[Federal Register Volume 76, Number 172 (Tuesday, September 6, 2011)]
[Notices]
[Pages 55125-55136]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-22541]
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NUCLEAR REGULATORY COMMISSION
[NRC-2011-0205]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
Background
Pursuant to Section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC) is publishing this regular biweekly notice. The Act
requires the Commission publish notice of any amendments issued, or
proposed to be issued and grants the Commission the authority to issue
and make immediately effective any amendment to an operating license
upon a determination by the Commission that such amendment involves no
significant hazards consideration, notwithstanding the pendency before
the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 11, 2011 to August 24, 2011. The last
biweekly notice was published on August 23, 2011 (76 FR 52699).
ADDRESSES: Please include Docket ID NRC-2011-0205 in the subject line
of your comments. Comments submitted in writing or in electronic form
will be posted on the NRC Web site and on the Federal rulemaking Web
site http://www.regulations.gov. Because your comments will not be
edited to remove any identifying or contact information, the NRC
cautions you against including any information in your submission that
you do not want to be publicly disclosed.
The NRC requests that any party soliciting or aggregating comments
received from other persons for submission to the NRC inform those
persons that the NRC will not edit their comments to remove any
identifying or contact information, and therefore, they should not
include any information in their comments that they do not want
publicly disclosed.
You may submit comments by any one of the following methods:
Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for documents filed under Docket ID NRC-
2011-0205. Address questions about NRC dockets to Carol Gallagher 301-
492-3668; e-mail [email protected].
Mail comments to: Chief, Rules, Announcements, and
Directives Branch (RADB), Office of Administration, Mail Stop: TWB-05-
B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
Fax comments to: RADB at 301-492-3446.
You can access publicly available documents related to this notice
using the following methods:
NRC's Public Document Room (PDR): The public may examine
and have copied, for a fee, publicly available documents at the NRC's
PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike,
Rockville, Maryland 20852.
NRC's Agencywide Documents Access and Management System
(ADAMS): Publicly available documents created or received at the NRC
are accessible electronically through ADAMS in the NRC Library at
http://www.nrc.gov/reading-rm/adams.html. From this page, the public
can gain entry into ADAMS, which provides text and image files of the
NRC's public documents. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC's PDR reference staff at 1-800-397-4209, 301-415-4737, or by e-mail
to [email protected].
Federal Rulemaking Web Site: Public comments and
supporting materials related to this notice can be found at http://www.regulations.gov by searching on Docket ID: NRC-2011-0205.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Title 10 of the Code of Federal
Regulations (10 CFR), Section 50.92, this means that operation of the
facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the
[[Page 55126]]
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.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the NRC's PDR, located at One White Flint North, Room O1-F21, 11555
Rockville Pike (first floor), Rockville, Maryland 20852. NRC
regulations are accessible electronically from the NRC Library on the
NRC Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If
a request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner 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 requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. 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 requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule (72 FR 49139,
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by e-mail at [email protected], or by
telephone at 301-415-1677, to request (1) A digital identification (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 (2) advise
the Secretary that the participant will be submitting a request or
petition for hearing (even in instances in which the participant, or
its counsel or representative, already holds an NRC-issued digital ID
certificate). Based upon this information, the Secretary will establish
an electronic docket for the hearing in this proceeding if the
Secretary has not already established an electronic docket.
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. System requirements for accessing
the E-Submittal server are detailed in NRC's ``Guidance for Electronic
Submission,'' which is available on the agency's public Web site at
http://www.nrc.gov/site-help/e-submittals.html. Participants may
attempt to use other software not listed on the Web site, but should
note that the NRC's E-Filing system does not support unlisted software,
and the NRC Meta System Help Desk will not be able to offer assistance
in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant 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
[[Page 55127]]
http://www.nrc.gov/site-help/e-submittals.html. A filing is considered
complete at the time the documents are submitted through the NRC's E-
Filing system. To be timely, an electronic filing must be submitted to
the E-Filing 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 E-Filing system also distributes an e-mail
notice that provides access to the document to the NRC Office of the
General Counsel and any others who have advised the Office of the
Secretary that they wish to participate in the proceeding, so that the
filer need not serve the documents on those participants separately.
