[Federal Register Volume 74, Number 192 (Tuesday, October 6, 2009)]
[Notices]
[Pages 51326-51339]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-23780]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0433]
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) 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
[[Page 51327]]
hazards consideration, notwithstanding the pendency before the
Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from September 10, 2009, to September 23, 2009.
The last biweekly notice was published on September 22, 2009 (74 FR
48316).
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 Title 10 of the Code of Federal
Regulations (10 CFR), Section 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
and Directives Branch (RDB), TWB-05-B01M, Division of Administrative
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be faxed to the RDB at 301-492-3446. Documents may be examined, and/or
copied for a fee, at the NRC's Public Document Room (PDR), located at
One White Flint North, Public File Area O1F21, 11555 Rockville Pike
(first floor), Rockville, Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area O1F21, 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 by the above date, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
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, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, any hearing held
would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other
[[Page 51328]]
document filed in the proceeding prior to the submission of a request
for hearing or petition to intervene, and documents filed by interested
governmental entities participating under 10 CFR 2.315(c), must be
filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve all adjudicatory documents
over the internet, or in some cases to mail copies on electronic
storage media. Participants may not submit paper copies of their
filings unless they seek an exemption in accordance with the procedures
described below.
To comply with the procedural requirements of E-Filing, at least
ten (10) days prior to the filing deadline, the petitioner/requestor
should contact the Office of the Secretary by e-mail at
[email protected], or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM
to access the Electronic Information Exchange (EIE), a component of the
E-Filing system. The Workplace Forms ViewerTM is free and is
available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a digital ID certificate is
available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory e-
filing system 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 Meta-System Help Desk, which is
available between 8 a.m. and 8 p.m., Eastern Time, Monday through
Friday, excluding government holidays. The Meta-System Help Desk can be
contacted by telephone at 1-866-672-7640 or by e-mail at
[email protected].
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, 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 request and/
or petition should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii). Documents submitted in adjudicatory proceedings
will appear in NRC's electronic hearing docket which is available to
the public at http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless
excluded pursuant to an order of the Commission, an Atomic Safety and
Licensing Board, or a Presiding Officer. Participants are requested not
to include personal privacy information, such as social security
numbers, home addresses, or home phone numbers in their filings, unless
an NRC regulation or other law requires submission of such information.
With respect to copyrighted works, except for limited excerpts that
serve the purpose of the adjudicatory filings and would constitute a
Fair Use application, participants are requested not to include
copyrighted materials in their submissions.
For further details with respect to this license amendment
application, 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 O1F21, 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. Persons who do not have
access to ADAMS or who encounter problems in accessing the documents
located in ADAMS, should contact the NRC PDR Reference staff at 1-800-
397-4209, 301-415-4737, or by email to [email protected].
Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling Water
Reactor, Genoa, Wisconsin (TAC J00359)
Date of amendment request: July 28, 2009.
Description of amendment request: The amendment application
proposes changes to Technical Specifications, in support of the dry
cask storage project at La Crosse Boiling Water Reactor. The
application specifically proposes lower Fuel Element Storage Well water
level limits and proposes changes to the definition of ``fuel
handling.''
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?
No.
The proposed change to the definition of FUEL HANDLING is an
administrative
[[Page 51329]]
clarification and does not affect the operation of the plant or the
postulated accidents in any way. The proposed changes to allow lower
Fuel Element Storage Well (FESW) water level limits do not alter the
manner in which individual fuel assemblies are moved or alter the
design function of the FESW or any other structures, systems, and
components used to ensure safe fuel storage. The total number of
fuel assembly moves to the Dry Cask Storage System is exactly the
same as that contemplated during original plant design when fuel was
assumed to be transported from the plant directly to a disposal
site. All of the accidents previously evaluated in the La Crosse
Boiling Water Reactor (LACBWR) Decommissioning Plan have been
reviewed for impact as a result of the proposed water level changes.
The proposed changes do not affect the plant in such a manner that
the likelihood or consequences of any previously evaluated accident
is increased.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No.
The proposed change to the definition of FUEL HANDLING is an
administrative clarification and does not affect the operation of
the plant in any way. The proposed changes to allow lower FESW water
level limits do not alter the manner in which individual fuel
assemblies are moved; or alter the design function of the FESW or
any other structures, systems, and components used to ensure safe
fuel storage. All of the accidents previously evaluated in the
LACBWR Decommissioning Plan have been reviewed for impact as a
result of the proposed water level changes. The existing accidents
remain applicable and bounding for the LACBWR facility with the
proposed changes in place and do not affect the plant in such a
manner that a new accident has been created.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety? No.
The proposed change to the definition of FUEL HANDLING is an
administrative clarification and does not affect plant operation or
safety margins in any way. The proposed changes to allow lower FESW
water level limits do not alter the manner in which individual fuel
assemblies are moved; or alter the design function of the FESW or
any other structures, systems, and components used to ensure safe
fuel storage. All of the accidents previously evaluated in the
LACBWR Decommissioning Plan have been reviewed for impact as a
result of the proposed water level changes. The likelihood and
consequences of previously evaluated accidents remain applicable and
bounding with the proposed changes in place; thus, safety margins
remain the same.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The U.S. Nuclear Regulatory Commission (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.
