[Federal Register Volume 74, Number 124 (Tuesday, June 30, 2009)]
[Notices]
[Pages 31318-31331]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-15117]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0261]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from June 4, 2009 to June 17, 2009. The last
biweekly notice was published on June 16, 2009 (74 FR 28575).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking
and Directives Branch, 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. Copies of written comments
received may be examined at the Commission's Public Document Room
(PDR), located at One White Flint North, Public File Area O1F21, 11555
Rockville Pike (first floor), Rockville, Maryland.
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 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
[[Page 31319]]
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
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 ViewerTM
to access the Electronic Information Exchange (EIE), a component of the
E-Filing system. The Workplace Forms ViewerTM is free and is
available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying for a digital ID certificate is
available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC electronic filing
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time,
Monday through Friday, excluding government holidays. The electronic
filing 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
[[Page 31320]]
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.
For further details with respect to this amendment action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: May 21, 2009.
Description of amendments request: The amendments would remove the
Table of Contents (TOC) from the Technical Specifications (TSs) and
place them under licensee control.
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?
No.
The proposed change is administrative and affects control of a
document, the TOC, listing the specifications in the plant TSs.
Transferring control from the Nuclear Regulatory Commission (NRC) to
CCNPP [Calvert Cliffs Nuclear Power Plant] (the licensee) does not
affect the operation, physical configuration, or function of plant
equipment or systems. It does not impact the initiators or
assumptions of analyzed events; nor does it impact the mitigation of
accidents or transient events. The change has no impact on, and
hence cannot increase, 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?
No.
The proposed change is administrative and does not alter the
plant configuration, require installation or new equipment, alter
assumptions about previously analyzed accidents, or impact the
operation or function of plant equipment or systems. Therefore, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No.
The proposed change is administrative. The TOC is not required
by regulation to be in the TS. Removal does not impact any safety
assumptions or have the potential to reduce a margin of safety as
described in the TS Bases. The change involves a transfer of control
of the TOC from the NRC to CCNPP. No change in the technical content
of the TS specifications is involved. Consequently, transfer from
the NRC to CCNPP has no impact on the margin of safety, and hence
cannot involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
NRC Acting Branch Chief: John Boska.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: May 5, 2009.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) Section 6.7.C to change
requirements related to the schedule for performing the 10 CFR Part 50,
Appendix J, Type A test. Specifically, the proposed change would change
the TS from requiring the test ``no later than April 2010'' to ``prior
to startup from the April 2010 refuel outage.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1.0 Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. The change does not impact the function of any
structure, system or component that affects the probability of an
accident or that supports mitigation or consequences of an accident
previously evaluated. The proposed change involves testing of
Primary Containment but does not impact containment design or
performance requirements. The proposed change ensures that the Type
A test is performed prior to establishing Primary Containment
following the April 2010 Refuel[ing] Outage. The proposed change
does not affect reactor operations or accident analysis and there is
no change to the radiological consequences of a previously analyzed
accident. The operability requirements for accident mitigation
systems remain consistent with the licensing and design basis.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2.0 Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The proposed change does not involve any physical
alteration of plant equipment and does not change the method by
which any safety-related system performs its function. The proposed
change involves the scheduling of the Type A test and does not alter
the way the test is performed. Type A tests have been previously
performed and are well within the design capability of station
structures, systems or components. No new or different types of
equipment will be permanently installed or operated. Operation of
existing installed equipment is unchanged. The methods governing
plant operation and testing remain consistent with current safety
analysis assumptions. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3.0 Does the proposed change involve a significant reduction in
a margin of safety?
Response: No. These changes do not change any existing design or
operational requirements and do not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analysis. The proposed change affects the
schedule for performing the Type A test and does not affect the way
the test is
[[Page 31321]]
performed or margins for the existing Primary Containment. As such,
there are no changes being made to safety analysis assumptions,
safety limits or safety system settings that would adversely affect
plant safety as a result of the proposed change. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Acting Branch Chief: John Boska.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: May 13, 2009.
Description of amendment request: The proposed change will modify
the Technical Specification (TS) 2.1.1.1, ``DNBR,'' to revise the
Departure from Nucleate Boiling Ratio (DNBR) safety limit based upon
the Combustion Engineering (CE) 16 x 16 Next Generation Fuel (NGF)
design and the associated Departure from Nucleate Boiling (DNB)
correlations.
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.
No changes to plant equipment or operating procedures are
required due to the change in the safety limit for DNBR. This change
does not impact any of the accident initiators. The analyses of the
reload are performed using NRC [U.S. Nuclear Regulatory Commission]
approved methodologies to ensure the Specified Acceptable Fuel
Design Limits (SAFDLs), of which DNBR is one, are not violated. The
current DNBR setpoint continues to ensure automatic protective
action is initiated to prevent exceeding the proposed DNBR safety
limit.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not result in any plant modifications
or change in the way the plant is designed to function. The proposed
change is not associated with any accident precursor or initiator.
