure no mechanistic fuel damage is calculated to occur
if the limit is not violated. These calculations do not change the
method of operating the plant and have no effect on the consequences
of an evaluated accident. Therefore, the proposed TS change does not
involve an increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed Technical Specification change create the
possibility of a new or different kind of accident from any accident
previously evaluated?
No. The proposed license amendment involves a revision of the
SLMCPR value for single and two recirculation loop operation based
on the results of an analysis of the Unit 1 Cycle 8 core. Creation
of the possibility of a new or different kind of accident would
require the creation of one or more new precursors of that accident.
New accident precursors may be created by modifications of the plant
configuration, including changes in the allowable methods of
operating the facility. This proposed license amendment does not
involve any modifications of the plant configuration or changes in
the allowable methods of operation. 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 Technical Specification change involve a
significant reduction in a margin of safety?
No. The margin of safety as defined in the TS bases will remain
the same. The new SLMCPR values were calculated using referenced
fuel vendor methods and procedures, which are in accordance with the
fuel design and licensing criteria. The SLMCPR remains high enough
to ensure that greater than 99.9 percent of all fuel rods in the
core are expected to avoid transition boiling if the limit is not
violated, thereby preserving the fuel cladding integrity. Therefore,
the proposed TS change does not involve a 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
[[Page 68457]]
NRC Branch Chief: Thomas H. Boyce.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit No. 1, DeWitt County, Illinois
Date of application for amendment: September 27, 2007, as
supplemented by letter dated September 5, 2008.
Brief description of amendment: The amendment modified the
technical specifications (TS) by relocating references to specific
American Society for Testing and Materials standards for fuel oil
testing to licensee-controlled documents as part of the implementation
of Technical Specification Task Force (TSTF) Traveler No. 374. This
proposed change to the standard technical specifications was submitted
by the TSTF in TSTF-374, ``Revision to TS 5.5.13 and Associated TS
Bases for Diesel Fuel Oil,'' and is applicable to all nuclear power
reactors.
Date of issuance: October 30, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 182.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: December 18, 2007 (72
FR 71705). The September 5, 2008 supplement, contained clarifying
information, did not expand the scope of the application as originally
noticed, and did not change the NRC staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 30, 2008.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Power Station, Unit Nos. 2 and 3, New London County,
Connecticut
Date of application for amendment: March 25, 2008, as supplemented
by letter dated September 30, 2008.
Brief description of amendment: The amendment revises the reactor
coolant system (RCS) specific activity to utilize a new indicator, Dose
Equivalent Xenon-133 and only take into account the noble gas activity
in the primary coolant, instead of using the average disintegration
energy (E Bar). Specifically, the Technical Specification 3.4.8,
``Specific Activity,'' limit on RCS gross specific activity has a new
limit on RCS noble gas specific activity. The changes are based on
Technical Specification Task Force (TSTF) change traveler TSTF-490,
``Deletion of E Bar Definition and Revision to RCS Specific Activity
Tech. Spec. [Technical Specification].''
Date of issuance: October 27, 2008.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 307 and 246.
Renewed Facility Operating License Nos. DPR-65 and NPF-49:
Amendment revised the License and Technical Specifications.
Date of initial notice in Federal Register: July 29, 2008 (73 FR
43955-43956). The supplement dated September 30, 2008, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 27, 2008.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of application of amendments: October 16, 2007, as
supplemented May 7, September 2 and October 23, 2008.
Brief description of amendments: The amendments revised the
Technical Specifications to accommodate plant modifications that
address water hammer concerns described in Generic Letter 96-06,
``Assurance of Equipment Operability and Containment Integrity During
Design-Basis Conditions,'' dated September 30, 1996.
Date of Issuance: October 29, 2008.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 363, 365, 364.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: November 20, 2007 (72
FR 65364). The supplements dated May 7, September 2 and October 23,
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 29, 2008.
[[Page 68458]]
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of application of amendments: October 22, 2007, supplemented
July 14, September 17, and October 27, 2008.
Brief description of amendments: The amendments revised the
Technical Specifications related to accommodate the use of AREVA NP
Mark-B-HTP fuel.
Date of Issuance: October 29, 2008.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 362, 364, 363.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: November 20, 2007 (72
FR 65365). The supplements dated July 14, September 17, and October 27,
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 29, 2008.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: December 13, 2007, as
supplemented by letter dated July 10, 2008.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) by adding three Emergency Core Cooling System
(ECCS) valves and removing four ECCS valves from a TS surveillance
requirement for checking valve position every 7 days.
Date of issuance: October 29, 2008.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 256.
Facility Operating License No. DPR-26: The amendment revised the
License and the TSs.
Date of initial notice in Federal Register: March 25, 2008 (73 FR
15784). The July 10, 2008, supplement provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the NRC staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 29, 2008.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendment: October 19, 2007, supplemented
by letters dated March 14, 2008, March 26, 2008, and July 18, 2008.
Brief description of amendment: The amendments consist of changes
to the technical specifications of each unit, increasing the allowed
surveillance interval for local power range monitor calibrations from
1000 effective full power hours (EFPH) to 2000 EFPH.
Date of issuance: October 28, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 195 and 156.
Facility Operating License Nos. NPF-39 and NPF-85. These amendments
revised the license and the technical specifications.
