[Federal Register Volume 73, Number 185 (Tuesday, September 23, 2008)]
[Notices]
[Pages 54862-54874]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-21925]
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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 August 28, 2008 to September 10, 2008. The
last biweekly notice was published on September 9, 2008 (73 FR 52412).
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 01F21, 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
[[Page 54863]]
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
[email protected], or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM
to access the Electronic Information Exchange (EIE), a component of the
E-Filing system. The Workplace Forms 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
[[Page 54864]]
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].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: July 7, 2008
Description of amendments request: The proposed change would revise
Surveillance Requirement (SR) 3.6.1.6.1 to add a new requirement to
verify that each vacuum breaker is closed within 6 hours following an
operation that causes any of the vacuum breakers to open and revises SR
3.6.1.6.2 by removing the requirement to perform functional testing of
each vacuum breaker within 12 hours following an operation that causes
any of the vacuum breakers to open.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR Part 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not involve physical changes to any
plant structure, system, or component. The suppression chamber-to-
drywell vacuum breakers only provide an accident mitigation
function. As such, the probability of occurrence for a previously
analyzed accident is not impacted by the change to the surveillance
frequency for these components.
The consequences of a previously analyzed accident are dependent
on the initial conditions assumed for the analysis, the behavior of
the fuel during the analyzed accident, the availability and
successful functioning of the equipment assumed to operate in
response to the analyzed event, and the setpoints at which these
actions are initiated. No physical change to suppression chamber-to-
drywell vacuum breakers is being made as a result of the proposed
change, nor does the change alter the manner in which the vacuum
breakers operate during an accident. As a result, no new failure
modes of the suppression chamber-to-drywell vacuum breakers are
being introduced. The surveillance requirements for the suppression
chamber-to-drywell vacuum breakers will continue to ensure testing
of the suppression chamber-to-drywell vacuum breakers following
plant transients involving the discharge of steam to the suppression
chamber from the SRVs, and such testing will continue to provide
assurance that the vacuum breakers are able to perform their design
function. Based on this evaluation, there is no significant increase
in the consequences of a previously analyzed event.
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 the surveillance requirements for the
suppression chamber-to-drywell vacuum breakers does not involve any
physical alteration of plant systems, structures, or components. No
new or different equipment is being installed. No installed
equipment is being operated in a different manner. There is no
alteration to the parameters within which the plant is normally
operated or in the setpoints that initiate protective or mitigative
actions. As a result no new failure modes are being introduced.
Therefore, the proposed change to the surveillance requirements for
the suppression chamber-to-drywell vacuum breakers 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.
Response: No.
The proposed change revises Surveillance Requirement 3.6.1.6.1
to add a new requirement to verify each vacuum breaker is closed
within 6 hours following an operation that causes any of the vacuum
breakers to open and revises Surveillance Requirement 3.6.1.6.2 by
removing the requirement to perform functional testing of each
vacuum breaker within 12 hours following an operation that causes
any of the vacuum breakers to open. The operability and functional
characteristics of the suppression chamber-to-drywell vacuum
breakers remains unchanged. The margin of safety is established
through the design of the plant structures, systems, and components,
through the parameters within which the plant is operated, through
the establishment of the setpoints for the actuation of equipment
relied upon to respond to an event, and through margins contained
within the safety analyses. The proposed change to the surveillance
requirements for the suppression chamber-to-drywell vacuum breakers
does not impact the condition or performance of structures, systems,
setpoints, and components relied upon for accident mitigation. The
proposed change to Surveillance Requirements 3.6.1.6.1 and 3.6.1.6.2
will avoid unnecessary cycling and wear of the vacuum breaker test
actuation mechanisms, will improve the reliability of the vacuum
breakers, and will minimize the potential for a plant shut down due
to a problem with a vacuum breaker test actuating mechanism from
excessive wear. The proposed change does not impact any safety
analysis assumptions or results. Therefore, the proposed change does
not result in 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 proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Thomas H. Boyce.
Dominion Nuclear Connecticut, Inc. Docket Nos. 50-245, 50-336, and 50-
423, Millstone Power Station, Units 1, 2, and 3, New London County,
Connecticut
Date of amendment request: August 21, 2008.
Description of amendment request: The proposed amendment removes
references to and limits imposed by Nuclear Regulatory Commission
Generic Letter (GL) 82-12, ``Nuclear Power Plant Staff Working Hours,''
from the subject plants'' technical specifications (TS). The guidelines
have been superseded by the requirements of Title 10 of the Code of
Federal Regulations, Part 26 (10 CFR 26), Subpart I, ``Managing
Fatigue.''