The proposed change supports the loading and use of Next Generation
Fuel (NGF) at ANO-2 [Arkansas Nuclear One, Unit 2] as previously
approved by the NRC.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The use the NRC-approved NGF WSSV-T correlation with the ABB-NV
correlation to establish a new bounding DNBR safety limit of 1.23,
preserves the DNBR margin of safety at a 95/95 level. The Core
Protection Calculator (CPC) DNBR power adjustment addressable
constant BERR1 is calculated based on the WSSV-T and ABB-NV CHF
[critical heat flux] correlations such that a CPC trip at a DNBR of
1.25 using the CE-1 CHF correlation assures that the bounding DNBR
safety limit of 1.23 for the WSSV-T and ABB-NV CHF correlations will
not be violated during normal operation and AOOs [anticipated
operational occurrences] to at least a 95/95 probability/confidence
level.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: May 15, 2009.
Description of amendment request: The amendment would modify
Technical Specification (TS) 6.6.5, ``Core Operating Limits Report
(COLR),'' to minimize the number of U.S. Nuclear Regulatory Commission
(NRC)-approved references consistent with the guidance provided in NRC
Generic Letter 88-16, ``Removal of Cycle-Specific Parameter Limits from
Technical Specifications,'' dated October 3, 1988. This request also
fulfills the commitment made in the licensee's letter to the NRC dated
March 11, 2008, ``Response to Request for Additional Information
License Amendment Request to Revise Technical Specification 6.6.5, Core
Operating Limits Report.''
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 to the list of NRC-approved methodologies
listed in TS 6.6.5 are administrative in nature and have no impact
on any plant configuration or system performance relied upon to
mitigate the consequences of an accident. Changes to the calculated
core operating limits may only be made using NRC-approved
methodologies, must be consistent with all applicable safety
analysis limits, and are controlled by the 10 CFR 50.59 [Title 10 of
the Code of Federal Regulations Section 50.59] process.
The proposed change will minimize and clarify the listing of the
NRC-approved methodologies that are currently being used in the ANO-
2 [Arkansas Nuclear One, Unit 2] core designs and the determination
of the operating limits for those cores. Assumptions used for
accident initiators and/or safety analysis acceptance criteria are
not altered by the proposed changes.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to the list of topical reports used to
determine the operating limits has no impact on any plant
configurations or on system performance that is relied upon to
mitigate the consequences of an accident. These changes are
administrative in nature and do not result in a change to the
physical plant or to the modes of operation defined in the facility
license.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not amend the cycle specific parameter
limits located in the COLR from the values presently required by the
TS. The individual specifications continue to require operation of
the plant within the bounds of the limits specified in COLR. The
proposed change to the list of analytical methods referenced in the
COLR is administrative in nature.
[[Page 31322]]
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana.
Date of amendment request: May 22, 2009.
Description of amendment request: The proposed amendment will
modify the Waterford Steam Electric Station, Unit 3 (Waterford 3),
Technical Specification (TS) 6.9.1.11 to minimize the number of
references that reflect U.S. Nuclear Regulatory Commission (NRC)-
approved methods used in establishing the Core Operating Limits Report
(COLR) parameter limits, consistent with the guidance provided in NRC
Generic Letter 88-16, ``Removal of Cycle-Specific Parameter Limits from
Technical Specifications,'' dated October 3, 1988.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the list of NRC-approved methodologies
listed in TS 6.9.1.11.1 are administrative in nature and have no
impact on any plant configuration or system performance relied upon
to mitigate the consequences of an accident. Changes to the
calculated core operating limits may only be made using NRC approved
methodologies, must be consistent with all applicable safety
analysis limits, and are controlled by the 10 CFR 50.59 [Title 10 of
the Code of Federal Regulations Section 50.59] process.
The proposed changes will minimize and clarify the listing of
the NRC-approved methodologies that are currently being used in the
Waterford 3 core designs and the determination of the operating
limits for those cores.
Assumptions used for accident initiators and/or safety analysis
acceptance criteria are not altered by the proposed changes.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the list of topical reports used to
determine the operating limits has no impact on any plant
configurations or on system performance that is relied upon to
mitigate the consequences of an accident. These changes are
administrative in nature and do not result in a change to the
physical plant or to the modes of operation defined in the facility
license.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not amend the cycle specific parameter
limits located in the COLR from the values presently required by the
TS. The individual specifications continue to require operation of
the plant within the bounds of the limits specified in COLR.
The proposed changes to the list of analytical methods
referenced in the COLR are administrative in nature.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois;
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station
(QCPS), Units 1 and 2, Rock Island County, Illinois
Date of application for amendment request: April 7, 2009.
Description of amendment request: The proposed amendment deletes a
no longer applicable footnote from the DNPS Technical Specifications
(TS), corrects administrative errors in the titles of analytical
methods, and deletes historical analytical methods no longer applicable
in DNPS and QCPS TS. The proposed amendment also deletes a license
condition from the DNPS and QCPS Renewed Facility Operating License
(FOL).