Therefore, applicants and other participants (or their counsel or
representative) must apply for and receive a digital ID certificate
before a hearing request/petition to intervene is filed so that they
can obtain access to the document via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at 1-866-672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/EHD/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information. With
respect to copyrighted works, except for limited excerpts that serve
the purpose of the adjudicatory filings and would constitute a Fair Use
application, participants are requested not to include copyrighted
materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Non-timely filings
will not be entertained absent a determination by the presiding officer
that the petition or request should be granted or the contentions
should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS, should contact the NRC PDR Reference staff at 1-800-397-4209,
301-415-4737, or by e-mail to [email protected].
Carolina Power and Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendment request: July 12, 2011.
Description of amendment request: The proposed license amendments
would revise Technical Specification (TS) 3.4.5, ``Reactor Coolant
System (RCS) Leakage Detection Instrumentation,'' to define a new time
limit for restoring inoperable RCS leakage detection instrumentation to
operable status and establish alternate methods of monitoring RCS
leakage when one or more required monitors are inoperable. These
proposed changes would be consistent with Standard Technical
Specifications Change Traveler (TSTF)-514, ``Revise BWR Operability
Requirements and Actions for RCS Leakage Instrumentation.'' The
availability of TSTF-514 was announced in the Federal Register on
December 17, 2010 (75 FR 79048), as part of the consolidated line item
improvement process.
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 clarifies the operability requirements for
the RCS leakage detection instrumentation and reduces the time
allowed for the plant to operate when the only TS-required operable
RCS leakage detection instrumentation monitor is the primary
containment atmosphere gaseous radioactivity monitor. The monitoring
of RCS leakage is not a precursor to any accident previously
evaluated. The monitoring of RCS leakage is not used to mitigate the
consequences of any accident previously evaluated.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
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 clarifies the operability requirements for
the RCS leakage detection instrumentation and reduces the time
allowed for the plant to operate when the only TS required operable
RCS leakage detection instrumentation monitor is the primary
containment atmosphere gaseous radioactivity monitor. 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 change in
the methods governing normal plant operation.
Therefore, it is concluded that 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
[[Page 55128]]
The proposed change clarifies the operability requirements for
the RCS leakage detection instrumentation and reduces the time
allowed for the plant to operate when the only TS-required operable
RCS leakage detection instrumentation monitor is the primary
containment atmosphere gaseous radioactivity monitor. Reducing the
amount of time the plant is allowed to operate with only the primary
containment atmosphere gaseous radioactivity monitor operable
increases the margin of safety by increasing the likelihood that an
increase in RCS leakage will be detected before it potentially
results in gross failure.
Therefore, it is concluded that the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Douglas A. Broaddus.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
1, Pope County, Arkansas
Date of amendment request: April 29, 2011.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.4.15, ``RCS [Reactor Coolant
System] Leakage Detection Instrumentation,'' to define a new time limit
for restoring inoperable RCS leakage detection instrumentation to
operable status; establish alternate methods of monitoring RCS leakage
when one or more required monitors are inoperable; and make TS Bases
changes which reflect the proposed changes and more accurately reflect
the contents of the facility design basis related to operability of the
RCS leakage detection instrumentation. New Condition C is applicable
when the reactor building atmosphere gaseous radioactivity monitor is
the only operable TS-required monitor. New Condition C Required Actions
require analyzing grab samples of the reactor building atmosphere every
12 hours and restoring another monitor within 7 days. These changes are
consistent with NRC-approved Revision 3 to Technical Specification Task
Force (TSTF) Standard Technical Specification (STS) Change Traveler
TSTF-513, ``Revise PWR [Pressurized-Water Reactor] Operability
Requirements and Actions for RCS Leakage Instrumentation.'' The
availability of this TS improvement was announced in the Federal
Register on January 3, 2011 (76 FRN 189), as part of the consolidated
line item improvement process (CLIIP).
