[Federal Register Volume 72, Number 242 (Tuesday, December 18, 2007)]
[Notices]
[Pages 71703-71719]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-24284]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from November 22, 2007, to December 5, 2007. The
last biweekly notice was published on December 4, 2007 (72 FR 68206).
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. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice,
person(s) may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
via electronic submission through the NRC E-Filing system for a hearing
and a petition for leave to intervene. Requests for a hearing and a
petition for leave to intervene shall be filed in accordance with the
Commission's ``Rules of Practice for Domestic Licensing Proceedings''
in 10 CFR Part 2. Interested person(s) should consult a current copy of
10 CFR 2.309, which is available at the Commission's PDR, located at
One White Flint North, Public File Area O1F21, 11555 Rockville Pike
(first floor), Rockville, Maryland. Publicly available records will be
accessible from the Agencywide Documents Access and Management System's
(ADAMS) Public Electronic Reading Room on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request
for a hearing or petition for leave to intervene is filed within 60
days, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or
[[Page 71704]]
fact to be raised or controverted. In addition, the petitioner/
requestor shall provide a brief explanation of the bases for the
contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for hearing or a petition for leave to intervene must be
filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007, (72 FR 49139). The E-Filing process
requires participants to submit and serve documents over the internet
or in some cases to mail copies on electronic storage media.
Participants may not submit paper copies of their filings unless they
seek a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
[email protected], or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms ViewerTM
to access the Electronic Information Exchange (EIE), a component of the
E-Filing system. The Workplace Forms 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 technical help line,
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time,
Monday through Friday. The help line number is (800) 397-4209 or
locally, (301) 415-4737.
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, in accordance
with 10 CFR 2.302(g), with their initial paper filing requesting
authorization to continue to submit documents in paper format. Such
filings must be submitted by: (1) First-class mail addressed to the
Office of the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff; or (2) courier, express mail, or expedited
delivery service to the Office of the Secretary, Sixteenth Floor, One
White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852,
Attention: Rulemaking and Adjudications Staff. Participants filing a
document in this manner are responsible for serving the document on all
other participants. Filing is considered complete by first-class mail
as of the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the petition
and/or request should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later
than 11:59 p.m. Eastern Time on the due date.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings. With respect to copyrighted
works, except for limited excerpts that serve the purpose of the
adjudicatory filings and would constitute a Fair Use application,
participants are requested not to include copyrighted materials in
their submission.
[[Page 71705]]
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].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit No.1 (CPS), DeWitt County, Illinois
Date of amendment request: September 27, 2007.
Description of amendment request: The proposed amendment would
modify technical specification (TS) by relocating references to
specific American Society for Testing and Materials (ASTM) standards
for fuel oil testing to licensee-controlled documents. In the
referenced letter, AmerGen (the licensee) previously received approval
for a change to the Unit No. 1, CPS TS that added the water and
sediment content test as alternative criteria to the ``clear and
bright'' acceptance test for new fuel oil.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The proposed changes relocate the specific ASTM standard
references from the Administrative Controls Section of TS to a
licensee-controlled document. Requirements to perform testing in
accordance with applicable ASTM standards are retained in the TS as
are requirements to perform surveillances of both new and stored
diesel fuel oil. Future changes to the licensee-controlled document
will be evaluated pursuant to the requirements of 10 CFR 50.59,
``Changes, tests and experiments,'' to ensure that such changes do
not result in more than a minimal increase in the probability or
consequences of an accident previously evaluated. In addition, the
``clear and bright'' test used to establish the acceptability of new
fuel oil for use prior to addition to storage tanks has been
expanded to recognize more rigorous testing of water and sediment
content. Relocating the specific ASTM standard references from the
TS to a licensee-controlled document and allowing a water and
sediment content test to be performed to establish the acceptability
of new fuel oil will not affect nor degrade the ability of the
emergency diesel generators (DGs) to perform their specified safety
function. Fuel oil quality will continue to meet ASTM requirements.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not adversely
affect the ability of structures, systems, and components (SSCs) to
perform their intended safety function to mitigate the consequences
of an initiating event within the assumed acceptance limits.
The proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of any accident previously evaluated.
Further, the proposed changes do not increase the types and amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures.
Therefore, the changes do not involve a significant increase in
the probability or consequences of any 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 relocate the specific ASTM standard
references from the Administrative Controls Section of TS to a
licensee-controlled document. In addition, the ``clear and bright''
test used to establish the acceptability of new fuel oil for use
prior to addition to storage tanks has been expanded to allow a
water and sediment content test to be performed to establish the
acceptability of new fuel oil. 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 requirements retained in the TS continue to require
testing of the diesel fuel oil to ensure the proper functioning of
the DGs. Therefore, the changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes relocate the specific ASTM standard
references from the Administrative Controls Section of TS to a
licensee-controlled document. Instituting the proposed changes will
continue to ensure the use of applicable ASTM standards to evaluate
the quality of both new and stored fuel oil designated for use in
the emergency DGs. Changes to the licensee-controlled document are
performed in accordance with the provisions of 10 CFR 50.59. This
approach provides an effective level of regulatory control and
ensures that diesel fuel oil testing is conducted such that there is
no significant reduction in a margin of safety.
The ``clear and bright'' test used to establish the
acceptability of new fuel oil for use prior to addition to storage
tanks has been expanded to allow a water and sediment content test
to be performed to establish the acceptability of new fuel oil. The
margin of safety provided by the DGs is unaffected by the proposed
changes since there continue to be TS requirements to ensure fuel
oil is of the appropriate quality for emergency DG use. The proposed
changes provide the flexibility needed to improve fuel oil sampling
and analysis methodologies while maintaining sufficient controls to
preserve the current margins of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell A. Gibbs.
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: November 8, 2007.
Description of amendments request: The amendment would clarify the
Technical Specification definitions for Channel Calibration and Channel
Functional Test. The proposed amendments would incorporate Technical
Specification Task Force (TSTF) Standard Technical Specification Change
Traveler, TSTF-205-A, ``Revision of Channel Calibration, Channel
Functional Test, and Related Definitions,'' Revision 3, dated July 31,
2003.
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. Would not involve a significant increase in the probability
or consequences of any accident previously evaluated.
The proposed change clarifies the Technical Specification
requirements for performance of channel calibrations and channel
functional tests. Specifically, the proposed change incorporates the
Nuclear Regulatory Commission-approved Technical Specification Task
Force Standard Technical Specification Change Traveler, TSTF-205-A,
``Revision of Channel Calibration, Channel Functional Test, and
Related Definitions,'' Revision 3, dated July 31, 2003. The change
does not adversely affect the performance or effectiveness of
required testing, as testing appropriate to the associated
Surveillance Requirements will continue to be performed. The
proposed change does not have a detrimental impact on the condition
or performance of any plant structure, system, or component that
could initiate an analyzed
[[Page 71706]]
event. Therefore, the probability of an accident previously
evaluated is not significantly increased.
The equipment being calibrated or tested is still required to be
operable and capable of performing the accident mitigation functions
assumed in the accident analysis. As a result, the consequences of
any accident previously evaluated are not significantly affected.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The scope of the proposed change is limited to the clarification
of existing calibration and test requirements. As such, 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.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. The proposed change will not involve a significant reduction
in [a] margin of safety.
The margin of safety in this case is the verification of
instrument channel operability. The proposed change clarifies
requirements for the performance of channel calibrations and channel
functional tests. Specifically, the proposed change incorporates the
Nuclear Regulatory Commission-approved Technical Specification Task
Force Standard Technical Specification Change Traveler, TSTF-205-A,
``Revision of Channel Calibration, Channel Functional Test, and
Related Definitions,'' Revision 3, dated July 31, 2003. No changes
of setpoints to plant process limits are involved. The surveillance
requirements, as revised, will continue to ensure that affected
equipment is tested in a manner that gives confidence that the
equipment can perform its appropriate safety function.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
NRC Branch Chief: Mark G. Kowal.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: November 9, 2007.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 3.8.a.7 related to the movement of
heavy loads over and in the spent fuel pools and would relocate the
modified requirements to a licensee-controlled document, the Kewaunee
Power Station Technical Requirements Manual (TRM). The proposed
amendment is needed to facilitate future spent fuel cask handling
activities associated with dry cask spent fuel storage. The proposed
amendment would incorporate the use of a single-failure-proof lifting
system for handling of necessary heavy loads over or in the spent fuel
pool with irradiated fuel in either the fuel storage racks or in the
just-loaded spent fuel canister in the spent fuel pool. The proposed
modified TS 3.8.a.7 would then be relocated to the TRM.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment revises Kewaunee Power Station (KPS)
heavy load handling Technical Specification (TS) 3.8.a.7
requirements consistent with modifications to the Auxiliary Building
(AB) crane and the NRC's [Nuclear Regulatory Commission] current
guidance for single-failure-proof lifting systems. The proposed
amendment also relocates the affected heavy load handling-related TS
to a licensee-controlled document, consistent with the NRC's
regulations.