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
DNPS TS 3.4.5, ``RCS Leakage Detection Instrumentation,''
establishes the applicability and requirements for equipment used to
quantify unidentified reactor coolant system operational leakage
(i.e., the drywell floor drain sump monitoring system). The proposed
change deletes a footnote that established a limited duration
alternative to these requirements for DNPS Unit 3.
The deletion of the footnote restores DNPS TS 3.4.5 requirements
to the requirements prior to NRC approval of an emergency license
amendment, which provided an alternative means to demonstrate TS
compliance. In that the condition necessitating the footnote (i.e.,
a failed component) has been resolved (i.e., repair of the failed
component), the footnote is no longer applicable. The proposed
change will have no effect on any accident initiator or precursor
previously evaluated and will not change the manner in which the
plant is operated. Thus, the proposed change does not have any
effect on the probability of an accident previously evaluated.
DNPS and QCNPS TS 5.6.5 ``Core Operating Limits Report (COLR),''
lists the NRC-approved analytical methods that are used at DNPS and
QCNPS to determine core operating limits. The proposed changes will
correct administrative errors in the titles of several analytical
methods in DNPS and QCNPS TS 5.6.5.b. The proposed changes will also
delete historical analytical methods from DNPS and QCNPS TS 5.6.5.b
that are no longer applicable, as well as renumber the remaining
analytical methods.
The correction of administrative errors in the titles of
analytical methods does not change the content or application of the
methods. Similarly, the deletion of non-applicable analytical
methods does not affect the ability to accurately model core
behavior, including the determination of core operating limits, for
the fuel that is currently loaded in the DNPS and QCNPS reactors.
Therefore, the proposed changes will have no effect on any accident
initiator or precursor previously evaluated and will not change the
manner in which the core is operated. Thus, the proposed changes do
not have any effect on the probability of an accident previously
evaluated.
Finally, the proposed changes will delete a license condition in
the DNPS Units 2 and 3 and QCNPS Units 1 and 2 Facility
[[Page 31323]]
Operating Licenses (FOLs) that limits the maximum average fuel rod
burnup to 60 gigawattdays per metric ton of uranium (GWD/MTU) until
a generic environmental assessment that supports an extended limit
is approved.
The proposed deletion of the license condition is justified by
completion of generic environmental assessments for DNPS and QCNPS
(i.e., as required by the license condition). As such, these license
conditions are no longer required or applicable. Therefore, the
proposed change will have no effect on any accident initiator or
precursor previously evaluated and will not change the manner in
which the core is operated. Thus, the proposed changes do not have
any effect on the probability of an accident previously evaluated.
The proposed changes to the DNPS TS 3.4.5, DNPS and QCNPS TS
5.6.5.b, and the deletion of the Renewed FOL license conditions do
not affect the ability to successfully respond to previously
evaluated accidents and does not affect the radiological assumptions
used in the evaluations for both DNPS and QCNPS.
Thus, the proposed changes will have no effect on the type or
amount of radiation released, and will have no effect on predicted
offsite doses in the event of an accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to DNPS TS Section 3.4.5, DNPS and QCNPS TS
Section 5.6.5, and the proposed deletion of Renewed FOL license
conditions do not affect the performance of any structure, system,
or component credited with mitigating any accident previously
evaluated.
The deletion of the footnote from DNPS TS 3.4.5 restores TS
requirements to the requirements prior to NRC approval of an August
2008 emergency license amendment. The proposed deletion of the
footnote does not affect the control parameters governing unit
operation or the response of plant equipment to transient
conditions. The proposed changes do not introduce any new modes of
system operation or failure mechanisms.
The NRC-approved analytical methodologies in TS 5.6.5.b are used
to accurately model core behavior, including the determination of
core operating limits, for the fuel that is currently loaded in the
DNPS and QCNPS reactors. These methodologies do not affect the
control parameters governing unit operation or the response of plant
equipment to transient conditions. The proposed changes do not
introduce any new modes of system operation or failure mechanisms.
The existing Renewed FOL license condition limits fuel burnup
until completion of a generic environmental assessment. In June
2004, the NRC issued NUREG-1437, ``Generic Environmental Impact
Statement for License Renewal of Nuclear Plants,'' Supplement 16,
``Quad Cities Nuclear Power Station, Units 1 and 2,'' and Supplement
17, ``Dresden Nuclear Power Station, Units 2 and 3.'' Based on the
completion and conclusions of these generic environmental
assessments for DNPS and QCNPS, the license condition limiting fuel
burnup for each unit has been satisfied. As such, these license
conditions are no longer required or applicable.
The proposed deletion of the license condition does not affect
the control parameters governing unit operation or the response of
plant equipment to transient conditions. The proposed changes do not
introduce any new modes of system operation or failure mechanisms.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
The proposed changes to DNPS TS 3.4.5, DNPS and QCNPS TS
5.6.5.b, and the DNPS and QCNPS Renewed FOLs (i.e., deletion of the
fuel burnup license condition) will not affect the ability to
quantify unidentified RCS leakage, accurately model core behavior
for the currently loaded fuel, and ensure compliance with NRC-
approved LTRs.