NRC Branch Chief: Andrew Persinko.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: July 23, 2009.
Description of amendment request: The proposed amendment would
remove the level indicating instrument from the Technical Specification
Surveillance Requirement (SR) for the refueling water storage tank, but
leave the low level alarm function in the SR.
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?
No. The proposed change revises the existing Indian Point 3
Refueling Water Storage Tank (RWST) Technical Specification (TS)
Surveillance Requirement (SR) to remove the level indication
function for the L-921 instrument loop. Removal of a TS SR for the
level indication does not increase the probability of an accident
occurring since it is not an accident initiator and does not
increase the consequences of an accident since it is not performing
any mitigating function and is not a post accident instrument. The
proposed revision will not affect RWST lo-lo level alarm function
used for operator guidance to begin sequencing to Recirculation Mode
of Safety Injection during a postulated loss of coolant accident
(LOCA). There will be no change in equipment qualification
requirements or changes to the surveillance requirement for the lo-
lo level alarm. 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?
No. The proposed change removes the RWST level indication
function from the RWST lo-lo level alarm surveillance requirement
for the L-921 instrument loop. The proposed change does not involve
installation of new equipment or modification of existing equipment,
so that no new equipment failure modes are introduced. Also, the
proposed change does not result in a change to the way that the
equipment or facility is operated so that no new accident initiators
are created. 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?
No. The proposed change removes the RWST level indication
function from the RWST io-lo level alarm surveillance requirement
for the L-921 instrument loop. There is no change to the design
requirements or the surveillance interval. The proposed change does
not add the level indicating function elsewhere in the TS because it
is a local level indication that is only used during normal
operation and was never a post accident monitoring instrument.
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., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Nancy L. Salgado.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York
Date of amendment request: July 31, 2009.
Description of amendment request: The proposed change would revise
the JAFNPP Technical Specifications (TSs) Surveillance Requirements
(SR) for testing of the Residual Heat Removal (RHR) System Shutdown
Cooling (SDC) mode Containment Isolation, Reactor Pressure--High
Function by replacing the current requirement to perform TS SR
3.3.6.1.3, Perform Channel Calibration, with TS SR 3.3.6.1.1 Perform
Channel Check, SR 3.3.6.1.2, Perform Channel Functional Test, SR
3.3.6.1.4, Calibrate the Trip Units, and SR 3.3.6.1.5, Perform Channel
Calibration. These changes are to support a proposed plant modification
to increase the reliability of SDC isolation logic by changing the
source of the reactor high pressure input signal.
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:
[[Page 51330]]
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the SRs that demonstrate the
operability of the SDC Isolation, Reactor Pressure--High function.
The current surveillance requirements include a 92-day calibration
and a 24-month logic system functional test. These surveillance
requirements are typical for pressure switches installed on
dedicated process measurement lines. The proposed change in
surveillance requirements is consistent with the use of ATTS [Analog
Transmitter Trip System] transmitters installed on shared process
measurement lines. The proposed surveillance requirements include
the standard requirements applied to all ATTS equipment and thus
will result in acceptable demonstration of the operability of the
SDC Isolation Reactor Pressure--High function.
The ATTS equipment that will be used for the SDC Isolation,
Reactor Pressure--High function is classified as safety related and
is environmentally qualified. The logic input configuration of the
ATTS equipment will be the same as the configuration of the pressure
switches. This will assure the same functionality currently
performed by the pressure switches currently used for the SDC
Isolation Reactor Pressure--High function. The reliability of the
ATTS has been proven in other RPS [Reactor Protection System], PCIS
[Primary Containment Isolation System], and ECCS [Emergency Core
Cooling System] functions and is comparable to the reliability of
the pressure switches that currently perform the SDC Isolation,
Reactor Pressure--High function. Therefore, the consequences of any
accident mitigated by the SDC Isolation, Reactor Pressure--High
function will not increase.
Based on these considerations, the proposed surveillance
requirement changes do not involve a significant increase in the
probability or consequences of an accident 'previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: No.
The proposed change aligns the TS surveillance requirements with
the type of equipment that will be used to supply the reactor
pressure input to the SDC Isolation Reactor Pressure--High logic.
Since the transmitters that will be used to supply the reactor
pressure input are currently installed equipment there are no new
accidents introduced by the equipment. The proposed change in SRs
aligns the requirements with the--requirements currently imposed on
the equipment in other JAF TS applications. The performance of the
SDC Isolation, Reactor Pressure--High function, is not altered by
changing the input source for reactor pressure parameter. Redundant
power sources within the ATTS assure the functionality of the system
during all plant operating modes that require the SDC Isolation,
Reactor Pressure--High function. The proposed change will not
introduce any new failure modes and, therefore, does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
The TS surveillance requirements that will be imposed on the SDC
Isolation, Reactor Pressure--High function reflect the equipment
that will perform that function. The proposed change in surveillance
requirements will appropriately demonstrate the operability of the
SDC Isolation, Reactor Pressure--High function.
Since the proposed changes to the SRs are consistent with the
SRs for ATTS transmitters in other RPS, PCIS, and ECCS applications
the proposed requirements have been demonstrated to provide an
adequate margin of safety. Therefore, the proposed change does not
involve a significant reduction in any 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: Nancy L. Salgado.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: August 5, 2009.