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The removal of references to GL 82-12 will not remove the
requirement to control work hours and manage fatigue. Removal of TS
references to GL 82-12 will be performed concurrently with the
implementation of the more conservative 10 CFR 26, Subpart I,
requirements.
[[Page 54865]]
The proposed changes do not impact the physical configuration or
function of plant structures, systems, or components (SSCs) or the
manner in which SSCs are operated, maintained, modified, tested, or
inspected. The proposed changes do not impact the initiators or
assumptions of analyzed events, nor do they impact the mitigation of
accidents or transient events.
Because these new requirements are administrative in nature and
further, are more conservative with respect to work hour controls
and fatigue management, the proposed change will not significantly
increase the probability or consequence of an accident previously
evaluated.
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 changes remove references to GL 82-12 from TS
consistent with the recently revised Subpart I to 10 CFR 26. These
regulations are more restrictive than the current guidance and would
add conservatism to work hour controls and fatigue management. Work
hours will continue to be controlled in accordance with NRC
requirements. The new rule continues to allow for deviations from
controls to mitigate or prevent a condition adverse to safety or
necessary to maintain the security of the facility. This ensures
that the new rule will not restrict work hours at the expense of the
health and safety of the public as well as plant personnel.
The proposed changes do not alter plant configuration, require
that new plant equipment be installed, alter assumptions made about
accidents previously evaluated, add any initiators, or impact the
function of plant SSCs or the manner in which SSCs are operated,
maintained, modified, tested, or inspected.
Because the proposed changes do not remove the station's
requirement to control work hours and increases the conservatism of
work hour controls by changing administrative scheduling
requirements, 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.
Compliance with the new rule adds conservatism to existing
fatigue management and contributes to the margin of safety. Deletion
of references to GL 82-12 in the TS is administrative in nature
since fatigue management is controlled through the new rule. MPS1,
MPS2 and MPS3 will continue their fitness-for-duty and behavioral
observation programs, both of which will be strengthened by
compliance with the new rule. The proposed changes add conservatism
to fatigue management and contribute to the margin of safety.
The proposed changes do not involve any physical changes to
plant SSCs or the manner in which SSCs are operated, maintained,
modified, tested, or inspected. The proposed changes do not involve
a change to any safety limits, limiting safety system settings,
limiting conditions of operation, or design parameters for any SSC.
The proposed changes do not impact any safety analysis
assumptions and do not involve a change in initial conditions,
system response times, or other parameters affecting an accident
analysis.
Therefore, the proposed changes do 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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
NRC Branch Chief: Harold K. Chernoff.
Duke Energy Carolinas, LLC, Docket No. 50-269, Oconee Nuclear Station,
Unit1, Oconee County, South Carolina
Date of amendment request: June 26, 2008.
Description of amendment request: The proposed amendment would
result in a revision of the current licensing basis (LB) in regard to
high-energy line break (HELB) events occurring outside of containment
for Oconee Nuclear Station, Unit 1 (ONS-1).
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.
Justification: The ONS-1 changes proposed in this LAR [license
amendment request] include revisions to the current HELB methodology
and mitigation strategy as documented in a new HELB report. This
report provides the completed analysis for ONS HELBs including the
descriptions of the station modifications that have been or will be
made as a result of this comprehensive HELB reanalysis.
The modifications associated with the revised HELB LB will be
designed and installed in accordance with applicable quality
standards such that the likelihood of failure of new or modified
SSCs will not initiate failures, malfunctions, or inadvertent
operations of existing accident mitigating SSCs [structures,
systems, and components], such as the KHUs [Keowee hydro units], SSF
[standby shutdown facility], HPI [high-pressure injection], or the
Central Tie Switchyard 100 kV alternate power systems. For Turbine
Building HELBs that could adversely affect equipment needed to
stabilize and cooldown the units, the addition of the PSW [protected
service water] System provides added assurances that safe shutdown
can be readily established and maintained beyond the 72-hour SSF
mission time.
In conclusion, the changes will collectively enhance the
station's overall design, safety, and risk margin; therefore, the
proposed change does not involve a significant increase in the
probability or consequence 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.
Justification: The proposed modifications address potential
adverse consequences from a HELB outside of containment. These
modifications will be designed and installed in compliance with
applicable quality standards such that there are reasonable
assurances that they will neither introduce nor cause new failure
mechanisms, malfunctions or accident initiators not already
considered in the current HELB design and licensing basis.
The overall effect of the changes to the HELB LB is considered
an enhancement to the station's ability to achieve safe and cold
shut down following a damaging HELB; 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.