The proposed change to TS 3.8.a.7 permits spent fuel cask
handling in the spent fuel pool, which is required for loading spent
fuel for dry storage at the on-site Independent Spent Fuel Storage
Installation (ISFSI). Proposed TS 3.8.a.7 includes a new requirement
that the AB crane and associated lifting devices meet the applicable
single-failure-proof criteria.
Heavy load handling will continue to be conducted in accordance
with the KPS heavy load handling program, which meets the NRC's
guidance in NUREG-0612, as described in this LAR, and as augmented
by Regulatory Information Summary 2005-25. With the upgrade of the
AB crane load handling system, drops of heavy loads will not be
considered credible. Notwithstanding the AB crane upgrade, heavy
loads will still be prohibited from being suspended over irradiated
fuel in the spent fuel pool storage racks under the revised
requirements.
The previously evaluated cask drop accident is not considered
credible with the upgraded AB crane because the crane trolley is
being upgraded to a single-failure-proof design, consistent with
applicable NRC-endorsed guidance. Lifting devices and interfacing
lifting points associated with spent fuel cask handling will also be
designed in accordance with applicable NRC guidance pertaining to
single-failure-proof lifting systems. The result of these design
upgrades is that the AB crane will retain the lifted load in the
event of a single failure in the load path, including a failure of a
wire rope. In addition, the crane will hold the load and the trolley
and bridge will be designed to stay on their respective rails during
a design basis seismic event.
The relocation of TS 3.8.a.7 to the KPS Technical Requirements
Manual (TRM) is an administrative change that does not affect plant
operation or heavy load handling.
Revised TS 3.8.a.7 and its associated Bases will be relocated to
the TRM after approval of this amendment request. Changes to the KPS
TRM are controlled by 10 CFR 50.59. Regulation 10 CFR 50.59 requires
that NRC approval be obtained prior to any change that would result
in more than a minimal increase in (1) the frequency of occurrence
of an accident previously evaluated, (2) likelihood of occurrence of
a malfunction of a SSC important to safety previously evaluated, or
(3) consequences of a malfunction of a SSC important to safety
previously evaluated. Accordingly, upon relocation of the
requirements of TS 3.8.a.7 and associated Bases to the TRM,
appropriate control of changes will be maintained, based on the
criteria in 10 CFR 50.59. Administrative relocation of the
requirements of TS 3.8.a.7 does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, configuration of KPS or the manner in which it is
operated.
Therefore, the proposed change does not significantly 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?
Response: No.
Heavy load handling will continue to be conducted in accordance
with the KPS heavy load handling program, which meets the NRC's
guidance in NUREG-0612, as approved for KPS. Drops of heavy loads
will continue to be very improbable events and the upgrade of the
KPS AB crane lifting system to a single-failure-proof design
provides additional defense-in-depth against such events.
Notwithstanding the AB crane upgrade, heavy loads will still be
prohibited from being suspended over irradiated fuel in the spent
fuel pool storage racks under the revised requirements.
Heavy load handling operations at KPS will continue to be
conducted as they currently are and no new heavy load handling
operations are required as a result of this amendment. The
previously evaluated cask drop accident is not considered credible
with the upgraded AB crane because the crane trolley is being
upgraded to a single-failure-proof design, consistent with
applicable NRC-endorsed guidance. Lifting devices and interfacing
lifting points associated with spent fuel cask handling will also be
designed in accordance with applicable NRC guidance pertaining to
single-failure-proof lifting systems. The result
[[Page 71707]]
of these design upgrades is that the AB crane will retain the lifted
load in the event of a single failure in the load path, including a
failure of a wire rope. In addition, the crane will hold the load
and the trolley and bridge will be designed to stay on their
respective rails during a design basis seismic event.
The relocation of TS 3.8.a.7 to the KPS Technical Requirements
Manual (TRM) is an administrative change that does not affect plant
operation or heavy load handling.
Accordingly, upon relocation of the requirements of TS 3.8.a.7
and associated Bases to the TRM, appropriate control of changes will
be maintained, based on the criteria in 10 CFR 50.59. Modification
of the requirements of TS 3.8.a.7 does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, configuration of KPS or the manner in which it is
operated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment revises KPS heavy load handling TS
3.8.a.7 requirements consistent with modifications to the AB crane
and the NRC's current guidance for single-failure-proof lifting
systems.
Heavy load handling will continue to be conducted in accordance
with the KPS heavy load handling program, which meets the NRC's
guidance in NUREG-0612, as approved for KPS. Drops of heavy loads
will continue to be very improbable events and the upgrade of the
KPS AB crane lifting system to a single-failure-proof design
provides additional defense-in-depth against such events and an
increase in overall design margin. Notwithstanding the AB crane
upgrade, heavy loads will still be prohibited from being suspended
over irradiated fuel in the spent fuel pool storage racks under the
revised requirements.
Further, the relocation of TS 3.8.a.7 to the KPS Technical
Requirements Manual (TRM) is an administrative change that does not
affect plant operation or heavy load handling.
Heavy load handling operations at KPS will continue to be
conducted as they currently are and no new heavy load handling
operations are required as a result of this amendment. The
previously evaluated cask drop accident is less probable with the
upgraded AB crane because the crane trolley is being upgraded to a
single-failure-proof design, consistent with applicable NRC-endorsed
guidance. Lifting devices and interfacing lifting points associated
with spent fuel cask handling will also be designed in accordance
with applicable NRC guidance pertaining to single-failure-proof
lifting systems. The result of these design upgrades is that the AB
crane will retain the lifted load in the event of a single failure
in the load path, including a failure of a wire rope. In addition,
the crane will hold the load and the trolley and bridge will be
designed to stay on their respective rails during a design basis
seismic event.
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: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc.,
120 Tredegar Street, Richmond, VA 23219.
NRC Acting Branch Chief: Cliff Munson.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: November 9, 2007.
Description of amendment request: The proposed amendment would
revise the Kewaunee Power Station (KPS) Updated Safety Analysis Report
(USAR) to modify the design and licensing basis for the auxiliary
building (AB) crane. The proposed amendment would allow the use of a
methodology for performing the seismic qualification analysis of the
upgraded crane. The crane is being upgraded to become a single-failure-
proof design. The new methodology includes rolling of the crane bridge
and trolley wheels during a seismic event.
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 amendment request pertains solely to an analysis method
supporting the upgrade of the KPS AB crane from a non-single-
failure-proof design to a single-failure-proof design. The AB crane
is used to lift and handle loads in the KPS spent fuel pool and
truck bay areas. The AB crane does not interface with operating
plant equipment. The design rated load of the AB crane remains the
same as previously approved. The proposed amendment does not change
the current heavy load handling practices that are in use at KPS.
Upgrading the AB crane to a single-failure-proof design will reduce
the probability of a heavy load drop in the areas where the AB crane
lifts and handles loads.
The seismic analysis method proposed for use recognizes the
inherent propensity for structures not fixed to one another (e.g.,
steel wheels on steel rails) to roll if sufficient lateral force is
applied to either object. This seismic analysis method is proposed
for use solely on the AB crane upgrade and not for any other plant
structures, systems, or components. The recognition of wheel rolling
between the AB crane trolley and bridge and their respective rails
reflects the true nature of the installed equipment and its response
to horizontal forces generated by a seismic event. Consideration of
rolling reduces the projected analyzed loads on the crane and
building structures and eliminates the need for unnecessary
modifications to both.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This amendment request pertains to an analysis method supporting
the upgrade of an existing plant component. Specifically, the
existing AB crane trolley is being replaced with a state-of-the-art
design that is single-failure-proof. The AB crane does not interface
with operating plant equipment. This seismic analysis method is
proposed for use solely on the AB crane upgrade and not for any
other plant structures, systems, or components.