Description of amendment request: Current Technical Specification
(TS) 5.5.8, ``Inservice Testing Program,'' contains references to the
American Society of Mechanical Engineers (ASME) Boiler and Pressure
Vessel Code, Section XI as the source of requirements for the inservice
testing (IST) of ASME Code Class 1, 2, and 3 pumps and valves. The
proposed amendment would delete the references to Section XI of the
Code and incorporate references to the ASME Code for Operation and
Maintenance of Nuclear Power Plants (ASME OM Code). The proposed
amendment would also indicate that there may be some nonstandard
frequencies utilized in the IST Program in which the provisions of
Surveillance Requirement 3.0.2 are applicable. The proposed changes are
consistent with Technical Specification Task Force (TSTF) Technical
Change Travelers 479-A, ``Changes to Reflect Revision to 10 CFR
50.55a,'' and 497-A, ``Limit Inservice Testing Program SR 3.0.2
Application to Frequencies of 2 Years or Less.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS 5.5.8, Inservice Testing Program,
for consistency with the requirements of 10 CFR 50.55a(f)(4) for
pumps and valves which are classified as American Society of
Mechanical Engineers (ASME) Code Class 1, Class 2 and Class 3. The
proposed change incorporates revisions to the ASME Code which is
consistent with the expectations of 10 CFR 50.55(a).
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. The proposed change does not involve the addition or removal
of any equipment, or any design changes to the facility. Therefore,
this proposed change does not represent a significant increase in
the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed change does not introduce a new accident
initiator, accident precursor, or malfunction mechanism. Therefore,
this proposed change does not create the possibility of an accident
or a different kind than previously evaluated.
3. 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 revises TS 5.5.8, Inservice Testing Program,
for consistency with the requirements of 10 CFR 50.55a(f)(4) for
pumps and valves which are classified as ASME Code Class 1, Class 2
and Class 3. The proposed change incorporates revisions to the ASME
Code, which is consistent with the expectations of 10 CFR 50.55a.
The safety function of the affected pumps and valves are maintained.
Therefore, this 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
[[Page 51331]]
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
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 amendment request: July 30, 2009.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) Section 3.6.3.1, ``Containment
Atmosphere Dilution (CAD) System,'' to modify containment combustible
gas control requirements as permitted by Title 10 of the Code of
Federal Regulations, Part 50 Section 50.44 (10 CFR 50.44). 10 CFR 50.44
was revised on September 16, 2003, as noticed in the Federal Register
(68 FR 54123).
The Nuclear Regulatory Commission (NRC) staff issued a ``Notice Of
Opportunity To Comment On Model Safety Evaluation, Model No Significant
Hazards Determination, And Model Application For Licensees that Wish To
Adopt TSTF-478, Revision 2, `BWR [Boiling-Water Reactor] Technical
Specification Changes that Implement the Revised Rule for Combustible
Gas Control,'' in the Federal Register on October 11, 2007 (72 FR
57970). The notice included a model safety evaluation (SE) and a model
no significant hazards consideration (NSHC) determination. On November
21, 2007, the NRC staff issued a notice in the Federal Register (72 FR
65610) announcing that the model SE and model NSHC determination may be
referenced in plant-specific applications to adopt the changes. In its
application dated July 30, 2009, the licensee affirmed the
applicability of the model NSHC determination which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC is presented below:
Criterion 1: The proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated
The Containment Atmosphere Dilution (CAD) system is not an
initiator to any accident previously evaluated. The TS Required
Actions taken when a drywell cooling system fan is inoperable are
not initiators to any accident previously evaluated. As a result,
the probability of any accident previously evaluated is not
significantly increased.
The revised 10 CFR 50.44 no longer defines a design-basis
accident (DBA) hydrogen release and the Commission has subsequently
found that the DBA loss-of-coolant accident (LOCA) hydrogen release
is not risk significant. In addition, CAD has been determined to be
ineffective at mitigating hydrogen releases from the more risk
significant beyond DBAs that could threaten containment integrity.
Therefore, elimination of the CAD system will not significantly
increase the consequences of any accident previously evaluated. The
consequences of an accident while relying on the revised TS Required
Actions for drywell cooling system fans are no different than the
consequences of the same accidents under the current Required
Actions. As a result, the consequences of any accident previously
evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2: The proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated
No new or different accidents result from utilizing the proposed
change. The proposed change permits physical alteration of the plant
involving removal of the CAD system. The CAD system is not an
accident precursor, nor does its existence or elimination have any
adverse impact on the pre-accident state of the reactor core or post
accident confinement of radionuclides within the containment
building from any design basis event. The changes to the TS do not
alter assumptions made in the safety analysis, but reflect changes
to the design requirements allowed under the revised 10 CFR 50.44.
The proposed change is consistent with the revised safety analysis
assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
Criterion 3: The proposed change does not involve a significant
reduction in a margin of safety
The Commission has determined that the DBA LOCA hydrogen release
is not risk significant, therefore is not required to be analyzed in
a facility accident analysis. The proposed change reflects this new
position and, due to remaining plant equipment, instrumentation,
procedures, and programs that provide effective mitigation of and
recovery from reactor accidents, including postulated beyond design
basis events, does not result in a significant reduction in a margin
of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, the NRC concludes that the proposed change
presents no significant hazards consideration under the standards set
forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. J. Bradley Fewell, Associate General
Counsel, Exelon Generation Company LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station
(FCS), Unit No. 1, Washington County, Nebraska
Date of amendment request: May 29, 2009.