As such, the proposed changes do not modify the safety limits or
setpoints at which protective actions are initiated and do not
change the requirements governing operation or availability of
safety equipment assumed to operate to preserve the margin of
safety. Therefore, the proposed changes provide an equivalent level
of protection as that currently provided.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Russell A. Gibbs.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: April 13, 2009.
Description of amendment request: The amendment would delete those
portions of Technical Specifications superseded by 10 CFR Part 26,
Subpart I. This change is consistent with 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.''
The NRC staff issued a notice of availability of the model safety
evaluation and model no significant hazards consideration (NSHC), using
the consolidated line-item improvement process for referencing in
license amendment applications in the Federal Register on December 30,
2008 (73 FR 79923). The licensee affirmed the applicability of the
following NSHC determination in its application dated April 13, 2009.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1: The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26.
Removal of the Technical Specification requirements will be
performed concurrently with the implementation of the 10 CFR Part
26, Subpart I, requirements. The proposed change does not impact the
physical configuration or function of plant structures, systems, or
components (SSCs) or the manner in which SSCs are operated,
maintained, modified, tested, or inspected. Worker fatigue is not an
initiator of any accident previously evaluated. Worker fatigue is
not an assumption in the consequence mitigation of any accident
previously evaluated.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2: The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. Working hours will
continue to be controlled in accordance with NRC requirements. The
new rule allows for deviations from controls to mitigate or prevent
a condition adverse to safety or as necessary to maintain the
security of the facility. This ensures that the new rule will not
unnecessarily restrict working hours and thereby create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change does not alter the plant configuration,
require new plant equipment to be installed, alter accident analysis
assumptions, add any initiators, or
[[Page 31324]]
effect the function of plant systems or the manner in which systems
are operated, maintained, modified, tested, or inspected. 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 proposed change removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. The proposed change
does not involve any physical changes to the plant or alter the
manner in which plant systems are operated, maintained, modified,
tested, or inspected. The proposed change does not alter the manner
in which safety limits, limiting safety system settings or limiting
conditions for operation are determined. The safety analysis
acceptance criteria are not affected by this change. The proposed
change will not result in plant operation in a configuration outside
the design basis. The proposed change does not adversely affect
systems that respond to safely shut down the plant and to maintain
the plant in a safe shutdown condition. Removal of plant-specific
Technical Specification administrative requirements will not reduce
a margin of safety because the requirements in 10 CFR Part 26 are
adequate to ensure that worker fatigue is managed.
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Thomas H. Boyce.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center (DAEC), Linn County, Iowa
Date of amendment requests: March 4, 2009.
Description of amendment requests: The proposed amendments would
change the Technical Specification (TS) Section 5.5.12 (Primary
Containment Leakage Rate Testing Program) and change TS Section 3.6.1.3
(Primary Containment Isolation Valves) to remove the repair criterion
for Main Steamline Isolation Valves (MSIVs) that fail their as-found
leakage rate acceptance criterion found in current Surveillance
Requirement 3.6.1.3.9.
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.
This proposed change to TS 5.5.12 does not modify existing
structures, systems or components (SSCs) of the plant, and it does
not introduce new SSCs. It does not change assumptions, methodology,
likelihood, or results of previously evaluated accidents in the
Updated Final Safety Analysis Report [UFSAR]. It does not change
operating procedures or administrative controls that affect the
functions of SSCs. By excluding Main Steam pathway leakage from Type
A, and Type B and C test results, this change will make the Primary
Containment Leakage Rate Testing Program more closely aligned with
the assumptions used in associated accident dose consequence
analyses.
The proposed change [to TS 3.6.1.3] to eliminate the repair
criterion (i.e., as-left leakage limit) for MSIVs that fail their
as-found leak test, does not change how the MSIVs function in
response to any event, nor the likelihood of occurrence of any
accident previously identified in the UFSAR. Repairing the MSIVs to
an as-left leakage value, which can be higher than the currently
specified value in TS that reliably assures the next as-found
leakage test will be within limits is sufficient to ensure that the
calculated dose consequences of any event involving MSIV leakage as
an effluent pathway remain within analyzed limits.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
changes. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. The changes do not alter assumptions made in the safety
analysis for MSIV performance.
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.
Since Main Steam pathway leakage bypasses the containment and
its filtration system (Standby Gas Treatment System) during a Loss-
of-Coolant Accident (LOCA), the effect on release to the environment
is analyzed and specifically accounted for in the DAEC dose analysis
methodology approved by Amendments 237 and 241. This proposed change
to exclude Main Steam pathway leakage from Type A, and Type B and C
test results does not change dose analysis values, and thus does not
affect actual margin in the dose analysis.