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 clarifies the operability requirements for
the RCS leakage detection instrumentation and reduces the time
allowed for the plant to operate when the only TS-required operable
RCS leakage detection instrumentation monitor is the reactor
building atmosphere gaseous radiation monitor. The monitoring of RCS
leakage is not a precursor to any accident previously evaluated. The
monitoring of RCS leakage is not used to mitigate the consequences
of any accident previously evaluated.
Therefore, it is concluded that 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 clarifies the operability requirements for
the RCS leakage detection instrumentation and reduces the time
allowed for the plant to operate when the only TS-required operable
RCS leakage detection instrumentation monitor is the reactor
building atmosphere gaseous radiation monitor. The proposed change
does not involve a physical alteration of the plant (no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operation. The proposed change
maintains sufficient continuity and diversity of leak detection
capability that the probability of piping evaluated and approved for
Leak-Before-Break progressing to pipe rupture remains extremely low.
Therefore, it is concluded that the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does the Proposed Change Involve a Significant Reduction in a
Margin of Safety?
Response: No.
The proposed change clarifies the operability requirements for
the RCS leakage detection instrumentation and reduces the time
allowed for the plant to operate when the only TS-required operable
RCS leakage detection instrumentation monitor is the reactor
building atmosphere gaseous radiation monitor. Reducing the amount
of time the plant is allowed to operate with only the reactor
building atmosphere gaseous radiation monitor operable increases the
margin of safety by increasing the likelihood that an increase in
RCS leakage will be detected before it potentially results in gross
failure.
Therefore, it is concluded that 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: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and
3, York and Lancaster Counties, Pennsylvania
Date of amendment request: April 6, 2011.
Description of amendment request: The proposed amendment would
modify the actions to be taken when the atmospheric gaseous
radioactivity monitor is the only operable reactor coolant leakage
detection instrument. The modified actions require additional, more
frequent monitoring of other indications of Reactor Coolant System
(RCS) leakage and provide appropriate time to restore another leakage
detection instrument to operable status. This change is consistent with
the U.S. Nuclear Regulatory Commission (NRC) approved safety evaluation
on Technical Specification Task Force (TSTF) Traveler, TSTF-514-A,
Revision 3, ``Revised BWR [boiling-water reactor] Operability
Requirements and Actions for RCS Leakage Instrumentation'' dated
November 24, 2010.
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, with NRC edits in brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No
The proposed changes [ ] modify the time allowed for the plant
to operate when the only Operable RCS leakage detection
[[Page 55129]]
instrumentation monitor is the atmospheric gaseous radiation
monitor. The monitoring of RCS leakage is not a precursor to any
accident previously evaluated. The monitoring of RCS leakage is not
used to mitigate the consequences of any accident previously
evaluated.
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.
The proposed changes [ ] modify the time allowed for the plant
to operate when the only Operable RCS leakage detection monitor is
the atmospheric gaseous radiation monitor. The proposed changes do
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, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes [ ] increase the time allowed for the plant
to operate when the only Operable RCS leakage detection
instrumentation monitor is the atmospheric gaseous radiation monitor
from 24 hours to 7 days. Increasing the amount of time the plant is
allowed to operate with only the atmospheric gaseous radiation
monitor Operable does not significantly decrease the margin of
safety due to the addition of compensatory Required Actions to
analyze grab samples of the primary containment atmosphere once per
12 hours and monitor Reactor Coolant System leakage by
administrative means once per 12 hours. The overall likelihood that
an increase in RCS leakage will be detected before it potentially
results in gross failure is maintained with the addition of the
Required Actions.
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, including the edits in brackets above, 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: J. Bradley Fewell, Esquire, Associate
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Richard B. Ennis, Acting.
Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and
3, York and Lancaster Counties, Pennsylvania
Date of amendment request: June 2, 2011.