Justification: The revised HELB LB will collectively enhance the
station's overall design, safety, risk margin, and the station's
ability to mitigate a HELB event; therefore, the proposed change
does not involve a significant reduction in a margin of safety.
Based on the above, Duke concludes that the proposed amendment
does not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significance 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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Melanie C. Wong.
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 amendment request: June 26, 2008.
Description of amendment request: The proposed amendments would
result
[[Page 54866]]
in a revision to portions of the Updated Final Safety Analysis Report
(UFSAR) regarding the tornado licensing basis (LB).
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:
(4) Does the proposed amendment involve a significant increase
in the probability or consequences of an accident previously
evaluated?
Response: No.
Justification: Although a tornado does not constitute a
previously-evaluated UFSAR Chapter 15 design basis accident or
transient as described in 10 CFR 50.36(c)(2), it is a design basis
criterion that is required to be considered in plant equipment
design. The possibility of a tornado striking the ONS is
appropriately considered in the UFSAR and Duke has concluded that
the proposed changes do not increase the possibility that a damaging
tornado will strike the site or increase the consequences from a
damaging tornado.
The modifications associated with the revised tornado LB will be
designed and installed such that failures in these new or modified
SSCs [structures, systems, and components will not initiate failures
or inadvertent operations of existing ONS accident mitigating SSCs,
such as the KHUs [Keowee hydro units], SSF [standby shutdown
facility], or HPI [high-pressure injection] systems. The use of the
NRC-approved TORMIS methodology confirmed that the risk from missile
damage was acceptably low to vulnerable areas of the SSF structures
and other SSCs required for SSD [safe shutdown]. As a result, there
is reasonable assurance that a tornado missile will not prohibit the
SSF system from fulfilling its tornado LB or other functions.
Also, there are additional electrical power sources available
which provide increased assurance that systems used to transition
the units to SSD can be readily powered following a damaging
tornado. The PSW [protected service water] System will provide
additional assurance that SSD can be established and maintained.
Overall, the changes proposed will increase assurance that
potential challenges to the integrity of the RCS due to the effects
of a damaging tornado will not result in a radioactive release to
the environment. In conclusion, the changes will collectively
enhance the station's overall design, safety, and risk margin;
therefore, the probability or consequences of accidents previously
evaluated are not significantly increased.
(5) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Justification: Although only the SSF is credited for
establishing and maintaining SSDHR [secondary side decay heat
removal] and RCMU [reactor coolant makeup] during the first 72 hours
following a damaging tornado, there are two relatively independent,
diverse and redundant systems capable of safely shutting down all
three units in the revised LB (SSF and PSW). Other modifications
improve the ability of the SSF and PSW systems to perform their
functions following a damaging tornado. The modifications will be
designed and installed such that they will not introduce new failure
mechanisms, malfunctions or accident initiators not already
considered in the design and LB.
In conclusion, the changes to the tornado LB will not degrade
existing plant systems and will significantly enhance the station's
ability to achieve SSD following a damaging tornado. The design and
installation of the PSW system will be such that there is reasonable
assurance that the system, including new power paths, will not
contribute to the possibility of new or different kind of accident
from any accident previously evaluated.
(6) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Justification: The revised tornado LB will collectively enhance
the station's overall design, safety, and risk margin; 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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Melanie C. Wong.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: June 26, 2008, as supplemented by
letters dated August 4 and August 26, 2008.
Description of amendment request: The proposed amendments would
make changes to the Technical Specifications that are conforming or
related to a change in fuel type from Westinghouse 0.400-inch OD
Vantage+ fuel to Westinghouse 0.422-inch OD Vantage+ fuel.
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 requested amendment is related to a change in the reload
fuel design. The design criteria for the reload fuel are consistent
with those for the existing fuel and ensure that the reload fuel is
compatible on the basis of coolant flow and neutronic
characteristics, as well as DNB and peak cladding temperature
requirements. The reload fuel design also ensures mechanical
compatibility with the existing fuel, reactor core, control rods,
steam supply system, and fuel handling tools and system.
The reactor fuel and its analysis are not accident initiators.
Therefore, the change in reload fuel design does not affect accident
or transient initiation.
The minimum boron accumulator concentration is also not an
accident initiator. The proposed change to the minimum accumulator
boron concentration Technical Specification limit ensures that the
plant will continue to operate in a manner that provides acceptable
levels of protection for health and safety of the public. Further,
all design basis accidents and transients affected by the fuel
upgrade were re-analyzed or evaluated using representative core
designs and the results for each fuel type show all acceptance
criteria will continue to be met.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Use of the 422V+ fuel is consistent with current plant design
bases and does not adversely affect any fission product barrier, nor
does it alter the safety function of safety significant systems,
structures and components or their roles in accident prevention or
mitigation. The operational characteristics of 422V+ fuel are
bounded by the safety analyses * * *. The 422V+ fuel design performs
within existing fuel design limits.