The design rated load of the AB crane remains the same at 125
tons. This load controls the design and supporting analysis. The
auxiliary hook design rated load is being increased from 10 tons to
15 tons. The proposed amendment does not change the currently
acceptable heavy load handling practices in use at KPS. The number
and types of lifts made using this crane in support of KPS plant
operations are not significantly changed from that contemplated
during original plant licensing. Furthermore, the basic operations
of the crane (i.e., hoisting and horizontal travel) remain the same,
although the electronic controls will be upgraded to current
standards.
Therefore, the proposed amendment does not create a new or
different kind of accident from any accident previously evaluated in
the KPS licensing basis.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Although the proposed change is made specifically to support the
upgrade of the KPS AB crane from a non-single-failure-proof to a
single-failure-proof design, the margin of safety under
consideration in this evaluation is mainly based on that contained
within the safety analysis (seismic analysis).
The purpose of this methodology is to determine the stress
placed on the AB cranes' structural components. The stresses
determined by this methodology are then compared to the yield
strength values contained in CMAA-70. If the stresses the structural
component are analyzed to receive during a postulated seismic event
are less than the values contained in CMAA-70 the structural
integrity of the crane is maintained and a suspended load will
remain suspended during a seismic event. Additional margin has been
added by reducing the analysis acceptance criteria to 90% of the
acceptance criteria values contained in CMAA-70, modifying the crane
support structure
[[Page 71708]]
through additional welds and material, and confirming the bolts are
of the proper material.
DEK [Dominion Energy Kewaunee] is modeling the AB crane to roll
during a seismic event when the postulated forces exceed the brake
holding force. This provides a more realistic approach because the
crane trolley is not fixed to the bridge rails. DEK has provided
additional conservatisms by doubling the calculated force needed to
overcome the brake holding force.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc.,
120 Tredegar Street, Richmond, VA 23219.
NRC Acting Branch Chief: Cliff Munson.
Entergy Operations, Inc., Docket Nos. 50-313, Arkansas Nuclear One,
Unit 1, Pope County, Arkansas
Date of amendment request: October 22, 2007.
Description of amendment request: The proposed amendment would
modify the Technical Specification (TS) to establish more effective and
appropriate action, surveillance, and administrative requirements
related to ensuring the habitability of the control room envelope (CRE)
in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task
Force (TSTF) Standard Technical Specification change traveler TSTF-448,
Revision 3, ``Control Room Habitability.'' Specifically, the proposed
amendment would modify TS 3.7.9, ``Control Room Emergency Ventilation
System (CREVS),'' and would establish a CRE habitability (CREH) program
in TS Section 5.5, ``Administrative Controls--Programs and Manuals.''
The NRC staff issued a ``Notice of Availability of Technical
Specification Improvement to Modify Requirements Regarding Control Room
Envelope Habitability Using the Consolidated Line Item Improvement
Process'' associated with TSTF-448, Revision 3, in the Federal Register
on January 17, 2007 (72 FR 2022). The notice included a model safety
evaluation, a model no significant hazards consideration (NSHC)
determination, and a model license amendment request. In its
application dated October 22, 2007, the licensee affirmed the
applicability of the model NSHC determination which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC 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 does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. 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 alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. 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. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on 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 amendments involves NSHC.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Operations, Inc., Docket Nos. 50-313, Arkansas Nuclear One,
Unit 1, Pope County, Arkansas
Date of amendment request: October 22, 2007.
Description of amendment request: The proposed amendment would
modify requirements of Technical Specification (TS) 3.4.12, ``RCS
Specific Activity,'' and TS 3.7.4, ``Secondary Specific Activity,'' as
related to the use of an alternate source term (AST) associated with
accident offsite and control room dose consequences. Implementation of
AST supports adoption of the control room envelope habitability
controls in accordance with Nuclear Regulatory Commission (NRC)-
approved TS Task Force (TSTF) Standard Technical Specification change
traveler TSTF-448, Revision 3, ``Control Room Habitability.''
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.
[[Page 71709]]
The use of an AST is recognized in 10 CFR 50.67 and guidance for
its implementation is provided in RG [Regulatory Guide] 1.183. The
AST involves quantities, isotopic composition, chemical and physical
characteristics, and release timing of radioactive material for use
as inputs to accident dose analyses. As such, the AST cannot affect
the probability of occurrence of a previously evaluated accident. In
addition, the reduction is specific activity limits within the TSs
is unrelated to accident initiators. No facility equipment,
procedure, or process changes are required in conjunction with
implementing the AST that could increase the likelihood of a
previously analyzed accident. The proposed changes in the source
term and the methodology for the dose consequence analyses follow
the guidance of RG 1.183. As a result, there is no increase in the
likelihood of existing event initiators.
Regarding accident consequences, the reduction in specific
activity limits within the TSs is more restrictive (more
conservative) and acts to support the analysis results given the
application of an AST. The results of accident dose analyses using
the AST are compared to TEDE [total effective dose equivalent]
acceptance criteria that account for the sum of deep dose equivalent
(for external exposure) and committed effective dose equivalent (for
internal exposure). Dose results were previously compared to
separate limits on whole body, thyroid, and skin doses as
appropriate for the particular accident analyzed. The results of the
revised dose consequences analyses demonstrate that the regulatory
acceptance criteria are met for each analyzed event. Implementing
the AST involves no facility equipment, procedure, or process
changes that could affect the radioactive material actually released
during an event. Consequently, no conditions have been created that
could significantly increase the consequences of any of the events
being evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any of the events
being 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 AST involves quantities, isotopic composition, chemical and
physical characteristics, and release timing of radioactive material
for use as inputs to accident dose analyses. As such, the AST cannot
create the possibility of a new or different kind of accident. In
addition, the reduction is specific activity limits within the TSs
is unrelated to accident initiators. No facility equipment,
procedure, or process changes have been made in conjunction with
implementing the AST that could initiate or substantially alter the
progression of an accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Implementing the AST is relevant only to calculated accident
dose consequences. The results of the revised dose consequences
analyses demonstrate that the regulatory acceptance criteria are met
for each analyzed event. In addition, the reduction is specific
activity limits within the TSs is unrelated to accident initiators.
No facility equipment, procedure, or process changes are required in
conjunction with implementing the AST that could increase the
exposure of control room or offsite individuals to radioactive
material. The AST does not affect the transient behavior of non-
radiological parameters (e.g., Reactor Coolant System pressure,
Containment pressure) that are pertinent to a margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
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: Thomas G. Hiltz.
Entergy Operations, Inc., Docket Nos. 50-313, Arkansas Nuclear One,
Unit 1, Pope County, Arkansas
Date of amendment request: October 22, 2007.
Description of amendment request: The proposed amendment would
modify Technical Specifications (TS) requirements for mode change
limitations in Limiting Condition for Operation (LCO) 3.0.4 and
Surveillance Requirement (SR) 3.0.4. The proposed TS changes are
consistent with Revision 9 of Nuclear Regulatory Commission (NRC)-
approved Industry TS Task Force (TSTF) Standard TS (STS) change
traveler, TSTF-359, ``Increase Flexibility in Mode Restraints.'' The
amendment would also modify other TSs to reflect the revisions to LCO
3.0.4. The spelling of the word ``not'' is corrected in Section 1.4 of
the TSs.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 2, 2002 (67 FR 50475), as part of the
Consolidated Line Item Improvement Process (CLIIP), on possible
amendments to revise the plant-specific TS to modify requirements for
model change limitations in LCO 3.0.4 and SR 3.0.4.
The NRC staff subsequently issued a notice of availability of the
models for Safety Evaluation and No Significant Hazards Consideration
Determination for referencing in license amendment applications in the
Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed
the applicability of the CLIIP, including the model No Significant
Hazards Consideration Determination, in its application dated October
22, 2007.