Description of amendment request: The proposed amendment would: (1)
Revise the definition for Operable-Operability in the FCS Technical
Specifications (TS); (2) modify the provisions under which equipment
may be considered operable when either its normal or emergency power
source is inoperable; (3) delete TS limiting condition for operation
(LCO) 2.0.1(2); (4) delete diesel generator surveillance requirement
(SR) 3.7(1)e; and (5) relocate the guidance for inoperable power
supplies and verifying operability of redundant components into the LCO
for electrical equipment 2.7, Electrical Systems.
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 revise the definition of operable-
operability, modify the provisions under which equipment may be
considered operable when either its normal or emergency power source
is inoperable, delete Technical Specification (TS) limiting
conditions for operation (LCO) 2.0.1(2), and relocate the guidance
for inoperable power supplies and verifying operability of redundant
components into the LCO for electrical equipment is more aligned
with NUREG-1432, Standard Technical Specifications [STS] for
Combustion Engineering Plants, and does not adversely impact the
probability of an accident
[[Page 51332]]
previously evaluated. The proposed changes are being made to address
inconsistencies in guidance provided in TS 2.0.1(2) and TS 2.7(2).
The proposed change does not affect the operability requirements for
the emergency diesel generators (EDGs) or the house service
transformers, and therefore does not impact the consequences of an
analyzed accident.
The new requirement added to TS 2.7 provides assurance that a
loss of offsite power during the period that an EDG (or house
service transformer) is inoperable, or loss of an EDG during the
period that a house service transformer is inoperable, or loss of a
house service transformer during the period that an EDG is
inoperable, does not result in a complete loss of safety function of
critical systems; thereby such a loss does not significantly
increase the probability of an accident.
Consistent with NUREG 1432, the 4-hour allowed time added to TS
2.7(2)j for the EDGs, takes into account the capacity and capability
of the remaining alternating current (AC) sources, a reasonable time
for repairs, and the low probability of a design basis accident
(DBA) occurring during this period. On a component basis, single
failure protection for the required feature's function may have been
lost; however, function has not been lost.
Additionally, consistent with NUREG-1432, the 24-hour allowed
time added to TS 2.7(2)b for the house service transformers takes
into account the capacity and capability of the remaining AC
sources, a reasonable time for repairs, and the low probability of a
DBA occurring during this period.
The proposed change removes the surveillance requirement (SR) to
perform an inspection of the EDG on a refueling inspection frequency
in accordance with the manufacturer's recommendations. This
inspection is considered a maintenance activity, not an SR, and has
no impact on the probability of an accident since EDGs are not
initiators for any analyzed event. Deletion of TS SR 3.7(1)e from
the TS does not impact the capability of the EDGs to perform their
accident mitigation functions. The required EDG maintenance
inspections will continue to be performed in accordance with the
licensee-controlled EDG maintenance process. The consequences of an
accident are not impacted because EDG operability is controlled by
other portions of TS 3.7, which ensures that required surveillances
are performed. The appropriate LCOs are entered in the event that
EDG surveillance criteria are not met.
As a result of redefining ``OPERABLE'' and adding the provision
to TS 2.7(2)j, the statements ``provided there are no inoperable
required engineered safeguards components which are redundant''
related to the electrical distribution components are being deleted
from the other 2.7(2) TS for the buses, transformer, and motor
control center (MCC) for clarification and consistency because these
statements restrict only to engineered safeguards components. In
addition, the administrative changes to renumber the existing TS
sections ``TS 2.0.1(3) to 2.0.1(2)'' and TS 3.7(1)f to TS 3.7(1)e.
are being made as a result of deletions to previous TS paragraphs
and are being made for consistency and clarification. Rearranging
the listing order of the MCCs in TS 2.7(1)f and TS 2.7(2)g in bus
order clarifies the TS. As such, these editorial changes are not
initiators of any accidents previously evaluated. As a result, the
probability of an accident previously evaluated is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not involve a physical alteration to the
plant (i.e., no new or different type of equipment will be
installed) or a change in methods governing normal plant operation.
The proposed changes to TS 2.0.1(2) and TS 2.7 do not create the
possibility of a new or different kind of accident since the design
function of the affected equipment is not changed. No new
interactions between systems or components are created. No new
failure mechanisms of associated systems will exist.
By deleting TS LCO 2.0.1(2) and including the guidance in TS
2.7, inconsistencies in the existing TS will be eliminated. The new
requirements added to TS 2.7 will include guidance to declare
required systems or components without a normal or emergency power
source available inoperable, when a redundant system or component is
also inoperable. This provides assurance that a loss of offsite
power, during the period that an EDG (or house service transformer)
is inoperable, or loss of an EDG during the period that a house
service transformer is inoperable (or vice versa), does not result
in a complete loss of safety function of critical systems.
No new failure mechanisms would be created. The proposed changes
do not alter any assumptions made in the safety analyses. For the
most part, the proposed changes are more aligned with the STS.