The proposed change to the minimum accumulator boron
concentration Technical Specification limit ensures that the plant
will continue to operate in a manner that provides acceptable levels
of protection for health and safety of the public. Further, all
design basis accidents and transients affected by the fuel upgrade
were re-analyzed or evaluated using representative core designs and
the results for each fuel type show all acceptance criteria will
continue to be met.
No equipment additions or modifications are included with the
proposed change.
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 changes do not alter the manner in which applicable
design basis limits are determined, nor do they result in exceeding
existing design basis limits. Thus, all licensed safety margins are
maintained.
[[Page 54867]]
Therefore, the proposed changes do 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 requests involve no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Lois M. James.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, San
Diego County, California
Date of amendment request: June 27, 2008.
Description of amendment request: These proposed changes consist of
Proposed Change Number 583 (PCN-583) and are in support of the
replacement of the steam generators (SGs) at SONGS Units 2 and 3. The
proposed changes reflect revised SG inspection and repair requirements,
and revised peak containment post-accident pressure resulting from
installation of the replacement SGs.
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 changes will reflect installation of Replacement
Steam Generators (RSGs) at San Onofre Nuclear Generating Station
(SONGS) Units 2 and 3. The proposed changes involve revising the
Steam Generator (SG) tube inspection and repair [requirements] and
revising the peak containment post-accident pressure.
The proposed change to revise the SG tube inspection and repair
[requirements] affect Technical Specifications (TSs) 3.4.17, ``Steam
Generator (SG) Tube Integrity,'' 5.5.2.11, ``Steam Generator (SG)
Program,'' and 5.7.2.c, ``Special Reports.'' The proposed TS 3.4.17,
5.5.2.11, and 5.7.2.c revisions remove the repair method (sleeving),
and Alternate Repair Criteria (ARC). The revisions replace the 44%
tube repair criterion applicable to the original SGs, with a 35%
(preliminary) tube repair criterion applicable to the RSGs. The
revisions replace inspection requirements applicable to the tubing
material of the original SGs with inspection requirements applicable
to the tubing material of the RSGs, thus maintaining consistency
with applicable material-specific regulatory guidance (TSTF-449,
Revision 4). Overall, these revisions will ensure that all RSG tubes
found by inservice inspection to contain flaws with a depth equal to
or exceeding 35% (preliminary) of the nominal tube wall thickness
will be plugged as required by revised TS 5.5.2.11.c.1.
The TS 5.5.2.11.b SG structural integrity, accident induced
leakage, and operational leakage performance criteria are unchanged
and will continue to be met for the RSGs. Meeting the SG performance
criteria provides reasonable assurance that the SG tubing will
remain capable of maintaining reactor coolant pressure boundary
integrity throughout each operating cycle and in the unlikely event
of a design basis accident.
The proposed change to the SG tube inspection and repair
[requirements] will not affect the probability of any accident
initiators. There will be no degradation in the performance of, or
an increase in the number of challenges imposed on, safety-related
equipment assumed to function during an accident. There will be no
change to accident mitigation performance. The proposed change will
not alter any assumptions or change any mitigation actions in the
radiological consequence evaluations in the Updated Final Safety
Analysis Report (UFSAR).
The proposed change to the peak containment post-accident
pressure will revise TS 5.5.2.15, ``Containment Leakage Rate Testing
Program,'' by changing the stated values for peak containment
internal pressure for the design-basis Loss-of-Coolant Accident
(LOCA) and Main Steam Line Break (MSLB) accidents. The current LOCA
value of 45.9 psig would be changed to 48.0 psig and the current
MSLB value of 56.5 psig would be changed to 51.5 psig.
The proposed change does not affect the probability of
occurrence of an accident previously evaluated because it relates
solely to the consequences of hypothesized accidents given that the
accident has already occurred.
The proposed change increases the calculated peak containment
internal pressure for the LOCA events from 45.9 psig to 48.0 psig.
The revised post-LOCA peak containment pressure is bounded by the
existing and revised post-MSLB peak containment pressure and the
containment design pressure of 60 psig. Despite the increase in the
post-LOCA peak containment pressure, any post-accident containment
leakage will still be limited to less than 0.1% containment air
volume per day, consistent with current TS 5.5.2.15. Therefore,
there is no increase in the radiological consequences of a LOCA as a
result of the change to the post-LOCA peak containment pressure.