The proposed TS changes are consistent with NRC-approved Industry
TSTF STS change, TSTF-359, Revision 8, as modified by 68 FR 16579.
TSTF-359, Revision 8, was subsequently revised to incorporate the
modifications discussed in the April 4, 2003, Federal Register notice
and other minor changes. TSTF-359, Revision 9, was subsequently
submitted to the NRC on April 28, 2003, and was approved by the NRC on
May 9, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the NRC staff's analysis
of the issue of no significant hazards consideration is presented
below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2 --The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, 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
[[Page 71710]]
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3 --The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS Limiting
Conditions for Operation (LCO). The risk associated with this
allowance is managed by the imposition of required actions that must
be performed within the prescribed completion times. The net effect
of being in a TS condition on the margin of safety is not considered
significant. The proposed change does not alter the required actions
or completion times of the TS. The proposed change allows TS
conditions to be entered, and the associated required actions and
completion times to be used in new circumstances. This use is
predicated upon the licensee's performance of a risk assessment and
the management of plant risk. The change also eliminates current
allowances for utilizing required actions and completion times in
similar circumstances, without assessing and managing risk. The net
change to the margin of safety is insignificant. Therefore, this
change does not involve a significant reduction in a margin of
safety.
The NRC staff proposes to determine that the request for amendment
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: Thomas G. Hiltz.
Entergy Operations, Inc., Docket Nos. 50-368, Arkansas Nuclear One,
Unit 2*, Pope County, Arkansas
Date of amendment request: October 22, 2007.
Description of amendment request: The proposed amendment would
modify the Technical Specification (TS) to establish more effective and
appropriate action, surveillance, and administrative requirements
related to ensuring the habitability of the control room envelope (CRE)
in accordance with Nuclear Regulatory Commission (NRC)-approved TS Task
Force (TSTF) Standard Technical Specification change traveler TSTF-448,
Revision 3, ``Control Room Habitability.'' Specifically, the proposed
amendment would modify TS 3.7.6.1, ``Control Room Emergency Ventilation
and Air Condition System,'' and would establish a CRE habitability
(CREH) program in TS Section 6.5, ``Administrative Controls--Programs
and Manuals.'' The NRC staff issued a ``Notice of Availability of
Technical Specification Improvement to Modify Requirements Regarding
Control Room Envelope Habitability Using the Consolidated Line Item
Improvement Process'' associated with TSTF-448, Revision 3, in the
Federal Register on January 17, 2007 (72 FR 2022). The notice included
a model safety evaluation, a model no significant hazards consideration
(NSHC) determination, and a model license amendment request. In its
application dated October 22, 2007, the licensee affirmed the
applicability of the model NSHC determination which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC 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 does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The proposed change
does not alter or prevent the ability of structures, systems, and
components (SSCs) to perform their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change revises the TS for the CRE emergency
ventilation system, which is a mitigation system designed to
minimize unfiltered air leakage into the CRE and to filter the CRE
atmosphere to protect the CRE occupants in the event of accidents
previously analyzed. An important part of the CRE emergency
ventilation system is the CRE boundary. The CRE emergency
ventilation system is not an initiator or precursor to any accident
previously evaluated. Therefore, the probability of any accident
previously evaluated is not increased. Performing tests to verify
the operability of the CRE boundary and implementing a program to
assess and maintain CRE habitability ensure that the CRE emergency
ventilation system is capable of adequately mitigating radiological
consequences to CRE occupants during accident conditions, and that
the CRE emergency ventilation system will perform as assumed in the
consequence analyses of design basis accidents. Thus, the
consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Accident Previously
Evaluated
The proposed change does not impact the accident analysis. The
proposed change does not alter the required mitigation capability of
the CRE emergency ventilation system, or its functioning during
accident conditions as assumed in the licensing basis analyses of
design basis accident radiological consequences to CRE occupants. No
new or different accidents result from performing the new
surveillance or following the new program. The proposed change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. The proposed
change does not alter any safety analysis assumptions and is
consistent with current plant operating practice. 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 alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The proposed change does not affect safety
analysis acceptance criteria. The proposed change will not result in
plant operation in a configuration outside the design basis for an
unacceptable period of time without compensatory measures. 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. Therefore, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the analysis adopted by the licensee
and, based on 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 amendments involves NSHC.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Operations, Inc., Docket Nos. 50-368, Arkansas Nuclear One,
Unit 2, Pope County, Arkansas
Date of amendment request: October 22, 2007.
Description of amendment request: The proposed amendment would
modify Technical Specifications (TS) requirements for mode change
limitations in Limiting Condition for Operation (LCO) 3.0.4 and
Surveillance Requirement (SR) 4.0.4. The proposed TS changes are
consistent with Revision
[[Page 71711]]
9 of Nuclear Regulatory Commission (NRC)-approved Industry TS Task
Force (TSTF) Standard TS (STS) change traveler, TSTF-359, ``Increase
Flexibility in Mode Restraints.'' The amendment would also modify other
TSs to reflect the revisions to LCO 3.0.4. In addition, a change to TS
3.4.3 was made which was determined to be equivalent to the TSTF-359
changes.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 2, 2002 (67 FR 50475), as part of the
Consolidated Line Item Improvement Process (CLIIP), on possible
amendments to revise the plant-specific TS to modify requirements for
model change limitations in LCO 3.0.4 and SR 4.0.4.
The NRC staff subsequently issued a notice of availability of the
models for Safety Evaluation and No Significant Hazards Consideration
Determination for referencing in license amendment applications in the
Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed
the applicability of the CLIIP, including the model No Significant
Hazards Consideration Determination, in its application dated October
22, 2007.
The proposed TS changes are consistent with NRC-approved Industry
TSTF STS change, TSTF-359, Revision 8, as modified by 68 FR 16579.
TSTF-359, Revision 8, was subsequently revised to incorporate the
modifications discussed in the April 4, 2003, Federal Register notice
and other minor changes. TSTF-359, Revision 9, was subsequently
submitted to the NRC on April 28, 2003, and was approved by the NRC on
May 9, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the NRC staff's analysis
of the issue of no significant hazards consideration is presented
below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS Limiting
Conditions for Operation (LCO). The risk associated with this
allowance is managed by the imposition of required actions that must
be performed within the prescribed completion times. The net effect
of being in a TS condition on the margin of safety is not considered
significant. The proposed change does not alter the required actions
or completion times of the TS. The proposed change allows TS
conditions to be entered, and the associated required actions and
completion times to be used in new circumstances. This use is
predicated upon the licensee's performance of a risk assessment and
the management of plant risk. The change also eliminates current
allowances for utilizing required actions and completion times in
similar circumstances, without assessing and managing risk. The net
change to the margin of safety is insignificant. Therefore, this
change does not involve a significant reduction in a margin of
safety.
The NRC staff proposes to determine that the request for amendment
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: Thomas G. Hiltz.
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: October 18, 2007.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to change requirements related to
Emergency Diesel Generator (EDG) fuel oil tank volume, EDG fuel oil
testing and Reactor Building crane inspections.
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 changes do not impact the operability of any
Structure, System or Component that affects the probability of an
accident or that supports mitigation of an accident previously
evaluated. The proposed change does not affect reactor operations or
accident analysis and has no radiological consequences. The
operability requirements for accident mitigation systems remain
consistent with the licensing and design basis. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The proposed change does not involve any physical
alteration of plant equipment and does not change the method by
which any safety-related system performs its function. As such, no
new or different types of equipment will be installed, and the
operation of installed equipment is unchanged. The methods governing
plant operation and testing remain consistent with current safety
analysis assumptions. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The specified margin for onsite fuel oil storage
is maintained and the applicable testing standards and methods
remain unchanged. These changes do not change any existing
requirements, and do not adversely affect existing plant safety
margins or the reliability of the equipment assumed to operate in
the safety analysis. As such, there are no changes being made to
safety analysis assumptions, safety limits or safety system settings
that would adversely affect plant safety as a result of the proposed
change. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
[[Page 71712]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: October 18, 2007.
Description of amendment request: The proposed amendment would
revise the Technical Specifications applicability requirements related
to primary containment oxygen concentration and drywell-to-suppression
chamber differential pressure limits. The associated actions would also
be revised to be consistent with exiting the applicability for each
specification.