Similarly, removing the as-left repair criterion for MSIVs from
the TS has no impact on the assumptions for MSIV leakage used in the
accident analysis, which are based upon the as-found MSIV leakage
limit, not the as-left leakage. As long as the as-left leakage value
gives high confidence that the as-found leakage will remain within
limits over the next operating cycle until the next as-found leak
test is conducted, the assumptions of the dose consequence analyses
are not adversely impacted and the previously calculated results
remain bounding.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. R. E. Helfrich, Florida Power & Light
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Lois M. James.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: April 17, 2009.
Description of amendment request: The proposed amendment would
revise Operating License No. DPR-49 by changing ``FPL Energy Duane
Arnold, LLC'' to ``NextEra Energy Duane Arnold, LLC,'' where
appropriate, to reflect the renaming of FPL Energy Duane Arnold, LLC to
NextEra Energy Duane Arnold, LLC.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This request is for administrative changes only. No actual
facility equipment or accident analyses will be affected by the
proposed changes. Therefore, this request will have no impact on 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.
[[Page 31325]]
This request is for administrative changes only. No actual
facility equipment or accident analyses will be affected by the
proposed changes and no failure modes not bounded by previously
evaluated accidents will be created. Therefore, this request will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, Reactor Coolant
System pressure boundary, and containment structure) to limit the
level of radiation dose to the public. This request is for
administrative changes only. No actual plant equipment or accident
analyses will be affected by the proposed changes. Additionally, the
proposed changes will not relax any criteria used to establish
safety limits, will not relax any safety system settings, and will
not relax the bases for any limiting conditions of operation.
Therefore, these proposed changes will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. R. E. Helfrich, Florida Power & Light
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Lois M. James.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 2, 2009.
Description of amendment request: The proposed amendment would (1)
delete Technical Specification (TS) surveillance requirement (SR)
3.1.3.2 and revise SR 3.1.3.3, (2) remove reference to SR 3.1.3.2 from
Required Action A.3 of TS 3.1.3, ``Control Rod OPERABILITY,'' and (3)
revise Example 1.4-3 in TS Section 1.4, ``Frequency,'' to clarify the
applicability of the 1.25 surveillance test interval extension. The
changes are in accordance with U.S. Nuclear Regulatory Commission
(NRC)-approved TS Task Force (TSTF) traveler TSTF-475, Revision 1,
``Control Rod Notch Testing Frequency and SRM [Source Range Monitor]
Insert Control Rod Action.''
The NRC issued a ``Notice of Availability of Model Application
Concerning Technical Specification Improvement To Revise Control Rod
Notch Surveillance Frequency, Clarify SRM Insert Control Rod Action,
and Clarify Frequency Example'' in the Federal Register on November 13,
2007 (72 FR 63935). In its application dated June 2, 2009, the licensee
affirmed the applicability of the model no significant hazards
consideration (NSHC).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC adopted by the licensee is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change generically implements TSTF-475, Revision 1,
``Control Rod Notch Testing Frequency and SRM Insert Control Rod
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and
NUREG-1434 (BWR/6) STS. The changes: (1) Revise TS testing frequency
for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod
OPERABILITY'', (2) clarify the requirement to fully insert all
insertable control rods for the limiting condition for operation
(LCO) in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring
Instrumentation'' (NUREG-1434 only), and (3) revise Example 1.4-3 in
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25
surveillance test interval extension. The consequences of an
accident after adopting TSTF-475, Revision 1 are no different than
the consequences of an accident prior to adoption. Therefore, this
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
proposed change will not introduce new failure modes or effects and
will not, in the absence of other unrelated failures, lead to an
accident whose consequences exceed the consequences of accidents
previously analyzed. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
TSTF-475, Revision 1 will: (1) [Revise the TS SR 3.1.3.2
frequency in TS 3.1.3, ``Control Rod OPERABILITY'', (2) clarify the
requirement to fully insert all insertable control rods for the
limiting condition for operation (LCO) in TS 3.3.1.2, ``Source Range
Monitoring Instrumentation,'' and (3)] revise Example 1.4-3 in
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25
surveillance test interval extension. [The GE Nuclear Energy Report,
``CRD Notching Surveillance Testing for Limerick Generating
Station,'' dated November 2006, concludes that extending the control
rod notch test interval from weekly to monthly is not expected to
impact the reliability of the scram system and that the analysis
supports the decision to change the surveillance frequency.]
Therefore, the proposed changes in TSTF-475, Revision 1 are
acceptable and do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based upon this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendment involves NSHC.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-321, 50-
366, 50-348, 50-364, 50-424, 50-425, Joseph M. Farley Nuclear Plant,
Unit Nos. 1 and 2 (FNP), Houston County, Alabama, Edwin I. Hatch
Nuclear Plant, Unit Nos. 1 and 2 (HNP), Appling County, Georgia, Vogtle
Electric Generating Plant, Units Nos. 1 and 2 (VEGP), Burke County,
Georgia
Date of amendment request: May 19, 2009.