The post-MSLB peak containment pressure decreases from 56.5 psig
to 51.5 psig. Thus, the peak containment post-accident pressure is
decreased as a result of this change, and there is no resulting
increase in the consequences of a previously evaluated accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
[Response: No.]
The proposed change to the SG tube inspection and repair
[requirements] deletes the repair method (sleeving) and the ARC
applicable to the original SGs, and provides repair criteria and
inspection requirements applicable to the RSGs. This will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
primary-to-secondary leakage that may be experienced during all
plant conditions will be monitored to ensure it remains within
current accident analysis assumptions. The proposed change does not
adversely affect the method of operation of the SGs or the primary
or secondary coolant chemistry controls and does not impact other
plant systems or components.
The proposed change to the peak containment post-accident
pressure relates to two accidents, LOCA and MSLB, which are already
evaluated in the Updated Final Safety Analysis Report (UFSAR).
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.
For the proposed change to the SG inspection and repair
[requirements], the safety function of the SGs is maintained by
ensuring the integrity of the tubes. SG tube integrity is a function
of the design, environment, and the physical condition of the SG
tubes. The proposed change, which deletes the repair method
(sleeving) and the ARC applicable to the original SGs, and provides
repair criteria and inspection requirements applicable to the RSGs,
does not adversely affect the SG tube design or operating
environment. SG tube integrity will continue to be maintained by
implementing the TS 5.5.2.11 SG Program to manage SG tube
inspection, assessment, and plugging. The requirements established
by the TS 5.5.2.11 SG Program are consistent with those in the
applicable design codes and standards.
For the change to the peak containment post-accident pressure,
the proposed change increases the calculated peak containment
internal pressure for the LOCA events from 45.9 psig to 48.0 psig.
The revised post-LOCA peak containment pressure is bounded by the
existing and revised post-MSLB peak containment pressure. The post-
MSLB peak containment pressure decreases from 56.5 psig to 51.5
psig. The proposed peak containment internal pressure for the MSLB
accident is less than the containment design pressure of 60 psig and
less than the previously calculated pressure.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
Based on the above, SCE concludes that the proposed amendments
present no significant hazards consideration under the standards set
forth in 10 CFR 50.92(c), and accordingly,
[[Page 54868]]
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 requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: Michael T. Markley.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3 (SONGS 2 and
3), San Diego County, California
Date of amendment request: June 27, 2008.
Description of amendment request: SONGS Units 2 and 3 requests
adoption of an approved change to the standard technical specifications
(STS) for Combustion Engineering Pressurized Water Reactor (PWR) Plants
(NUREG-1432) and plant-specific technical specifications (TS), to allow
replacing the departure from nucleate boiling (DNB) parameter limits
with references to the core operating limits report (COLR) in
accordance with Generic Letter 88-16, ``Removal of Cycle Specific
Parameter Limits from Technical Specifications,'' dated October 4,
1988. The changes are consistent with NRC approved Industry/Technical
Specification Task Force (TSTF) Standard Technical Specification Change
Traveler, TSTF-487, Revision 1, using the consolidated line-item
improvement process (CLIIP).
The NRC staff issued a notice of availability in the Federal
Register on June 5, 2007 (72 FR 31108), including a model safety
evaluation and model no significant hazards consideration (NSHC)
determination, using the CLIIP process. The licensee affirmed the
applicability of the model NSHC determination in its application dated
June 27, 2008.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1: Does the Proposed Change Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated?
Response: No.
The proposed amendment replaces the limit values of the reactor
coolant system (RCS) DNB parameters (i.e., pressurizer pressure, RCS
cold leg temperature, and RCS flow rate) in TS with references to
the COLR, in accordance with the guidance of Generic Letter 88-16,
to allow these parameter limit values to be recalculated without a
license amendment. The proposed amendment does not involve operation
of any required structures, systems, or components (SSCs) in a
manner or configuration different from those previously recognized
or evaluated. The cycle-specific values in the COLR must be
calculated using the NRC-approved methodologies listed in TS 5.6.3,
``Core Operating Limits Report (COLR).'' Replacing the RCS DNB
parameter limits in TS with references to the COLR will maintain
existing operating fuel cycle analysis requirements. Because these
parameter limits are determined using the NRC approved
methodologies, the acceptance criteria established for the safety
analyses of various transients and accidents will continue to be
met. Therefore, neither the probability nor consequences of any
accident previously evaluated will be increased by the proposed
change.
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 preciously evaluated.
Criterion 2: Does the Proposed Change Create the Possibility of
a New or Different Kind of Accident from any Previously Evaluated?
Response: No.