Therefore, 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 proposed changes to delete TS 2.0.1(2) and relocate the
guidance for inoperable power supplies and verifying operability of
redundant components to TS LCO 2.7(2)j, to delete the statement that
MCC-3C1 may be inoperable in excess of 8 hours if battery chargers
No. 1 and No. 2 are operable, and to delete the SR for inspecting
the DG on a refueling frequency in accordance with the
manufacturer's recommendations do not alter the manner in which
safety limits or limiting safety system settings are determined. The
safety analysis acceptance criteria are not affected by these
proposed changes. The sources of power credited for design basis
events are not affected by the proposed changes.
The proposed changes to modify the provisions under which
equipment may be considered operable when either its normal or
emergency power source is inoperable, delete TS LCO 2.0.1(2), and
relocate the guidance for inoperable power supplies and verifying
operability of redundant components into the LCO for electrical
equipment is more aligned with the STS. These changes are being made
to address inconsistencies in guidance provided in TS 2.0.1(2) and
TS 2.7(2). The proposed change does not reduce the operability
requirements for the transformers, buses, MCCs, or EDGs and
therefore will not result in plant operation in a configuration
outside of the design basis.
Further, the proposed change does not change the design function
of any equipment assumed to operate in the event of an accident.
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: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES Units 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: March 24, 2009, as supplemented by
letters dated April 24, and September 11, 2009.
Description of amendment request: The proposed change revises the
allowable value in the Technical Specification (TS) Table 3.3.5.1-1
(Function 3.d) for the high-pressure coolant injection (HPCI) automatic
pump suction transfer from the condensate storage tank (CST) to the
suppression pool (SP). The present allowable value for this transfer is
greater than or equal to 36 inches above the CST bottom. The proposed
change is to increase the allowable value for this transfer to occur at
greater than or equal to 40.5 inches above the CST bottom.
Additionally, the proposed amendment also includes an editorial/
administrative change which corrects a typographical error in the SSES
Units 1 and 2 TS Section 3.10.8.f.
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:
[[Page 51333]]
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change to TS Table 3.3.5.1-1 increases the
Technical Specification allowable value for the HPCI suction low
level automatic transfer function from >= 3 6 inches to >= 40.5
inches above the CST bottom. There are no process setpoint changes
associated with this TS allowable value change. This TS change does
not introduce the possibility of an increase in the probability or
consequences of an accident because the HPCI automatic transfer
function is not an initiator of any new accidents nor does it
introduce any new failure modes. The CST is not safety related and
therefore not credited in any design basis accident analyses.
However, the CST reserve volume is credited in anticipated
transients without scram (ATWS), Appendix R and station blackout
(SBO) evaluations. The reserve volume available in the CST at the
proposed allowable value of 40.5 inches above the CST bottom remains
adequate to fully support these HPCI system support functions and
the change fully supports HPCI system operation. The reserve volume
is not reduced as a result of the proposed change in the TS
allowable value since the transfer will still occur at the CST low
level instrument setpoint of 43.5 inches above tank bottom, which
remains unchanged.
The HPCI system automatic transfer function occurs at the point
in a design basis accident (DBA) when the CST level reaches the low
level transfer setpoint. This proposed change will require the HPCI
pump suction to be transferred from the CST to the SP at 40.5 inches
versus 36 inches above the CST bottom. Currently, the TS allow this
transfer to occur at 36 inches. This proposed change is conservative
because it assures the suction transfer will occur while there is
more water in the tank, thus eliminating the possibility of vortex
formation and air intrusion to the HPCI pump suction. Since this
proposed change ensures the HPCI system automatic suction transfer
function occurs without adversely impacting HPCI system operation,
it does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed editorial/administrative change is necessary to
correct a typographical error in the SSES Units 1 and 2 TS Section
3.10.8.f. This editorial 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?
No. As discussed above, the proposed change to TS Table 3.3.5.1-
1 involves increasing the TS allowable value for the HPCI low level
automatic transfer function from the CST to the SP at >= 36 inches
to >= 40.5 inches above the CST tank bottom. This change ensures the
HPCI automatic transfer function occurs without introducing the
possibility of vortex formation or air intrusion in the HPCI pump
suction path. All HPCI system support functions remain unaffected by
this change. This TS change does not introduce the possibility of a
new accident because the HPCI automatic transfer function is not an
initiator of any accident and no new failure modes are introduced.
There are no new types of failures or new or different kinds of
accidents or transients that could be created by these changes.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed editorial/administrative change only corrects a
typographical error in the SSES Units 1 and 2 TS Section 3.10.8.f.
This editorial 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?
No. The margin of safety is established through equipment
design, operating parameters, and the setpoints at which automatic
actions are initiated. The proposed change to TS Table 3.3.5.1-1
involves increasing the allowable level at which the HPCI automatic
suction transfer from the CST to the SP must occur to avoid the
possibility of vortex formation or air intrusion into the HPCI pump.
This change does not result in a change to the level switch
setpoint, which initiates the HPCI suction transfer from the CST to
the SP. Although the allowable value for the transfer is now closer
to the process setpoint for activation of the level switch, this
reduction in operating margin was reviewed and determined to be
acceptable. The level switch setpoint tolerances were established
based on historical instrument data and instrument characteristics.