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 does not increase the
probability of an accident since it does not involve the
modification of any plant equipment or affect how plant systems or
components are operated, it only changes the requirements for when
inerting and differential pressure need to be established. Whether
the containment is inerted or differential pressure is established
does not impact the likelihood of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability of an accident previously evaluated. The
technical limits (i.e., oxygen concentration and differential
pressure) imposed by the associated Technical Specifications remain
unchanged. Brief periods where the requirements for maintaining
these technical limits are relaxed are currently considered in the
Technical Specifications and associated licensing basis. The
proposed change clarifies the definition of these periods however,
any changes are not considered significant and are supported by
remaining consistent with the recommended allowances of NUREG 1433,
Revision 3. The consequences of analyzed events are therefore not
affected. Therefore, the proposed change does not involve a
significant increase in the consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The proposed 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. As such, no
new or different types of equipment will be installed, and the
operation of installed equipment is unchanged. The methods governing
plant operation and testing remain consistent with current safety
analysis assumptions. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The proposed change does not involve the
modification of any plant equipment or affect basic plant operation.
Additionally, the associated limitations remain unchanged. These
changes do not negate any existing requirement, and do not adversely
affect existing plant safety margins or the reliability of the
equipment assumed to operate in the safety analysis. As such, there
are no changes being made to safety analysis assumptions, safety
limits or safety system settings that would adversely affect plant
safety as a result of the proposed change.
The revised plant conditions reflecting the applicability and
the duration allowed to restore limits are not credited in any
design basis event. These changes do not reflect any significant
adverse impact to the overall risk of operating during brief periods
without the required primary containment oxygen concentration or
differential pressure since the total time for any occurrence is
only marginally extended and reflects times consistent with NUREG-
1433, Revision 3. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: November 20, 2007.
Description of amendment request: The proposed amendment would
revise the values of the safety limit minimum critical power ratio
(SLMCPR) in Technical Specification (TS) Section 2.1.1, ``Reactor Core
SLs.'' Specifically, the proposed change would delete the Quad Cities
Nuclear Power Station (QCNPS) Unit 2 fuel-specific SLMCPR requirements
for Global Nuclear Fuel (GNF) GE14 fuel and consolidate the Unit 1 and
Unit 2 SLMCPR requirements into a bounding dual-unit requirement. This
change is needed to support the next cycle of Unit 2 operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed changes involve a significant increase in
the probability or consequences of an accident previously evaluated?
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC-
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
change to delete the QCNPS Unit 2 fuel-specific SLMCPR requirements
for Global Nuclear Fuel (GNF) GE14 fuel conservatively establishes
the SLMCPR for QCNPS, Unit 2, Cycle 20 at the SLMCPR value for the
co-resident Westinghouse SVEA-96 Optima2 fuel, such that the fuel is
protected during normal operation and during plant transients or
anticipated operational occurrences (AOOs).
The proposed change to delete the GE14 SLMCPR and establish the
requirement at the SLMCPR value for the co-resident Westinghouse
SVEA-96 Optimal fuel does not increase the probability of an
evaluated accident. The change does not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. Therefore, no individual precursors of
an accident are affected.
The proposed change to delete the GE14 SLMCPR and establish the
requirement at the SLMCPR value for the co-resident Westinghouse
SVEA-96 Optimal fuel revises the QCNPS Unit 2 SLMCPR requirement to
protect the fuel during normal operation as well as during plant
transients or AOOs. Operational limits will be established based on
the proposed SLMCPR to ensure that the SLMCPR is not violated. This
will ensure that the fuel design safety criterion (i.e., that at
least 99.9% of the fuel rods do not experience transition boiling
during normal operation and AOOs) is met. Since the proposed change
does not affect operability of plant systems designed to mitigate
any consequences of accidents, the consequences
[[Page 71713]]
of an accident previously evaluated will not increase.
The proposed consolidation of the Unit 1 and Unit 2 SLMCPR
requirements into a bounding dual-unit requirement is
administrative. As such, the proposed consolidation does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
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?
Creation of the possibility of a new or different kind of
accident requires creating one or more new accident precursors. New
accident precursors may be created by modifications of plant
configuration, including changes in allowable modes of operation.
The proposed changes do not involve any plant configuration
modifications or changes to allowable modes of operation. The
proposed change to delete the GE14 SLMCPR and establish the
requirement at the SLMCPR value for the co-resident Westinghouse
SVEA-96 Optimal fuel assures that safety criteria are maintained for
QCNPS, Unit 2, Cycle 20. The proposed consolidation of the Unit 1
and Unit 2 SLMCPR requirements into a bounding dual-unit requirement
is administrative.
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?
The SLMCPR provides a margin of safety by ensuring that at least
99.9% of the fuel rods do not experience transition boiling during
normal operation and AOOs if the SLMCPR limit is not violated. The
proposed change will ensure the current level of fuel protection is
maintained by continuing to ensure that at least 99.9% of the fuel
rods do not experience transition boiling during normal operation
and AOOs if the SLMCPR limit is not violated. The proposed SLMCPR
values were developed using NRC-approved methods. Additionally,
operational limits will be established based on the proposed SLMCPR
to ensure that the SLMCPR is not violated. This will ensure that the
fuel design safety criterion (i.e., that no more than 0.1% of the
rods are expected to be in boiling transition if the MCPR limit is
not violated) is met.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: October 29, 2007.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) for Prairie Island Nuclear
Generating Plant (PINGP) Units 1 and 2 Surveillance Requirement (SR)
3.8.1.9, to require that the test is performed at or below a power
factor of 0.85. The proposed amendments fulfill the commitment made in
Amendments 178 to Unit 1, and 168 to Unit 2, issued on May 30, 2007
(Agency wide Documents Access and Management System (ADAMS) Accession
No. ML071310023).
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 license amendment request proposes more restrictive changes
to the Technical Specification Surveillance Requirements for the
emergency diesel generators which will require testing at a
specified power factor, grid conditions permitting.
The emergency diesel generators are not accident initiators and
therefore, these changes do not involve a significant increase in
the probability of an accident. The proposed changes increase the
load testing requirements, are consistent with the intent of current
regulatory guidance for testing emergency diesel generators, and
will continue to assure that this equipment performs its design
function. Thus these changes do not involve a significant increase
in the consequences 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 proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This license amendment request proposes more restrictive changes
to the Technical Specification Surveillance Requirements for the
emergency diesel generators which will require testing at a
specified power factor, grid conditions permitting.
The changes proposed for the emergency diesel generators do not
change any system operations or maintenance activities. Testing
requirements will be revised and will continue to demonstrate that
the Limiting Conditions for Operation are met and the system
components are functional. These changes do not create new failure
modes or mechanisms which are not identifiable during testing and no
new accident precursors are generated.
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.
This license amendment request proposes more restrictive changes
to the Technical Specification Surveillance Requirements for the
emergency diesel generators which will require testing at a
specified power factor, grid conditions permitting.
The current Technical Specification Surveillance Requirements do
not specify testing at any power factor. The Technical Specification
Surveillance Requirements proposed in this license amendment request
are thus more restrictive in that they place additional restraints
on the test conditions. These changes may make the testing more
rigorous and thus more difficult for the emergency diesel generators
to meet the test acceptance criteria. The addition of a power factor
is consistent with the intent of current regulatory guidance for
testing emergency diesel generators. Since these changes are an
increase in the test requirements and are consistent with the intent
of current regulatory guidance, these changes do not involve a
significant reduction in a margin of safety.
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 requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Cliff Munson.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant (PINGP), Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: November 19, 2007.
Description of amendment request: The proposed amendments would
revise Technical Specifications for the PINGP, Units 1 and 2, to
replace the
[[Page 71714]]
current fixed Frequency for testing the containment spray nozzles in
Surveillance Requirement 3.6.5.8 with a maintenance or event based
Frequency.
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 license amendment request proposes Technical Specification
Surveillance Requirement changes which will require verification
that the containment spray system spray nozzles are unobstructed
following maintenance which could result in nozzle blockage.