Description of amendment request: The proposed amendment would
modify Technical Specification Limiting Condition for Operation 3.1.2,
``Reactor Anomalies,'' to allow performance of the surveillance on a
comparison of predicted to actual (or monitored) effective core
reactivity (Keff). The reactivity anomaly verification is
currently determined by a comparison of predicted vs. actual control
rod density.
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 with changes by the NRC staff
noted in brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed Technical Specifications changes do not
[substantively] affect any plant systems, structures, or components
designed for the prevention or mitigation of previously evaluated
accidents. The amendment would only change how the reactivity
anomaly surveillance is performed. Verifying that the core
reactivity is consistent with predicted values ensures that accident
and transient safety analyses remain valid. This amendment changes
the Technical Specification requirements such that, rather than
performing the surveillance by comparing predicted to actual control
rod density, the surveillance is performed by a direct comparison of
keff. Present day on-line core monitoring systems, such
as the one in use at Peach Bottom Atomic Power Station (PBAPS),
Units 2 and 3 are capable of performing the direct measurement of
reactivity.
Therefore, since the reactivity anomaly surveillance will
continue to be performed by a viable method, the proposed amendment
does not involve a significant increase in the probability or
consequence of a previously evaluated accident.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This Technical Specifications amendment request does not
[substantively change] the operation, testing, or maintenance of any
safety-related, or otherwise important to safety systems. All
systems important to safety will continue to be operated and
maintained within their design bases. The proposed changes to the
reactivity anomaly Technical Specifications will only provide a new,
more efficient method of detecting an unexpected change in core
reactivity.
Since all systems continue to be operated within their design
bases, no new failure modes are introduced and the possibility of a
new or different kind of accident is not created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This proposed Technical Specifications amendment proposes to
change the method for performing the reactivity anomaly surveillance
from a comparison of predicted to actual control rod density to a
comparison of predicted to actual keff. The direct
comparison of keff provides a technically superior method
of calculating any differences in the expected core reactivity. The
reactivity anomaly surveillance will continue to be performed at the
same frequency as is currently required by the Technical
Specifications, only the method of performing the surveillance will
be changed. Consequently, core reactivity assumptions made in safety
analyses will continue to be adequately verified.
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, including the changes made by the NRC staff as noted in
brackets, 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: J. Bradley Fewell, Esquire, Associate
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: G. Edward Miller, Acting.
Florida Power Corporation, et al. (FPC), Docket No. 50-302, Crystal
River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida
Date of amendment request: December 20, 2010, as supplemented by
the July 20, 2011 letter.
Description of amendments request: FPC will be constructing and
operating an on-site independent spent fuel storage installation at CR-
3, as a general licensee under the provisions of 10 CFR part 72,
Subpart K to maintain full-core offload capacity in the spent fuel
pools. The spent fuel pools are located in the CR-3 Auxiliary Building
(AB). In support of future dry shielded canister/transfer cask loading
operations, FPC is replacing the existing AB overhead crane with a new
single failure proof crane designed in accordance with American Society
of Mechanical Engineers (ASME) NOG-1-2004, ``Rules
[[Page 55130]]
for Construction of Overhead and Gantry Cranes (Top Running Bridge,
Multiple Girder).'' The licensee requested NRC approval of the
following:
1. An exception to ASME NOG-1-2004 pertaining to the application of
tornado wind and tornado generated missile loading to auxiliary
building overhead crane (FHCR-5) and its support structure.
2. Revisions to the CR-3 Final Safety Analysis Report (FSAR)
Sections 5.1.1.1.h and 9.6.1.5.a.5 to specifically identify the design
parameters for FHCR-5 and its support structure.