The proposed amendment to replace the RCS DNB parameter limits
in TS with references to the COLR does not involve a physical
alteration of the plant, nor a change or addition of a system
function. 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 proposed change. Therefore, the proposed
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
Criterion 3: Does the Proposed Change Involve a Significant
Reduction in the Margin of Safety?
Response: No.
The proposed amendment to replace the RCS DNB parameter limits
in TS with references to the COLR will continue to maintain the
margin of safety. The DNB parameter limits specified in the COLR
will be determined based on the safety analyses of transients and
accidents, performed using the NRC-approved methodologies that show
that, with appropriate measurement uncertainties of these parameters
accounted for, the acceptance criteria for each of the analyzed
transients are met. This provides the same margin of safety as the
limit values currently specified in the TS. Any future revisions to
the safety analyses that require prior NRC approval are identified
per the 10 CFR 50.59 review process.
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 requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: Michael T. Markley.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the 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.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendments: October 26, 2007.
Brief description of amendments: The proposed amendment would
revise the Technical Specification requirements related to control room
envelope habitability in accordance with the NRC-approved Revision 3 of
Technical Specification Task Force (TSTF) Standard Technical
Specifications Change Traveler TSTF-448, ``Control Room Habitability.''
Date of publication of individual notice in the Federal Register:
August 29, 2008 (73 FR 51014).
Expiration date of individual notice: September 29, 2008 (Public
Comments) and October 28, 2008 (Requests for Hearing).
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
[[Page 54869]]
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].
Duke Energy Carolinas, LLC, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: September 27, 2007.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) TS 3.7.2, ``Main Steam Isolation
Valves,'' and TS 3.7.3, ``Main Feedwater Isolation Valves, Main
Feedwater Control Valves, Associated Bypass Valves and Tempering
Valves,'' by removing the specific isolation time for the isolation
valves from the associated surveillance requirements.
Date of issuance: September 8, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 244 and 238.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the licenses and the technical specifications.
Date of initial notice in Federal Register: February 26, 2008 (73
FR 10 10297).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 8, 2008.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: March 13, 2008.
Brief description of amendment: The amendment replaces the current
Arkansas Nuclear One, Unit No. 2 (ANO-2) TS 3.4.8, ``RCS [reactor
coolant system] Specific Activity,'' limit on RCS gross specific
activity with a new limit on RCS noble gas specific activity. The noble
gas specific activity limit would be based on a new dose equivalent Xe-
133 (DEX) definition that would replace the current E Bar average
disintegration energy definition. In addition, the current dose
equivalent I-131 (DEI) definition would be revised to allow the use of
additional thyroid dose conversion factors (DCFs). This request adopted
Technical Specification Task Force (TSTF) change traveler TSTF-490,
Revision 0, ``Deletion of E Bar Definition and Revision to RCS [reactor
coolant system] Specific Activity Technical Specification'' (Agencywide
Documents Access and Management System Accession No. ML052630462), for
pressurized water reactor Standard Technical Specifications (STS) for
ANO-2.
Date of issuance: September 8, 2008.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: Unit 2-282.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications and license.
Date of initial notice in Federal Register: May 6, 2008 (73 FR
25039).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 8, 2008.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: October 22, 2007, as
supplemented by letters dated April 22, and July 8, 2008.
Brief description of amendment: The amendment revises Technical
Specifications (TS) Limiting Condition for Operation (LCO) 3.0.4 and
Surveillance Requirement (SR) 4.0.4 to adopt the provisions of
Industry/TS Task Force (TSTF) change TSTF-359, ``Increased Flexibility
in Mode Restraints.'' This operating license improvement was made
available by the U.S. Nuclear Regulatory Commission (NRC) on April 4,
2003, as part of the consolidated line item improvement process. The
proposed TS changes also include an additional application of LCO
3.0.4.c for TS 3.4.3, ``Pressurizer Spray Valves.''
Date of issuance: August 28, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: Unit 2-281.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: December 18, 2007 (72
FR 71710). The supplements dated April 22, and 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 amendment is contained in a Safety Evaluation
dated August 28, 2008.
No significant hazards consideration comments received: No.
[[Page 54870]]
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket No. 50-352 and No. 50-353,
Limerick Generating Station, Unit 1 and 2, Montgomery County,
Pennsylvania
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendments: August 8, 2007.