The containment spray system and its spray nozzles are not
accident initiators and therefore, these changes do not involve a
significant increase the probability of an accident. The revised
surveillance requirement will require event based verification in
lieu of fixed Frequency verification which may require either fewer
or more verifications of operability. The proposed changes to verify
system operability following maintenance is considered adequate to
ensure operability of the containment spray system. Since the system
continues to be available to perform its accident mitigation
function, the consequences of accidents previously evaluated are not
significantly increased.
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.
This license amendment request proposes Technical Specification
Surveillance Requirement changes which will require verification
that the containment spray system spray nozzles are unobstructed
following maintenance which could result in nozzle blockage.
The proposed change does not introduce a new mode of plant
operation and does not involve physical modification to the plant.
The change does not introduce new accident initiators or impact the
assumption made in the safety analysis. Testing requirements will be
revised and will continue to demonstrate that the Limiting
Conditions for Operation are met and the system components are
functional.
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.
This license amendment request proposes Technical Specification
Surveillance Requirement changes which will require verification
that the containment spray system spray nozzles are unobstructed
following maintenance which could result in nozzle blockage.
The containment spray system is not susceptible to corrosion-
induced obstruction or obstruction from sources external to the
system. Maintenance activities that could introduce foreign material
into the system would require subsequent verification to ensure
there is no spray nozzle blockage. The spray header nozzles are
expected to remain unblocked and available in the event that the
safety function is required. Therefore, the capacity of the system
would remain unaffected.
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 requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Clifford G. Munson.
Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power
Plant (HBPP), Unit 3 Humboldt County, California J00336
Date of amendment request: November 5, 2007.
Description of amendment request: The licensee has proposed
amending the technical specifications (TS) to delete many operational
and administrative requirements upon transfer of spent nuclear fuel
assemblies and fuel fragment containers from the Spent Fuel Pool (SFP)
to the Humboldt Bay Independent Spent Fuel Storage Installation
(ISFSI). Some TS requirements will be relocated to the HBPP Quality
Assurance Plan.
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.
The proposed changes reflect the transfer of spent fuel from the
Spent Fuel Pool to the Humboldt Bay (HB) Independent Spent Fuel
Storage Installation. Design basis accidents related to the SFP are
discussed in the Humboldt Bay Power Plant Unit 3 Defueled Safety
Analysis Report (DSAR). These postulated accidents are predicated on
spent fuel being stored in the SFP. With the removal of the spent
fuel from the SFP, there are no important-to-safety systems,
structures or components required to function or to be monitored. In
addition, there are no remaining credible accidents involving spent
fuel or the SFP that require actions of a Certified Fuel Handler or
Noncertified Fuel Handler to prevent occurrence or to mitigate
consequences. The proposed change to the Design Features section of
the Technical Specifications (TS) clarifies that the spent fuel is
being stored in dry casks within an ISFSI. The probability or
consequences of accidents at the ISFSI are evaluated in the HB ISFSI
Final Safety Analysis Report (FSAR) and are independent of the
accidents evaluated in the HBPP Unit 3 DSAR. Therefore, the proposed
changes will 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 evaluated?
Response: No.
The proposed changes reflect the reduced operational risks as a
result of the spent fuel being transferred to dry casks within an
ISFSI. The proposed changes do not modify any systems, structures or
components. The plant conditions for which the HBPP Unit 3 DSAR
design basis accidents relating to spent fuel and the SFP have been
evaluated are no longer applicable. The aforementioned proposed
changes do not affect any of the parameters or conditions that could
contribute to the initiation of an accident. Design basis accidents
associated with the dry cask storage of spent fuel are already
considered in the HB ISFSI FSAR. No new accident scenarios are
created as a result of deleting nonapplicable operational and
administrative requirements. Therefore, the proposed changes will
not create the possibility of a new or different kind of accident
from those previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
The proposed changes reflect the reduced operational risks as a
result of the spent fuel being transferred to dry casks within an
ISFSI. The design basis and accident assumptions within the HBPP
Unit 3 DSAR and the TS relating to spent fuel are no longer
applicable. The proposed changes do not affect remaining plant
operations, nor structures, systems, or components supporting
decommissioning activities. In addition, the proposed changes do not
result in a change in initial conditions, system response time, or
in any other parameter affecting the course of a decommissioning
activity accident analysis. Therefore, the proposed changes will not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
[[Page 71715]]
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Jennifer K. Post, Pacific Gas and
Electric Company, 77 Beale Street, B30A, San Francisco, CA.
NRC Branch Chief: Andrew Persinko.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: November 30, 2007.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) Sections TS 5.5.9, ``Steam
Generator (SG) Program'' and TS 5.6.10, ``Steam Generator Tube
Inspection Report.'' The proposed changes to TS 5.5.9 modify the
inspection and plugging requirements for portions of SG tubes within
the hot leg side of the tubesheet region of the SGs only. The proposed
changes to TS 5.6.10 will add requirements to report specific data
related to indications, leakage detected, and calculated accident
leakage.
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 license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No. The previously analyzed accidents are initiated by the
failure of plant structures, systems, or components. The proposed
changes that alter the SG inspection criteria do not have a
detrimental impact on the integrity of any plant structure, system,
or component that initiates an analyzed event. The proposed changes
will not alter the operation of, or otherwise increase the failure
probability of, any plant equipment that initiates an analyzed
accident. Therefore, the proposed change does not involve a
significant increase in the probability of an accident previously
evaluated.
Of the applicable accidents previously evaluated, the limiting
transients with consideration to the proposed changes to the SG tube
inspection criteria, are the SG tube rupture (SGTR) event and the
steam line break (SLB) accident.
During the SGTR event, the required structural integrity margins
of the SG tubes will be maintained by the presence of the SG
tubesheet. SG tubes are hydraulically expanded in the tubesheet
area. Tube rupture in tubes with cracks in the tubesheet is
precluded by the constraint provided by the tubesheet. This
constraint results from the hydraulic expansion process, thermal
expansion mismatch between the tube and tubesheet and from the
differential pressure between the primary and secondary side. Based
on this design, the structural margins against burst discussed in RG
1.121 (Reference 4) [Regulatory Guide 1.121, ``Bases for Plugging
Degraded PWR Steam Generator Tubes,'' dated August 1976], are
maintained for both normal and postulated accident conditions.
The proposed changes do not affect other systems, structures,
components or operational features. Therefore, the proposed changes
result in no significant increase in the probability of the
occurrence of a SGTR accident.
At normal operating pressures, leakage from primary water stress
corrosion cracking (PWSCC) below the proposed limited inspection
depth is limited by both the tube-to-tubesheet crevice and the
limited crack opening permitted by the tubesheet constraint.
Consequently, negligible normal operating leakage is expected from
cracks within the tubesheet region. The consequences of a SGTR event
are affected by the primary-to-secondary leakage flow during the
event. Primary-to-secondary leakage flow through a postulated broken
tube is not affected by the proposed change since the tubesheet
enhances the tube integrity in the region of the hydraulic expansion
by precluding tube deformation beyond its initial hydraulically
expanded outside diameter.
The probability of a SLB is unaffected by the potential failure
of a SG tube, since this failure is not an initiator for a SLB.
The consequences of a SLB are also not significantly affected by
the proposed changes. During a SLB accident, the reduction in
pressure above the tubesheet on the shell side of the SG creates an
axially uniformly distributed load on the tubesheet due to the
reactor coolant system pressure on the underside of the tubesheet.
The resulting bending action constrains the tubes in the tubesheet,
thereby restricting primary-to-secondary leakage below the midplane.
The purpose of the tube-end weld is to ensure the hydraulically
expanded tube-to-tubesheet joints in Model F SGs are leak-tight.
Considerations were also made with regard to the potential for
primary-to-secondary leakage during postulated faulted conditions.