Description of amendment request: The proposed amendment would
delete those portions of technical specifications (TS) superseded by
Title 10 of the Code of Federal Regulations (10 CFR) Part 26, Subpart
I. This change is consistent with the Nuclear Regulatory Commission
(NRC)-approved Revision 0 to Technical Specification Task Force (TSTF)
Traveler, TSTF-511, ``Eliminate Working Hour Restrictions from TS 5.2.2
to Support Compliance with 10 CFR Part 26.'' The availability of this
TS improvement was announced in the Federal Register on December 30,
2008, (73 FR 79923) as part of the consolidated line item improvement
process.
Basis for proposed no significant hazards consideration
determination: SNC has reviewed the no significant hazards
determination published on December 30, 2008 (73 FR 79925), as part of
the CLIIP Notice of Availability. SNC has concluded that the
determination presented in the notice is applicable to FNP, HNP, and
VEGP. SNC has evaluated the proposed changes to the TS using the
criteria in 10 CFR 50.92 and has determined that the proposed changes
do not involve a significant hazards consideration. An analysis of the
issue of no significant hazards consideration is presented below:
[[Page 31326]]
Criterion 1: The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. Removal of the
Technical Specification requirements will be performed concurrently
with the implementation of the 10 CFR Part 26, Subpart I,
requirements. The proposed change does not impact the physical
configuration or function of plant structures, systems, or
components (SSCs) or the manner in which SSCs are operated,
maintained, modified, tested, or inspected. Worker fatigue is not an
initiator of any accident previously evaluated. Worker fatigue is
not an assumption in the consequence mitigation of any accident
previously evaluated.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2: The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. Working hours will
continue to be controlled in accordance with NRC requirements. The
new rule allows for deviations from controls to mitigate or prevent
a condition adverse to safety or as necessary to maintain the
security of the facility. This ensures that the new rule will not
unnecessarily restrict working hours and thereby create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change does not alter the plant configuration,
require new plant equipment to be installed, alter accident analysis
assumptions, add any initiators, or affect the function of plant
systems or the manner in which systems are operated, maintained,
modified, tested, or inspected.
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 proposed change removes Technical Specification restrictions
on working hours for personnel who perform safety related functions.
The Technical Specification restrictions are superseded by the
worker fatigue requirements in 10 CFR Part 26. The proposed change
does not involve any physical changes to the plant or alter the
manner in which plant systems are operated, maintained, modified,
tested, or inspected. The proposed change does not alter the manner
in which safety limits, limiting safety system settings or limiting
conditions for operation are determined. The safety analysis
acceptance criteria are not affected by this change. The proposed
change will not result in plant operation in a configuration outside
the design basis. The proposed change does not adversely affect
systems that respond to safely shutdown the plant and to maintain
the plant in a safe shutdown condition.
Removal of plant-specific Technical Specification administrative
requirements will not reduce a margin of safety because the
requirements in 10 CFR Part 26 are adequate to ensure that worker
fatigue is managed.
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: FNP: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201, HNP: Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037, VEGP: Mr.
Arthur H. Domby, Troutman Sanders, NationsBank Plaza, Suite 5200, 600
Peachtree Street, NE., Atlanta, Georgia 30308-2216.
NRC Branch Chief: Melanie C. Wong.
Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: June 5, 2009.
Description of amendment request: The proposed amendment would
correct an error by changing a logic connector from ``OR'' to ``AND''
between Technical Specification (TS) 3.3.2, ``ESFAS [Engineered Safety
Feature Actuation System] Instrumentation,'' Condition I, Actions I.2.1
and I.2.2.
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.
This proposed amendment corrects an identified error by only
changing a logic connector between two TS actions. The change only
restores the sequential nature of these required actions consistent
with other similar TS actions where, if conditions warrant, the
movement of the plant to lower modes is required (i.e., to Mode 3,
to Mode 4, etc.). In addition, this change does not alter the
completion times for these actions. 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.
By correcting the logic connector between these two TS actions,
this change only restores consistency with other similar TS actions
where movement of the plant to lower modes is required. The change
does not alter the expected outcome of the required actions nor does
it change the completion times for these actions. Therefore, the
possibility of a new or different kind of accident from those
previously analyzed has not been created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
By only correcting the logic connector between the required
actions, the proposed change does not alter the expected outcome of
the required actions nor does it change the completion times for
these actions. 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: June 5, 2009.
Description of amendment request: The proposed amendment would
change the technical specifications to revise the completion time (CT)
from 1 hour to 24 hours for Condition B of TS 3.5.1, ``Accumulators,''
and its associated Bases. Condition B of TS 3.5.1 currently specifies a
CT of one hour to restore a reactor coolant system (RCS) accumulator to
operable status when declared inoperable due to any reason except not
being within the required boron concentration range.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration, adopted by the licensee is
presented below:
[[Page 31327]]
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The basis for the accumulator limiting condition for operation
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a
sufficient volume of borated water will be immediately forced into
the core through each of the cold legs in the event the RCS pressure
falls below the pressure of the accumulators, thereby providing the
initial cooling mechanism during large RCS pipe ruptures. As
described in Section 9.2 of the WCAP-15049, ``Risk-Informed
Evaluation of an Extension to Accumulator Completion Times,''
evaluation, the proposed change will allow plant operation in a
configuration outside the design basis for up to 24 hours, instead
of 1 hour, before being required to begin shutdown. The impact of
the increase in the accumulator CT on core damage frequency for all
the cases evaluated in WCAP-15049 is within the acceptance limit of
1.0E-06/yr for a total plant core damage frequency (CDF) less than
1.0E-03/yr. The incremental conditional core damage probabilities
calculated in WCAP-15049 for the accumulator CT increase meet the
criterion of 5E-07 in Regulatory Guides (RG) 1.174 and 1.177 for all
cases except those that are based on design basis success criteria.