Brief description of amendments: The amendment replaces references
to Section XI of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code with references to the ASME Code for
Operation and Maintenance of Nuclear Power Plants (OM Code) in the
applicable technical specification (TS) section for the Inservice
Testing Program (IST) for the Exelon Generation Company, LLC, and
AmerGen Energy Company, LLC, plants that have implemented industry
Improved Technical Specifications. The changes are based on Technical
Specification Task Force (TSTF) 479, Revision 0, ``Changes to Reflect
Revision of 10 CFR 50.55a.'' For all units except Oyster Creek and TMI-
1, the amendments also incorporate TSTF-497, Revision 0, ``Limit
Inservice Testing Program SR [Surveillance Requirement] 3.0.2
Application to Frequencies of 2 Years or Less,'' which adds a provision
in the applicable TS section to only apply the extension allowance of
SR 3.0.2 to the frequency table listed in the TS as part of the IST
program and to normal and accelerated inservice testing frequencies of
two years or less, as applicable.
Date of issuance: August 28, 2008.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 153, 153, 157, 157, 229, 222, 194, 155, 268, 268,
272, 241, 236 and 266.
Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66,
DPR-19, DPR-25, NPF-39, NPF-85, DPR-16, DPR-44, DPR-56, DPR-29, DPR-30,
and DPR-50: The amendments revised the Technical Specifications/
Licenses.
Date of initial notice in Federal Register: December 4, 2007 (72 FR
68213). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 28, 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: August 24, 2007, supplemented by
letter dated June 11, 2008.
Brief description of amendment: The amendments consist of changes
to the technical specifications of each unit, increasing the minimum
required volume of fuel oil in the emergency diesel generator day tanks
from 200 gallons to 250 gallons.
Date of issuance: August 27, 2008.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 193 and 154.
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: June 20, 2008 (73 FR
35168). The NRC staff's original proposed no significant hazards
determination was based on the supplement dated June 11, 2008. The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated August 27, 2008.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and
3, York and Lancaster Counties, Pennsylvania
Date of application for amendments: July 13, 2007, as supplemented
on February 28, 2008, March 28, 2008, April 17, 2008, May 23, 2008,
July 29, 2008, August 7, 2008, and August 21, 2008.
Brief description of amendments: The amendments modify the
Technical Specifications to support application of Alternative Source
Term (AST) methodology at PBAPS Units 2 and 3. The fission product
release from the reactor core into containment is referred to as the
``source term,'' and is characterized by the composition and magnitude
of the radioactive material, the chemical and physical properties of
the material, and the timing of the release from the reactor core as
discussed in Technical Information Document (TID) 14844, ``Calculation
of Distance Factors for Power and Test Reactor Sites.'' Since the
publication of TID 14844, advances have been made in understanding the
composition and magnitude, chemical form, and timing of fission product
releases from severe nuclear power plant accidents. In light of these
insights, NUREG-1465, ``Accident Source Terms for Light-Water Nuclear
Power Plants,'' was published in 1995 with revised ASTs for use in the
licensing of future light-water reactors.
The Nuclear Regulatory Commission (NRC), in Title 10 of the Code of
Federal Regulations, Section 50.67 (10 CFR 50.67), ``Accident source
term,'' subsequently allowed the use of the ASTs described in NUREG-
1465 at operating plants. This request to apply the AST methodology is
made in accordance with 10 CFR 50.67, with the exception that TID 14844
will continue to be used as the radiation dose basis for equipment
qualification at PBAPS Units 2 and 3. Application of the AST
methodology at PBAPS Units 2 and 3 requires that radiation dose limits
specified in 10 CFR 50.67 are adhered to for the exclusion area
boundary, the low population zone outer boundary, and the facility
control room.
Date of issuance: September 5, 2008.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 269 and 273.
Renewed Facility Operating License Nos. DPR-44 and DPR-56:
Amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: May 6, 2008 (73 FR
25040). The supplements dated February 28, 2008, March 28, 2008, April
17, 2008, May 23, 2008, July 29, 2008, August 7, 2008, and August 21,
2008, clarified the application, did not expand the scope of the
application as originally noticed, and did not change the initial
proposed
[[Page 54871]]
no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 5, 2008.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: February 20, 2008.
Brief description of amendment: This amendment revised an
Applicability footnote in Technical Specification (TS) Table 3.3.2.1-1,
``Control Rod Block Instrumentation,'' to permit use of an improved
optional Banked Position Withdrawal Sequence (BPWS) reactor shutdown
process. Corresponding changes are in accordance with the Bases of TS
3.1.6, ``Control Rod Pattern,'' and the Bases of TS 3.3.2.1, to
reference the new BPWS shutdown method. This amendment is consistent
with Technical Specification Task Force (TSTF) Traveler TSTF-476-A,
Revision 1, ``Improved BPWS Control Rod Insertion Process (NEDO-
33091),'' and the Consolidated Line Item Improvement Process Notice of
Availability dated May 23, 2007.