3. Deletion of a commitment in FSAR Section 9.6.3.1, ``Spent Fuel
Assembly Removal,'' due to the expansion of spent fuel storage over
that originally credited in the CR-3 Safety Evaluation Report dated
July 5, 1974.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed LAR [license amendment request] does not involve
plant equipment used to operate or shut down the reactor or in the
mitigation of accidents described in Chapter 14 of the FSAR. FHCR-5
will be restricted from movement over fuel stored in either of the
spent fuel pools by administrative controls and designated safe load
paths when moving spent fuel casks, and it will be single failure
proof so a cask load drop accident affecting stored spent fuel is
prevented. The change provides justification for an exception to a
Code requirement pertaining to the design and qualification of the
new single failure proof crane in the AB. The new crane will meet
the design specifications in ASME NOG-1-2004, with the exception of
Section 4134(c). The change also includes a commitment not to
operate the crane if an Approaching or Potential Tropical Storm, an
Approaching or Potential Hurricane, or a Tornado Watch or Warning
has been declared for the site. The revised FSAR description of the
crane will meet the intent of the original description and will
ensure the crane will exceed the design requirements of the original
design. With the replacement of the crane, the occurrence of a cask
load drop accident is not considered credible. As a result, the
proposed change does not increase the probability or consequences of
a load drop accident previously evaluated that could impact stored
fuel and/or pool structural integrity.
Therefore, the proposed change does not involve significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The power generation portion of the plant is unaffected by the
proposed change, which is limited to the design and analysis of a
new overhead crane in the AB. The location and design functions of
the AB overhead crane remain as they are currently described in the
CR-3 FSAR. Overall, the design of the crane is being enhanced to
single failure proof in order to reduce the likelihood of an
uncontrolled lowering of the load due to an unforeseen malfunction
or subcomponent failure. Portions of the design and analysis of the
crane require NRC approval because they deviate from the NRC-
endorsed design code for single failure proof cranes and the CR-3
licensing basis. The new single failure proof crane will be used to
move a loaded or unloaded transfer cask between the cask loading
pit, the decontamination pit, and the transfer trailer in the truck
bay. Any credible event involving the fuel handling evolutions are
bounded by existing FSAR analyses.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does not involve a significant reduction in a margin of
safety.
This proposed LAR involves the replacement of the existing non-
single failure proof AB overhead crane with a new single failure
proof crane. The new crane will meet the design specifications found
in ASME NOG-1-2004, with the exception of Section 4134(c). ASME NOG-
1-2004 has been endorsed by the NRC in Regulatory Issue Summary
(RIS) 2005-25, Supplement 1, ``Clarification of NRC Guidelines for
Control of Heavy Loads,'' as an acceptable means of meeting the
criteria in NUREG-0554, ``Single Failure Proof Cranes for Nuclear
Power Plants.'' The ASME NOG-1-2004 design code has been found by
the NRC to provide adequate protection and safety margin against the
uncontrolled lowering of the lifted load. The occurrence of a cask
load drop accident is considered not credible when the load is
lifted with a single failure proof lifting system meeting the
guidance in NUREG-0612, ``Control of Heavy Loads at Nuclear Power
Plants,'' Section 5.1.6, ``Single Failure Proof Handling Systems.''
As a result, the proposed change has no adverse impact on new fuel,
stored spent fuel, cooling capacity of the pool, or structural
integrity of the pool. Similarly, the margin of safety for the
operation and safe shutdown of the plant will not be affected by the
proposed change.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, NC 27602.
NRC Branch Chief: Douglas A. Broaddus.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, Rockingham County, New Hampshire
Date of amendment request: July 14, 2011.
Description of amendment request: The proposed change would replace
the Technical Specification (TS) required 10-year surveillance
frequency for testing the containment spray nozzles in accordance with
TS surveillance 4.6.2.1.d with an event-based frequency. Specifically,
verification that the spray nozzle is unobstructed would only be
required following activities that could result in nozzle blockage.