As indicated in WCAP-15049, design basis accumulator success
criteria are not considered necessary to mitigate large break loss-
of-coolant accident (LOCA) events, and were only included in the
WCAP-15049 evaluation as a worst case data point. In addition, WCAP-
15049 states that the NRC has indicated that an incremental
conditional core damage frequency (ICCDP) greater than 5E-07 does
not necessarily mean the change is unacceptable.
The proposed technical specification change does not involve any
hardware changes nor does it affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, engineered safety feature (ESF) actuation setpoints,
accident mitigation capabilities, accident analysis assumptions or
inputs.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. As described in Section 9.1 of the WCAP-
15049 evaluation, the plant design will not be changed with this
proposed technical specification CT increase. All safety systems
still function in the same manner and there is no additional
reliance on additional systems or procedures. The proposed
accumulator CT increase has a very small impact on core damage
frequency. The WCAP-15049 evaluation demonstrates that the small
increase in risk due to increasing the accumulator allowed outage
time (AOT) is within the acceptance criteria provided in RGs 1.174
and 1.177. No new accidents or transients can be introduced with the
requested change and the likelihood of an accident or transient is
not impacted.
The malfunction of safety related equipment, assumed to be
operable in the accident analyses, would not be caused as a result
of the proposed technical specification change. No new failure mode
has been created and no new equipment performance burdens are
imposed.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not involve a significant reduction in
a margin of safety. There will be no change to the departure from
nucleate boiling ratio (DNBR) correlation limit, the design DNBR
limits, or the safety analysis DNBR limits.
The basis for the accumulator LCO, as discussed in Bases Section
3.5.1, is to ensure that a sufficient volume of borated water will
be immediately forced into the core through each of the cold legs in
the event the RCS pressure falls below the pressure of the
accumulators, thereby providing the initial cooling mechanism during
large RCS pipe ruptures. As described in Section 9.2 of the WCAP-
15049 evaluation, the proposed change will allow plant operation in
a configuration outside the design basis for up to 24 hours, instead
of 1 hour, before being required to begin shutdown. The impact of
this on plant risk was evaluated and found to be very small. That
is, increasing the time the accumulators will be unavailable to
respond to a large LOCA event, assuming accumulators are needed to
mitigate the design basis event, has a very small impact on plant
risk. Since the frequency of a design basis large LOCA (a large LOCA
with loss of offsite power) would be significantly lower than the
large LOCA frequency of the WCAP-15049 evaluation, the impact of
increasing the accumulator CT from 1 hour to 24 hours on plant risk
due to a design basis large LOCA would be significantly less than
the plant risk increase presented in the WCAP-15049 evaluation.
Therefore, this 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
Tennessee Valley Authority, Docket No. 50 390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: June 5, 2009.
Description of amendment request: The proposed amendment would
provide alternatives for valve position verification in various
Required Actions and Surveillance Requirements in Technical
Specification 3.6.3, ``Containment Isolation Valves.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment will revise the position verification
requirements for manual containment isolation devices that are
locked, sealed, or otherwise secured in the closed position.
Revising the verification requirements will not introduce any
physical changes or result in the equipment being operated in a new
or different manner. All systems, structures, and components
previously required for mitigation of a transient remain capable of
performing their designed functions. Furthermore, although the
proposed change would revise the position verification requirements,
no physical change is being made to the assumed position of the
valves for accident analysis. Therefore, the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
No new accident scenarios or failure mechanisms are introduced
as a result of this proposed change. The proposed amendment would
revise the position verification requirements but not alter any
valve positions. With no changes to the plant lineup, no new or
different accidents are possible. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Changes to the position verification requirements of normally
closed manual containment isolation valves that are locked, sealed,
or otherwise secured do not change the position/status of these
valves. The proposed amendment does not impact the ability of these
valves to perform their design function of controlling containment
leakage rates during design basis radiological accidents. Therefore,
the proposed change
[[Page 31328]]
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: February 12, 2008, as
supplemented by letters dated August 28, 2008, September 15, 2008,
October 17, 2008, December 15, 2008, December 18, 2008 (two letters),
April 9, 2009, and May 20, 2009.
Brief description of amendment: The amendment revised the Technical
Specification (TS) Sections 2.1, ``Limiting Safety System Setting,''
3.1, ``Reactor Protection System,'' 3.2, ``Protective Instrument
Systems,'' associated Surveillance Requirements, and other TS with
similar requirements as these instrumentation TS sections.