These tolerances provide adequate margin to the proposed TS
allowable value of 40.5 inches above the CST bottom. The tolerances
further ensure the transfer will occur prior to level reaching the
technical specification allowable value. Therefore, the proposed
change does not result in a significant reduction in a margin of
safety.
The proposed editorial/administrative change only corrects a
typographical error in the SSES Units 1 and 2 TS Section 3.10.8.f.
This editorial change does not result in 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: Nancy L. Salgado.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: July 30, 2009.
Description of amendment request: The proposed amendment would
relocate Technical Specification (TS) surveillance requirements (SRs)
for the reactor recirculation system motor-generator (MG) set scoop
tube stop settings to the Technical Requirements Manual (TRM).
Specifically, the proposed amendment would relocate TS SR 4.4.1.1.3 to
the TRM which is a licensee-controlled document. SR 4.4.1.1.3 requires
that each MG set scoop tube mechanical and electrical stop be
demonstrated operable with overspeed setpoints less than or equal to
109% and 107%, respectively, of rated core flow, at least once per 18
months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The Nuclear Regulatory Commission (NRC) staff's review
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 major components in the MG set consist of a motor, fluid
coupler and a generator. The motor drives the generator through the
fluid coupler. The speed and output of the generator rise and fall
as the volume of fluid in the coupler is varied by changing the
position of the scoop tube. As the generator's output increases or
decreases, the speed of the recirculation pump follows suit. The
scoop tube mechanism has both mechanical and electrical overspeed
stops that limit recirculation flow by limiting the MG set speed.
The electrical stop actuates first. The mechanical stop is designed
to prevent the scoop tube motion if the electrical stop fails or to
mitigate overshoot of the electrical stop. The electrical stops are
not credited in any of the accident or transient analyses. The
mechanical stop settings are an input used in the determination of
the flow dependent minimum critical power ratio (MCPR) and the
linear heat generation rate (LHGR) or average planar linear heat
generation rate (APLHGR) operating limits. These operating limits
are established and documented on a cycle-specific basis in the core
operating limits report (COLR) in accordance with TS 6.9.1.9.
Operation within the MCPR, LGHR and APLHGR operating limits is
required in accordance with TSs 3.2.3, 3.2.4, and 3.2.1,
respectively.
Once relocated, any future changes to the surveillance
requirements for the MG set scoop tube mechanical and electrical
stop settings would be controlled by 10 CFR 50.59.
There are no physical plant modifications associated with this
change. The proposed amendment would not alter the way any
structure, system, or component (SSC) functions and would not alter
the way the plant is operated. As such, the proposed
[[Page 51334]]
amendment would have no impact on the ability of the affected SSCs
to either preclude or mitigate an accident. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment would not change the design function or
operation of the SSCs involved and would not impact the way the
plant is operated. As such, the proposed change would not introduce
any new failure mechanisms, malfunctions, or accident initiators not
already considered in the design and licensing bases. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety is associated with the confidence in the
ability of the fission product barriers (i.e., fuel cladding,
reactor coolant pressure boundary, and containment structure) to
limit the level of radiation to the public. There are no physical
plant modifications associated with the proposed amendment. The
proposed amendment would not alter the way any SSC functions and
would not alter the way the plant is operated. The proposed
amendment would not introduce any new uncertainties or change any
existing uncertainties associated with any safety limit. The
proposed amendment would have no impact on the structural integrity
of the fuel cladding, reactor coolant pressure boundary, or
containment structure. Based on the above considerations, the NRC
staff concludes that the proposed amendment would not degrade the
confidence in the ability of the fission product barriers to limit
the level of radiation to the public. Therefore, the proposed change
does not involve a significant reduction in a margin of safety.
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: Vincent Zabielski, PSEG Nuclear LLC-N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
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 System (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: May 21, 2009.
Brief description of amendments: The amendments removed the Table
of Contents from the Technical Specifications and place them under
licensee control.
Date of issuance: September 21, 2009.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment Nos.: 293 and 269.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: June 30, 2009 (74 FR
31320).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated September 21, 2009.
No significant hazards consideration comments received: No.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina.
Duke Power Company 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 for amendments: February 27, 2009.
Brief description of amendments: The amendments deleted those
portions of the Technical Specifications (TSs) superseded by the Code
of Federal Regulations, Part 26, Subpart I. The changes are consistent
with Nuclear Regulatory Commission (NRC)-approved Revision 0 to
Technical Specification Task Force (TSTF) Improved Standard Technical
Specification Change Traveler, TSTF-511, ``Eliminate Working Hour
Restrictions from TS 5.2.2 to Support Compliance with 10 CFR Part 26.''
Date of issuance: September 21, 2009.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 251 and 246.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the licenses and technical specifications.
Amendment Nos.: 253 and 233.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and technical specifications.
Amendment Nos.: 365, 367, and 366.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the licenses and technical specifications.
Date of initial notices in Federal Register: August 11, 2009 (74 FR
40236) Catawba and McGuire; and August 11, 2009 (74 FR 40237) Oconee.
The Commission's related evaluation and final finding of no
significant hazards consideration of the
[[Page 51335]]
amendments is contained in a Safety Evaluation dated September 21,
2009.
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: November 13, 2008, as
supplemented by letters dated June 1, July 14, and August 17, 2009.