However, the leak rate during postulated accident conditions would
be expected to be less than that during normal operation for
indications near the bottom of the tubesheet based on the evaluation
(Reference 1) [Westinghouse Electric Company WCAP-16794-P, ``Steam
Generator Tube Alternate Repair Criteria for the Portion of the Tube
Within the Tubesheet at the Vogtle 1 & 2 Electric Generating
Plants,'' dated October 2007] which shows that while the driving
pressure increases by about a factor of almost two, the flow
resistance increases, because the tube-to-tubesheet contact pressure
also increases. Depending on the depth within the tubesheet, the
relative increase in resistance could easily be larger than that of
the pressure potential. Therefore, the leak rate under normal
operating conditions could exceed its allowed value before the
accident condition leak rate would be expected to exceed its allowed
value. This approach is termed an application of the ``bellwether
principle.'' While such a decrease in the leak rate is expected, the
postulated accident leak rate could conservatively be taken to be
bounded by twice the normal operating leak rate if the increase in
contact pressure is ignored.
Since normal operating leakage is limited by VEGP TS 3.4.13 and
by NEI 97-06 (Reference 3) [NEI 97-06, ``Steam Generator Program
Guidelines,'' Revision 2, dated May 2, 2005] to less than 150 gpd
throughout one SG in the VEGP Units 1 and 2 SGs, the attendant
accident condition leak rate, assuming all leakage to be from lower
tubesheet indications, would be bounded by 0.20 gpm in the faulted
SG which is less than the accident analysis assumption of 0.35 gpm
to the affected SG included in Section 15.1.5 of the VEGP FSAR.
Hence, it is reasonable to omit any consideration of inspection of
the tube, tube end weld, bulges/overexpansions or other anomalies
below 17 inches from the top of the hot leg tubesheet.
Based on the above discussion, the proposed changes do not
involve an increase in the consequences of an accident previously
evaluated.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
No. The proposed changes do not involve the use or installation
of new equipment and the currently installed equipment will not be
operated in a new or different manner. No new or different system
interactions are created and no new processes are introduced. The
proposed changes will not introduce any new failure mechanisms,
malfunctions, or accident initiators not already considered in the
design and licensing bases.
Based on this evaluation, 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?
No. The proposed changes maintain the required structural
margins of the SG tubes for both normal and accident conditions. NEI
97-06 (Reference 3) and RG 1.121 (Reference 4), are used as the
bases in the development of the limited tubesheet inspection depth
methodology for determining that SG tube integrity considerations
are maintained within acceptable limits. RG 1.121 (Reference 4)
describes a method acceptable to the NRC for meeting the following
General Design Criteria (GDC).
GDC 14, ``Reactor coolant pressure boundary,''
GDC 15, ``Reactor coolant system design,''
GDC 31, ``Fracture prevention of reactor coolant
pressure boundary,'' and,
GDC 32, ``Inspection of reactor coolant pressure
boundary.''
RG 1.121 concludes that by determining the limiting safe
conditions for tube wall degradation, the probability and
consequences of a SGTR are reduced. This RG uses safety factors on
loads for tube burst that are consistent with the requirements of
Section III of the ASME Code [American Society of Mechanical
Engineers, Boiler and Pressure Vessel Code].
Application of the limited tubesheet inspection depth criteria
will preclude
[[Page 71716]]
unacceptable primary-to-secondary leakage during all plant
conditions. The methodology for determining leakage provides for
large margins between calculated and actual leakage values in the
proposed limited tubesheet inspection depth criteria.
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. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Branch Chief: Evangelos C. Marinos.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: October 23, 2007.
Description of amendment request: The amendments will relocate the
surveillance test intervals of various Technical Specifications (TSs)
to a licensee-controlled program (risk-informed Initiative 5(b)) in
accordance with the Surveillance Frequency Control Program, which is
being added to the Administrative Controls section of the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. [Do] the proposed change[s] involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change[s] [involve] the relocation of various
surveillance test intervals from Technical Specifications (TS) to a
licensee-controlled program. The proposed change[s] [do] not involve
the modification of any plant equipment or affect basic plant
operation. The proposed change[s] will have no impact on the design
or function of any safety related structures, systems or components.
Surveillance test intervals are not assumed to be an initiator of
any analyzed event, nor are they assumed in the mitigation of
consequences of accidents. The surveillance requirements themselves
will be maintained in the TS along with the applicable Limiting
Conditions for Operation (LCOs) and Action statements. The
surveillances performed at the intervals specified in the licensee-
controlled program will assure that the affected system or component
function is maintained, that the facility operation is within the
Safety Limits, and that the LCOs are met.
Therefore, the proposed change[s] [do] not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. [Do] the proposed change[s] create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change[s] [do] not involve any physical alteration
of plant equipment and does not change the method by which any
safety-related structure, system, or component performs its function
or is tested. As such, no new or different types of equipment will
be installed, and the basic operation of installed equipment is
unchanged.
The methods governing plant operation and testing remain
consistent with current safety analysis assumptions.
Therefore, the proposed change[s] will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. [Do] the proposed change[s] involve a significant reduction
in a margin of safety?
Response: No.
The proposed change[s] [do] not negate any existing requirement,
and [do] not adversely affect existing plant safety margins or the
reliability of the equipment assumed to operate in the safety
analysis. As such, there are no changes being made to safety
analysis assumptions, safety limits or safety system settings that
would adversely affect plant safety as a result of the proposed
change. Margins of safety are unaffected by relocation of the
surveillance test intervals to a licensee-controlled program.
Therefore, the proposed change[s] [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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Thomas G. Hiltz.
U.S. Department of Transportation (USDOT), United States Maritime
Administration (MARAD), License No. NS-1, Docket No. 50-238, Nuclear
Ship Savannah (NSS)
Date of amendment request: October 9, 2007.
Description of amendment request:
The proposed license amendment would modify the Technical
Specification (TS) requirements to clarify the TS and make the
requirements commensurate with the current ship status and
decommissioning schedule. Thirty-nine TS changes are proposed. The
proposed changes modify the TS as follows:
Delete requirements more appropriate for the Final Safety
Analysis Report;
Provide consistent titles and phrases;
Delete duplicate requirements;
Organize similar requirements into single locations;
Remove requirements that can be implemented through
current regulations;
Delete archaic requirements;
Invoke requirements commensurate with current ship status
and decommissioning schedule;
Format and renumber, as appropriate;
Revise requirements to reflect historical practices;
Revise TS to be consistent with the Decommissioning
Quality Assurance Plan; and
Correct errors introduced in License Amendment 13,
Reference (a).
The application for license amendment is available electronically
at the NRC's Electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this site, you can access the NRC's Agencywide
Document Access and Management System (ADAMS), which provides text and
image files of NRC's public documents. The ADAMS accession number for
the October 9, 2007, request is ML072880143.
If you do not have access to ADAMS, or if there are problems in
accessing the documents located in ADAMS, contact the NRC Public
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737, or
by e-mail to [email protected]. These documents may also be viewed
electronically on the public computers located at the NRC's PDR, O 1
F21, One White Flint North, 11555 Rockville Pike, Rockville, MD 20852.
The PDR reproduction contractor will copy documents for a fee.
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.
The proposed changes are administrative and do not involve
modification of any plant equipment or affect basic plant operation.
The NSS's reactor is not operational and the level of radioactivity
in the NSS has significantly decreased from the levels that
[[Page 71717]]
existed when the 1976 Possession-only License was issued. No aspect
of any of proposed changes is and initiator of any accident
previously evaluated. Consequently, the probability of an accident
previously evaluated is not significantly increased.
Therefore, the proposed changes no 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 evaluated?
Response: No.
All of the proposed changes are administrative and do not
involve physical alteration of plant equipment that was not
previously allowed by Technical Specifications. These proposed
changes do not change the method by which any safety-related system
performs its function. As such, no new or different types of
equipment will be installed, and the basic operation of installed
equipment is unchanged. The methods governing plant operation and
testing remain consistent with current safety analysis assumptions.
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.
All of the proposed changes are administrative in nature. No
margins of safety exist that are relevant to the ship's defueled and
partially dismantled reactor. 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 changes. The proposed changes involve movement of the ship,
changes in the performance of responsibilities and reflect
significantly improved radiological conditions since 1976.
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 upon
the staff's review of the licensee's analysis, as well as the staff's
own evaluation, the staff concludes 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.
Senior Technical Advisor, N.S. Savannah: Erhard W. Koehler, MARAD,
Office of Ship Disposal Programs.
NRC Branch Chief: Andrew Persinko.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: October 24, 2007.