Date of Issuance: June 12, 2009.
Effective date: As of the date of issuance, and shall be
implemented within 180 days.
Amendment No.: 236.
Facility Operating License No. DPR-28: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: April 22, 2008 (73 FR
21659).
The supplemental letters dated August 28, 2008, September 15, 2008,
October 17, 2008, December 15, 2008, December 18, 2008 (two letters),
April 9, 2009, and May 20, 2009, the application as originally noticed,
and did not change the staff's original proposed no significant hazards
consideration determination. The Commission's related evaluation of
this amendment is contained in a Safety Evaluation dated June 12, 2009.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: September 22, 2008.
Brief description of amendment: The proposed amendment would revise
the Technical Specification (TS) to remove the requirement to perform
quarterly closure time testing of the Main Steam Isolation Valves
(MSIVs) by deleting TS Surveillance Requirement 4.7.D.1.c. Operability
testing of the MSIVs will continue to be required by the Vermont Yankee
Inservice Test Program and the safety functions of the MSIVs will
continue to be contained in the Vermont Yankee Updated Final Safety
Analysis Report and Vermont Yankee Technical Requirements Manual.
Date of Issuance: June 17, 2009.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 237.
Facility Operating License No. DPR-28: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: November 4, 2008 (73 FR
65692).
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: March 2, 2009.
Brief description of amendment: The amendment deletes those
portions of Technical Specifications superseded by 10 CFR Part 26,
Subpart I. This change is consistent with NRC-approved Revision 0 to
Technical Specification Task Force Traveler, TSTF-511, ``Eliminate
Working Hour Restrictions from TS 5.2.2 to Support Compliance with 10
CFR Part 26,'' as announced in the Federal Register on December 30,
2008 (73 FR 79923) as part of the consolidated line item improvement
process.
Date of issuance: June 9, 2009.
Effective date: As of the date of issuance and shall be implemented
by October 1, 2009.
Amendment No.: 181.
Renewed Facility Operating License No. NPF-12: Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: March 24, 2009 (74 FR
12395). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 9, 2009.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: October 23, 2008.
Brief description of amendments: The amendments revised the
Sequoyah
[[Page 31329]]
Nuclear Plant (SQN), Units 1 and 2 Technical Specifications (TSs) by a
partial adoption of Technical Specifications Task Force (TSTF)
Traveler, TSTF-491, Revision 2, ``Removal of Main Steam and Feedwater
Valve Isolation Times.'' The amendments only revised TS 3.7.1.5, ``Main
Steam Line Isolation Valves,'' by relocating the main steam isolation
valve closure time from Surveillance Requirement 4.7.1.5.1 to the
Bases. The amendments deviated from TSTF-491 in that the current SQN TS
3.7.1.6 ``Main Feedwater Isolation, Regulating, and Bypass Valves,''
and associated surveillance requirements do not include the main
feedwater valve closure times, and thus, TSTF-491 changes to TS 3.7.1.6
were not applied to the SQN TSs.
Date of issuance: June 12, 2009.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 324 and 316.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: January 13, 2009 (74 FR
1716). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 12, 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 issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to
[email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, 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 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland, and electronically on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected].
If a request for a hearing or petition for leave to intervene is filed
by the above date, the Commission or a presiding officer designated by
the Commission or by the Chief Administrative Judge of the Atomic
Safety and Licensing Board Panel, will rule on the request and/or
petition; and the Secretary or the Chief Administrative Judge of the
Atomic Safety and Licensing Board will issue a
[[Page 31330]]
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
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 ViewerTM
to access the Electronic Information Exchange (EIE), a component of the
E-Filing system.
The Workplace Forms ViewerTM is free and is available at
http://www.nrc.gov/site-help/e-submittals/install-viewer.html.
Information about applying for a digital ID certificate is available on
NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC electronic filing
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time,
Monday through Friday,
[[Page 31331]]
excluding government holidays. The electronic filing 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.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: June 4, 2009, as supplemented by
letter dated June 6, 2009.
Brief description of amendment: The amendment authorizes a
temporary one-time change to Technical Specification (TS) 3.8.1
Required Action B.4 Completion Time. The amendment would add a note
allowing a Completion Time of ``17 days'', on a temporary one-time
basis. This one-time allowance will expire at 10:15 a.m. on June 12,
2009.
Date of issuance: June 8, 2009.
Effective date: As of the date of issuance, and shall be
implemented immediately.
Amendment No.: 294.
Facility Operating License No. DPR-59: The amendment revised the
License and the Technical Specifications.
Public comments requested as to the proposed no significant hazards
consideration (NSHC): No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, State consultation, and final NSHC
determination are contained in a safety evaluation dated June 8, 2009.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Acting Branch Chief: John P. Boska.
Dated at Rockville, Maryland, this 19th day of June 2009.
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E9-15117 Filed 6-29-09; 8:45 am]
BILLING CODE 7590-01-P