Brief description of amendment: The amendment modified Technical
Specification 3.3.1.1, Reactor Protective Instrumentation, specifically
Table 4.3-1 and associated Notes 7 and 8, to clarify and streamline
Reactor Coolant System flow verification requirements associated with
the Departure from Nucleate Boiling Ratio reactor trip signal.
Date of issuance: September 16, 2009.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 286.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications/license.
Date of initial notice in Federal Register: January 27, 2009 (74 FR
4769). The supplemental letters dated June 1, July 14, and August 17,
2009, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 16, 2009.
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: May 13, 2009, as supplemented by
letter dated July 8, 2009.
Brief description of amendment: The amendment modified Technical
Specification 2.1.1.1, departure from nucleate boiling ratio safety
limit based upon the Combustion Engineering 16 $x 16 Next Generation
Fuel design and the associated departure from nucleate boiling
correlations.
Date of issuance: September 18, 2009.
Effective date: As of the date of issuance and shall be implemented
after the current cycle (Cycle 20) is completed and prior to startup
for operating Cycle 21.
Amendment No.: 287.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications/license.
Date of initial notice in Federal Register: June 30, 2009 (74 FR
31321). The supplemental letter dated July 8, 2009, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 18, 2009.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: July 25, 2008 (Agencywide
Documents Access and Management System (ADAMS) Accession No.
ML082110187), as supplemented by letters dated October 31, 2008 (ADAMS
Accession No. ML083080059), February 17, 2009 (ADAMS Accession No.
ML090480372), May 8, 2009 (ADAMS Accession No. ML092380433) and July
27, 2009 (ADAMS Accession No. ML092100162).
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.3.1.1, ``Reactor Protection System (RPS)
Instrumentation,'' Surveillance Requirement (SR) 3.3.1.1.8 and TS
3.3.1.3, ``Oscillation Power Range Monitor (OPRM) Instrumentation,'' SR
3.3.1.3.2 to increase the frequency interval between Local Power Range
Monitor calibrations from 1000 effective full power hours (EFPH) to
2000 EFPH.
Date of issuance: September 16, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 195/182.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications and License.
Date of initial notice in Federal Register: January 23, 2009 (74 FR
4250-4251). The October 31, 2008, February 17, 2009, May 8, 2009, and
July 27, 2009 supplements, contained clarifying information and did not
change the NRC staff's initial proposed finding of no significant
hazards consideration nor expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 16, 2009.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: October 9, 2008, supplemented by
letter dated April 2, 2009.
Brief description of amendment: The amendment reflects the planned
installation of replacement steam generators (SGs). Specifically, the
amendment modified the technical specifications to eliminate the
existing requirements associated with tube sleeve repairs and alternate
repair criteria which are not applicable to the replacement SGs. It
also incorporated a revised primary-to-secondary leakage criterion,
changes the required reporting period for SG inspection results, and
incorporated revised tube integrity surveillance frequency requirements
to reflect the new Alloy 690 tubing material.
Date of issuance: September 15, 2009.
Effective date: Upon installation of the replacement SGs and shall
be implemented prior to exiting cold shutdown from the TMI-1 SG
replacement refueling outage (T1R18), which is scheduled to begin in
the fall of 2009.
Amendment No.: 271.
Facility Operating License No. DPR-50: Amendment revised the
license and the technical specifications.
Date of initial notice in Federal Register: March 10, 2009 (74 FR
10310).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 15, 2009.
No significant hazards consideration comments received: No.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: January 30, 2009, as
supplemented by a letter dated July 30, 2009.
Brief description of amendment: The amendment deleted the Duane
Arnold Energy Center (DAEC) Technical Specification (TS) Section
5.2.2.e regarding work hour controls.
Date of issuance: September 18, 2009.
Effective date: As of the date of issuance and shall be implemented
by October 1, 2009. Amendment No.: 274.
[[Page 51336]]
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 24, 2009 (74 FR
12393). The supplemental letter contained clarifying information, did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 18, 2009.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: October 13, 2008, as supplemented by
letters dated April 8, May 29, June 12, and September 1, 2009.
Brief description of amendment: The amendment revised the licensing
basis by approving adoption of the Alternative Source Term (AST), in
accordance with Section 50.67 of Title 10 of the Code of Federal
Regulations (10 CFR), for use in calculating the loss-of-coolant
accident (LOCA) dose consequences. The amendment revised the Technical
Specifications (TSs) to (1) change the TS definition for DOSE
EQUIVALENT I-131 to adopt Federal Guidance Report 11 dose conversion
factors; (2) require operability of the Standby Liquid Control system
in Mode 3, to reflect its credit in the LOCA analysis; (3) establish a
Main Steam (MS) Pathway leakage limit that effectively increases the
previous MS isolation valve leakage limit; and (4) change TS Section
5.5.12 to reflect a requested permanent exemption from the requirements
of 10 CFR Part 50, Appendix J, Option B, Section III.A, to allow
exclusion of MS Pathway leakage from the overall integrated leakage
rate measured during the performance of a Type A test, and from the
requirements of Appendix J, Option B, Section III.B, to allow exclusion
of the MS Pathway leakage from the combined leakage rate of the
penetrations and valves subject to Type B and C tests.