Basis for proposed no significant hazards consideration (NSHC)
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of NSHC, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The spray nozzles and the associated containment spray system
(CBS) are designed to perform accident mitigation functions. The
proposed change to reduce the frequency and remove specific details
of surveillance testing that verifies the spray nozzles are
unobstructed does not impact the physical function of plant
structures, systems, or components (SSCs) or the manner in which
SSCs perform their design function. The proposed change neither
adversely affects accident initiators or precursors, nor alters
design assumptions. The proposed change does not alter or prevent
the ability of operable SSCs to perform their intended function to
mitigate the consequences of an initiating event within assumed
acceptance limits. The capability of the CBS system to perform its
accident mitigation functions is not adversely affected by the
proposed change.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change will not impact the accident analysis. The
change does not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed), a significant
change in the method of plant operation, or new operator actions.
The change does not make any physical modifications to the CBS
system, changes to setpoints, or changes to the method of delivering
borated water to the CBS spray nozzles. The proposed change will not
introduce failure modes that could result in a new accident, and the
change does not alter assumptions made in the safety analysis.
[[Page 55131]]
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed change does not
involve a significant change in the method of plant operation, and
no accident analyses will be affected by the proposed changes.
Additionally, the proposed changes will not relax any criteria used
to establish safety limits and will not relax any safety system
settings. The safety analysis acceptance criteria are not affected
by this change. The proposed change will not result in plant
operation in a configuration outside the design basis. The proposed
change does not adversely affect systems that respond to safely shut
down the plant and to maintain the plant in a safe shutdown
condition.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves NSHC. Attorney for licensee: M.S. Ross,
Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-
0420.
NRC Branch Chief: Harold K. Chernoff.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2 (VEGP), Burke
County, Georgia
Date of amendment request: July 26, 2011.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TSs). Specifically, the proposed
change would revise the minimum indicated nitrogen cover pressure
specified for the accumulators in TS Surveillance Requirement (SR)
3.5.1.3 from 617 psig (pounds per square inch, gauge) to 626 psig. The
proposed change is necessary to account for the uncertainty associated
with the accumulator pressure indication instrumentation. Currently, in
accordance with NRC Administrative Letter 98-10, ``Dispositioning of
Technical Specifications that Are Insufficient to Assure Plant
Safety,'' VEGP is administratively controlling the minimum indicated
accumulator pressure to greater than or equal to 626 psig. In addition,
an editorial error in the text of TS SR 3.6.2.1 would also be
corrected.
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 the minimum indicated nitrogen
cover pressure specified for the SI [safety injection] accumulators
in SR 3.5.1.3 from 617 psig to 626 psig. In addition, the proposed
change includes an administrative change to correct an editorial
error in the text of TS SR 3.6.2.1.
The SI accumulators are not a precursor to any accident
previously evaluated. The SI accumulators are used to mitigate the
consequences of accidents previously evaluated. The proposed change
to the indicated minimum SI accumulator nitrogen cover pressure
provides assurance that the requirements of the TS continue to bound
the acceptance limits of the SI accumulators with respect to the
assumptions in the LOCA [loss-of-coolant accident] analyses.
Thus, the proposed change does not affect the probability or the
consequences of any accident previously evaluated. The proposed
change to correct an editorial error in the text of SR 3.6.2.1 has
no impact on the probability or consequences of any accident
previously evaluated.
Therefore, it is concluded that the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No
The proposed change revises the minimum indicated nitrogen cover
pressure specified for the SI accumulators in SR 3.5.1.3 from 617
psig to 626 psig. In addition, the proposed change includes an
administrative change to correct an editorial error in the text of
TS SR 3.6.2.1.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
proposed change to the requirements of the TS assure that the
acceptance limits of the SI accumulators with respect to the
assumptions in the LOCA analyses continue to be met, and correct an
editorial error in the text of an SR. Thus, the proposed change does
not adversely affect the design function or operation of any
structures, systems, and components important to safety.
Therefore, it is concluded that 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 change revises the minimum indicated nitrogen cover
pressure specified for the SI accumulators in SR 3.5.1.3 from 617
psig to 626 psig. In addition, the proposed change includes an
administrative change to correct an editorial error in the text of
TS SR 3.6.2.1.