Date of initial notice in Federal Register: July 8, 2008 (73 FR
39055). The supplements dated March 14, 2008, March 26, 2008 and July
18, 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 amendments is contained
in a Safety Evaluation dated October 28, 2008.
No significant hazards consideration comments received: No.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: April 4, 2008, as supplemented
by letter dated August 6, 2008.
Brief description of amendment: The amendment revised the Technical
Specifications by adding a new Limiting Condition for Operation (LCO),
LCO 3.0.9. This LCO establishes conditions under which systems would
remain operable when required physical barriers are not capable of
providing their related support function. This amendment is consistent
with approved Technical Specification Task Force (TSTF) Improved
Standard Technical Specifications Change Traveler, TSTF-427, Revision
2.
Effective date: As of the date of issuance and shall be implemented
within 90 days following startup from the 2009 Refueling Outage.
Amendment No.: 157.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 9, 2008 (73
FR 52418).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 22, 2008.
No significant hazards consideration comments received: None.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: October 29, 2007, as
supplemented by letters dated April 24 and June 13, 2008.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) for Prairie Island Nuclear Generating
Plants, Units 1 and 2. The amendments revise TS 3.8.1 ``AC Sources--
Operating'' by revising Surveillance Requirement 3.8.1.9 to require the
emergency diesel generator 24-hour load test be performed at or below a
power factor of 0.85.
Date of issuance: October 21, 2008.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 189, 178.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 18, 2007 (72
FR 71713). The supplemental letters contained clarifying information
and did not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in Safety Evaluation dated October 21, 2008.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: October 15, 2007, as
supplemented by letter dated July 8, 2008.
[[Page 68459]]
Brief description of amendments: The amendments relocate
surveillance frequencies of most surveillance tests from the Technical
Specifications (TS) to a licensee-controlled document, the Surveillance
Frequency Control Program (SFCP). Once relocated, changes to the
surveillance frequencies may be made using a risk-informed methodology,
Nuclear Energy Institute (NEI) document NEI 04-10 Rev. 1, as specified
in the Administrative Controls of the TS. The NRC staff has previously
approved NEI 04-10 Rev. 1, as acceptable for referencing in licensing
applications.
Date of issuance: October 30, 2008.
Effective date: As of its date of issuance and shall be implemented
within 360 days from the date of issuance.
Amendment Nos.: Unit 1-200; Unit 2-201.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: November 20, 2007 (72
FR 65370). The supplement dated July 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 amendments is contained
in a Safety Evaluation dated October 30, 2008.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: January 17, 2008, as
supplemented August 15, 2008.
Brief description of amendment: The amendment will strengthen the
control room envelope habitability requirements, adds a new
administrative controls program, and adds an additional condition as
described in Technical Specification Task Force traveler 448, Revision
3, ``Control Room Habitability.''
Date of issuance: October 27, 2008.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 180.
Renewed Facility Operating License No. NPF-12: Amendment revises
the Appendix A Technical Specifications and the Appendix C Additional
Conditions.
Date of initial notice in Federal Register: February 12, 2008 (73
FR 8071). The supplement dated August 18, 2008, provided clarifying
information that did not change the scope of the January 17, 2008,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 27, 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: October 23, 2007, as supplemented by
letter dated May 20, 2008.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) to relocate surveillance frequencies of
most surveillance tests from the TS to a licensee-controlled
surveillance frequency control program (SFCP). Once relocated, the
surveillance frequency changes are permitted based on the risk-informed
methodology as specified in the Administrative Controls section of the
TS.
Date of issuance: October 31, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 1-188; Unit 2-175.
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: December 18, 2007 (72
FR 71716). The supplemental letter dated May 20, 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 31, 2008.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Unit Nos. 1 and 2, Hamilton County, Tennessee
Date of amendment request: October 26, 2007.
Description of amendment request: The amendments modify the
Technical Specifications (TSs) to establish more effective and
appropriate action, surveillance, and administrative requirements
related to ensuring the habitability of the control room envelope in
accordance with NRC-approved Technical Specification Task Force (TSTF)
Standard Technical Specification change traveler TSTF-448, Revision 3,
``Control Room Habitability.'' Specifically, the amendments modify TS
3.7.7, ``Control Room Emergency Ventilation System'' and TS Section 6,
``Administrative Controls.'' The amendments also add a new license
condition regarding initial performance of the new surveillance and
assessment requirements of the revised TSs.
Date of issuance: October 28, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos: 321 and 313.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the license and the TSs.
Date of initial notice in Federal Register: December 4, 2007 (72 FR
68219).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 28, 2008.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: October 31, 2007, as
supplemented by letters dated February 21, March 7, April 17, May 6,
July 10, and August 13, 2008.
Brief description of amendment: The amendment revises Technical
Specifications to extend for one time the Completion Times for both
essential service water trains and the emergency diesel generators from
72 hours to 14 days. The revision to TS would apply when each train of
ESW system is inoperable during respective ESW system piping
replacements.
Date of issuance: October 31, 2008.
Effective date: As of its date of issuance and shall be implemented
by December 31, 2008.
Amendment No.: 186.
[[Page 68460]]
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 31, 2008 (72
FR 74362). The supplements dated February 21, March 7, April 17, May 6,
July 10, and August 13, 2008, provided additional information that
clarified the application did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 2008.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 6th day of November 2008.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E8-27110 Filed 11-17-08; 8:45 am]
BILLING CODE 7590-01-P