Date of issuance: September 15, 2009.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment No.: 234.
Facility Operating License No. DPR-46: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 23, 2009 (74 FR
4251). The supplemental letters dated April 8, May 29, June 12, and
September 1, 2009, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 15, 2009.
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: March 30, 2009.
Brief description of amendment request: The amendments revised
Technical Specification (TS) by deleting the Reactor Coolant Pump
breaker position reactor trip in TS 3.3.1, ``Reactor Trip System (RTS)
Instrumentation.''
Date of Issuance: September 18, 2009.
Amendment Nos.: Unit 1-183; Unit 2-176.
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: May 19, 2009 (74 FR
23448).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 18, 2009.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of application for amendment: June 5, 2009.
Brief description of amendment: The amendment revised WBN Unit 1
technical specifications (TSs) to revise the completion time from 1
hour to 24 hours for Condition B of TS 3.5.1, ``Accumulators'' and its
associated Bases.
Date of issuance: September 9, 2009.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment No.: 81.
Facility Operating License No. NPF-90: Amendment revises TS 3.5.1.
Date of initial notice in Federal Register: June 30, 2009 (74 FR
31326).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 9, 2009.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been
[[Page 51337]]
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 O1F21,
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 email to
[email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, any person(s) whose interest may be
affected by this action may file a request for a hearing and a petition
to intervene with respect to issuance of the amendment to the subject
facility operating license. Requests for a hearing and a petition for
leave to intervene shall be filed in accordance with the Commission's
``Rules of Practice for Domestic Licensing Proceedings'' in 10 CFR Part
2. Interested person(s) should consult a current copy of 10 CFR 2.309,
which is available at the Commission's PDR, located at One White Flint
North, Public File Area O1F21, 11555 Rockville Pike (first floor),
Rockville, Maryland, and electronically on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected].
If a request for a hearing or petition for leave to intervene is filed
by the above date, the Commission or a presiding officer designated by
the Commission or by the Chief Administrative Judge of the Atomic
Safety and Licensing Board Panel, will rule on the request and/or
petition; and the Secretary or the Chief Administrative Judge of the
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------
\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
[[Page 51338]]
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007, (72 FR 49139). The E-Filing process
requires participants to submit and serve all adjudicatory documents
over the internet or in some cases to mail copies on electronic storage
media. Participants may not submit paper copies of their filings unless
they seek a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
[email protected], or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms Viewer\TM\ to
access the Electronic Information Exchange (EIE), a component of the E-
Filing system. The Workplace Forms Viewer\TM\ is free and is available
at http://www.nrc.gov/site-help/e-submittals/install-viewer.html.
Information about applying for a digital ID certificate is available on
NRC's public website 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 using the agency's adjudicatory e-
filing system 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 Meta-System Help Desk, which is
available between 8 a.m. and 8 p.m., Eastern Time, Monday through
Friday, excluding government holidays. The Meta-System Help Desk can be
contacted by telephone at 1-866-672-7640 or by e-mail at
[email protected].
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, in accordance
with 10 CFR 2.302(g), with their initial paper filing requesting
authorization to continue to submit documents in paper format. Such
filings must be submitted by: (1) First class mail addressed to the
Office of the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff; or (2) courier, express mail, or expedited
delivery service to the Office of the Secretary, Sixteenth Floor, One
White Flint North, 11555 Rockville, Pike, Rockville, Maryland, 20852,
Attention: Rulemaking and Adjudications Staff. Participants filing a
document in this manner are responsible for serving the document on all
other participants. Filing is considered complete by first-class mail
as of the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the petition
and/or request should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii).
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings, unless an NRC regulation or
other law requires submission of such information. With respect to
copyrighted works, except for limited excerpts that serve the purpose
of the adjudicatory filings and would constitute a Fair Use
application, participants are requested not to include copyrighted
materials in their submission.
Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California
Date of application for amendment: September 3, 2009, as
supplemented on September 8, 2009.
Brief description of amendment: The amendment revised the Diablo
Canyon Power Plant, Unit No. 2 Technical Specification (TS) 3.7.1,
``Main Steam Safety Valves (MSSVs),'' by increasing the Power Range
Neutron Flux High setpoint in TS Table 3.7.1-1 from 87 percent rated
thermal power (RTP) to 106 percent RTP. This will allow the unit to
operate at full power with one main steam safety valve, MS-2-RV-224,
inoperable for the remainder of Cycle 15.
Date of issuance: September 17, 2009.
Effective date: As of its date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 208.
Facility Operating License No. DPR-82: The amendment revised the
Facility Operating License and Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. A public notice of the proposed amendment
was published in The Tribune newspaper, located in San Luis Obispo,
California,
[[Page 51339]]
on September 11 and 12, 2009. The notice provided an opportunity to
submit comments on the NRC staff's proposed NSHC determination.
The supplemental letter dated September 8, 2009, 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 Tribune.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consideration of public comments, state
consultation, and final NSHC determination are contained in a safety
evaluation dated September 17, 2009.
Attorney for licensee: Jennifer Post, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: Michael T. Markley.
Dated at Rockville, Maryland, this 25th day of September 2009.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E9-23780 Filed 10-5-09; 8:45 am]
BILLING CODE 7590-01-P