Description of amendment request: The amendments would revise the
Technical Specifications (TS) Limiting Condition for Operations (LCO)
3.8.7 and 3.8.9, pertaining to electrical power systems and
distribution associated with the 120 Volt AC vital bus inverters. The
TS changes are intended to support operability of components shared
between Unit 1 and Unit 2. The proposed changes will add new
Conditions, Required Action statements and Completion Times for LCO
3.8.7 and LCO 3.8.9 to address shared components.
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?
The proposed amendment does not involve a significant increase
in the probability or consequence of an accident previously
analyzed. There is no change to how or under what conditions the
inverters or 120 VAC vital buses are operated, nor are there any
changes to acceptable operating parameters. Operability
requirements, which are consistent with current operation of the
inverters and vital buses, are being established for the inverters
and vital buses associated with shared systems. The proposed change
will ensure that there is an operable electrical control circuit for
the Auxiliary Building Central Exhaust subsystem filter and bypass
dampers for each train of the [Emergency Core Cooling System Pump
Room Exhaust Air Cleanup System] ECCS PREACS which will ensure that
the evaluated dose consequences for [design basis accidents] DBAs
will not be exceeded.
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?
The implementation of the proposed changes does not create the
possibility of an accident of a different type than was previously
evaluated in the [Updated Final Safety Analysis Report] UFSAR. There
is no change to how or under what conditions the inverters or 120
VAC vital buses are operated nor are there any changes to acceptable
operating parameters. The proposed operability requirements, which
are consistent with current operation of the inverters and vital
buses, are being established for the inverters and vital buses
associated with shared systems. The proposed changes ensure vital
120 VAC power is available to support operation of the Auxiliary
Building Central Exhaust subsystems. These changes do not alter the
nature of events postulated in the UFSAR nor do they introduce any
unique precursor 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 proposed amendment] involve a significant reduction
in the margin of safety?
The implementation of the proposed changes does not reduce the
margin of safety. The proposed changes for the 120 VAC Vital Bus
System and Inverters do not affect the ability of these systems or
components to perform their intended safety functions to provide
power to required safety and monitoring systems or components.
Operability requirements, which are consistent with current
operation of the inverters and vital buses, are being established
for the inverters and vital buses associated with shared systems.
These changes provide additional assurance that the Auxiliary
Building Central Exhaust subsystems will operate to maintain the
margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond,
VA 23219.
NRC Branch Chief: Evangelos C. Marinos.
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
[[Page 71718]]
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].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina.
Date of application for amendments: January 22, 2007, as
supplemented by letter dated September 28, 2007.
Brief Description of amendments: The amendments change the
Technical Specifications (TSs) related to the fuel design description
and the fuel criticality methods to accommodate the transition to AREVA
NP fuel.
Date of issuance: November 27, 2007.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment Nos.: 243 and 271.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
changed the TSs.
Date of initial notice in Federal Register: August 29, 2007 (72 FR
49742). The supplement dated September 28, 2007, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 27, 2007.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 17, 2007.
Brief description of amendment: The amendment revised the date for
performing the ``Type A test'' in the River Bend Station, Unit 1,
Technical Specification 5.5.13, ``Primary Containment Leak Rate Testing
Program,'' from ``prior to December 14, 2007,'' to ``prior to April 14,
2008.''
Date of issuance: December 3, 2007.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 155.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 11, 2007 (72
FR 51857). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 3, 2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC (EGC), Docket Nos. STN 50-454 and STN
50-455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Units 1
and 2, Will County, Illinois.
Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units
2 and 3, Grundy County, Illinois.
Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and
2, LaSalle County, Illinois.
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois.
EGC and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-278, Peach
Bottom Atomic Power Station, Units 2 and 3 (PBAPS), York and Lancaster
Counties, Pennsylvania.
Date of application for amendments: December 15, 2006.
Brief description of amendments: The amendments modify the
technical specifications (TSs) by replacing the term ``plant-specific''
with ``generic'' when discussing job titles in TS Section 5.2.1.a. This
revision will ensure the TS description is consistent with the licensee
Quality Assurance Topical Report (QATR). The proposed amendment will
also revise the PBAPS TS Section 5.2.1.a to replace the reference to
the Updated Final Safety Analysis Report with reference to the EGC
QATR. This change aligns the PBAPS TS wording with the rest of the
licensee fleet.
Date of issuance: November 19, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days of the date of issuance.
Amendment Nos.: 152, 152, 147, 147, 225, 217, 187, 174, 265, 269,
236, and 231.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72, NPF-77,
DPR-19, DPR-25, NPF-11, NPF-18, DPR-29, DPR-30, DRP-44, and DPR-56: The
amendments revised the Technical Specifications and Operating Licenses.
Date of initial notice in Federal Register: March 13, 2007 (72 FR
11387).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 19, 2007.
No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell
County, Texas
Date of amendment request: December 19, 2006.
Brief description of amendments: Amendments revise the requirements
in Technical Specification (TS) 5.5.8, ``Inservice Testing Program,''
to update references to the American Society of Mechanical Engineers
(ASME) Boiler and Pressure Vessel Code, Section XI, as the source of
requirements for the inservice testing of ASME Code Class 1, 2, and 3
pumps and valves, and address the applicability of Surveillance
Requirement 3.0.2 to other normal and accelerated frequencies specified
as 2 years or less in the Inservice Testing Program.
Date of issuance: December 4, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days of the date of issuance.
Amendment Nos.: Unit 1-140; Unit 2-140.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: May 22, 2007 (72 FR
28724). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 4, 2007.
No significant hazards consideration comments received: No.
[[Page 71719]]
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: August 16, 2007, as supplemented by
letter dated November 5, 2007.
Brief description of amendment: The amendment revised Technical
Specification 5.5.6, ``Inservice Testing Program,'' to allow a one-time
extension of the 5-year frequency requirement for setpoint testing of
safety valve MS-RV-70ARV.
Date of issuance: December 4, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 228.
Facility Operating License No. DPR-46: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 25, 2007 (72
FR 54476). The supplement dated November 5, 2007, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
initially published in the Federal Register. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
December 4, 2007.
No significant hazards consideration comments received: No.
Sacramento Municipal Utility District, Docket No. 50-312, Rancho Seco
Nuclear Generating Station, Sacramento County, California
Date of application for amendment: April 12, 2006, and supplemented
November 21, 2006.
Brief description of amendment: The amendment incorporates the
Nuclear Regulatory Commission (NRC) approved, License Termination Plan
(LTP), and associated addendum, into the Rancho Seco license and
specifies limits on the changes the licensee is allowed to make to the
approved LTP without prior NRC review and approval.
Date of issuance: November 26, 2007.
Effective date: November 26, 2007.
Amendment No: 133.
Facility Operating License No. DPR-54: The amendment revised the
License.
Date of initial notice in Federal Register: February 13, 2007 (72
FR 6789).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 26, 2007.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: July 14, 2006, as supplemented
by letters dated June 28, September 26, and November 2, 2007.
Brief description of amendments: The amendments incorporate a
description of the parent tube inspection limitation adjacent to the
nickel band portion of the lower sleeve joint and provide the basis for
the structural and leakage integrity of the joint being ensured with
the existing inspection of the parent tube adjacent to the nickel band
region.
Date of issuance: November 29, 2007.
Effective date: As of its date of issuance, to be implemented
within 60 days of issuance.
Amendment Nos.: Unit 2-215; Unit 3-207.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: September 12, 2006 (71
FR 53720). The supplements dated June 28, September 26, and November 2,
2007, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register. The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated November 29, 2007.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: June 5, 2007, as supplemented
June 11, 2007.
Brief description of amendments: The amendments revised the
Technical Specifications testing frequency for surveillance requirement
3.1.4, ``Control Rod Scram Times,'' from ``120 days cumulative
operation in MODE 1'' to ``200 days cumulative operation in MODE 1.''
Date of issuance: November 26, 2007.
Effective date: As of the date of issuance and shall be implemented
within 45 days from the date of issuance.
Amendment Nos.: 254, 198.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: July 17, 2007, (72 FR
39084).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 26, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 10th day of December 2007.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E7-24284 Filed 12-17-07; 8:45 am]
BILLING CODE 7590-01-P