Date of issuance: August 28, 2008.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 150.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: April 22, 2008 (73 FR
21659).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 28, 2008.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1 (NMP1), Oswego County, New York
Date of application for amendment: September 27, 2007, as
supplemented by letter dated June 5, 2008.
Brief description of amendment: The amendment changes the NMP1
Technical Specifications (TSs) by revising the operability requirements
contained in TS Section 3.2.7, ``Reactor Coolant System Isolation
Valves,'' and associated requirements contained in TS Section 3.6.2,
``Protective Instrumentation.'' The amendment will modify the
conditions for which reactor coolant system isolation valves (RCSIVs)
and associated isolation instrumentation must be operable to include
the hot shutdown reactor operating condition. In addition, it will be
required that the RCSIVs in the shutdown cooling (SDC) system and
associated isolation instrumentation be operable during the cold
shutdown reactor operating condition and the refueling reactor
operating condition. Lastly, TS Section 3.6.2 (Table 3.6.2b) will be
revised to delete unnecessary operability requirements for the cleanup
system and SDC system high area temperature isolation instrumentation,
consistent with the proposed revisions to the RCSIV operability
requirements.
Date of issuance: August 27, 2008.
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment No.: 197.
Renewed Facility Operating License No. DPR-63: Amendment revised
the License and TSs.
Date of initial notice in Federal Register: November 20, 2007 (72
FR 65367). The supplement dated June 5, 2008, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the Nuclear
Regulatory Commission staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 27, 2008.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: November 5, 2007, as supplemented April
7, 2008.
Brief description of amendment request: TS Section 5.5.17,
``Containment Leakage Rate Testing Program,'' is changed to resolve a
timing conflict between the FNP, Unit 2 R20 refueling outage schedule
and the 15-year test date for the FNP, Unit 2 Type A Containment
Integrated Leak Rate Test (ILRT). Although Unit 1 does not have a
current timing conflict, a similar Unit 1 change was requested for
consistency. The change adds approximately 1 month to the previously
approved required date.
Date of issuance: September 2, 2008.
Effective Date: As of its date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1-177; Unit 2-170.
Facility Operating License Nos. NPF-2 and NPF-8: The amendment
revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 29, 2008 (73 FR
5229).
The supplement dated April 7, 2008, provided clarifying information
that did not change the scope of the application or the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 2, 2008.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: August 29, 2006, as
supplemented November 6, November 27, 2006, January 30, June 22, July
16, August 13, October 18, December 11, 2007, January 24, February 4,
February 25 (two letters, nos. 1389 and 0175), February 27, March 13,
April 1, May 5, June 25, July 2, July 14, and August 14, 2008.
Brief description of amendments: The amendments revise the
licensing basis with a full scope implementation of an alternative
source term (AST) for HNP.
Date of issuance: August 28, 2008.
Effective date: As of the date of issuance and shall be implemented
by May 31, 2012 for Hatch Unit 1 and by May 31, 2011, for Hatch Unit 2.
Amendment Nos.: Unit 1-256, Unit 2-200.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: July 23, 2008 (73 FR
42834).
The supplement dated August 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 staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated August 28, 2008.
No significant hazards consideration comments received: No.
[[Page 54872]]
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 [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, 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 [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases
[[Page 54873]]
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.
---------------------------------------------------------------------------
\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.
---------------------------------------------------------------------------
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
[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
[[Page 54874]]
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.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: August 26, 2008, as supplemented on
August 28, 2008.
Description of amendment request: The amendments revise Functional
Unit 6.f of Table 3.3-3, ``Engineered Safety Feature Actuation System
Instrumentation,'' modifying the mode of applicability with two
footnotes. The first footnote indicates that the auxiliary feedwater
(AFW) auto-start function associated with the trip of main feedwater
(MFW) pumps in Mode 2 is only required when one or more MFW pumps are
supplying feedwater to the steam generators. The second footnote, which
annotates the minimum channels operable column for Functional Unit 6.f
of TS Table 3.3-3, indicates that one channel may be inoperable during
Mode 1 for up to 4 hours when starting up or shutting down a MFW pump.
Functional Unit 6.f of technical specification Table 3.3-3 is an
anticipatory trip function that provides early actuation of the AFW
system.
Date of issuance: August 29, 2008.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos: 319 and 312.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, state consultation, and
final NSHC determination are contained in a safety evaluation dated
August 29, 2008.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Thomas H. Boyce.
Dated at Rockville, Maryland, this 11th day of September 2008.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E8-21925 Filed 9-22-08; 8:45 am]
BILLING CODE 7590-01-P