The proposed change to the indicated SI accumulator nitrogen
cover pressure provides assurance that the requirements of the TS
continue to bound the acceptance limits of the SI accumulators with
respect to the assumptions in the LOCA analyses. Thus the proposed
change to the SI accumulator minimum nitrogen cover pressure assures
the existing margin of safety is maintained. The proposed change to
correct an editorial error in the text of SR 3.6.2.1 has no impact
on the margin of safety.
Therefore, it is concluded that 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. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216. NRC Branch Chief: Gloria Kulesa.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee
Date of amendment request: July 15, 2011 (TS-SQN-2011-01).
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) requirements for steam
generator (SG) tube inspections to reflect the replacement steam
generators (RSGs) to be installed during Sequoyah Nuclear Plant (SQN),
Unit 2, refueling outage 18 presently scheduled for the fall of 2012.
Previous changes to the SQN, Unit 2, TSs to reflect the Technical
Specification Task Force (TSTF) Standard Technical Specification
Traveler, TSTF-449, ``Steam Generator Tube Integrity,'' Revision 4,
were approved by Nuclear Regulatory Commission (NRC) on May 22, 2007.
The changes proposed in this amendment reflect the inspection
requirements of TSTF-449, Revision 4. The RSG tubes will be made of
Alloy 690 thermally treated (TT) material, and the existing SGs have
Alloy 600 tubes. The revisions to TSs are required because the
inspection frequency for Alloy 690 TT tube material, as defined
[[Page 55132]]
in TSTF-449, differs from the inspection frequency for Alloy 600, and
the tube repair processes and products in the existing TSs are not
applicable to the RSGs.
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 for RSGs continues to implement the current
SG Program that includes performance criteria which provide
reasonable assurance that the RSG tubing will retain integrity over
the full range of operating conditions (including startup, operation
in the power range, hot standby, cooldown, and all anticipated
transients included in the design specifications). This change
removes repair criteria from the SG Program that were approved by
previous License Amendments for the existing SGs which are not
applicable to the RSGs. It removes references to use of repairs and
reporting of repair results in other TS sections. This change
removes inspection requirements that are designated for specific
damage conditions in the existing SGs. The change also revises the
inspection interval for 100 percent inspections of SG tubes and the
maximum interval for inspection of a single SG consistent with
Technical Specification Task Force (TSTF) Standard Technical
Specification Traveler, TSTF-449, ``Steam Generator Tube
Integrity,'' Revision 4 for the Alloy 690 tube material in the RSGs.
The revised inspection requirements are based on properties and
experience with the improved Alloy 690 tube material. The revised
inspection requirements will result in the same outcome that SG tube
integrity will continue to be maintained.
This change continues to implement SG performance criteria for
tube structural integrity, accident induced leakage, and operational
leakage for the RSGs. Meeting the performance criteria provides
reasonable assurance that the RSG tubing will remain capable of
fulfilling its specific safety function of maintaining reactor
coolant pressure boundary integrity throughout each operating cycle
and in the unlikely event of a design basis accident (DBA). The
performance criteria are only a part of the SG Program required by
the existing TS. The program, defined by NEI [Nuclear Energy
Institute] 97-06, ``Steam Generator Program Guidelines,'' includes a
framework that incorporates a balance of prevention, inspection,
evaluation, repair, and leakage monitoring. These features will
continue to be implemented as they are currently approved. The
proposed changes do not, therefore, significantly increase the
probability of an accident previously evaluated.
The consequences of DBAs are, in part, functions of the Dose
Equivalent 1-131 in the primary coolant and the primary to secondary
leakage rates resulting from an accident. Therefore, limits are
included in the TS for Operational Leakage and for Dose Equivalent
1-131 in the primary coolant to ensure the plant is operated within
its analyzed condition. The analysis of the limiting DBA assumes
that the primary to secondary leak rate, after the accident, is 1
gallon per minute with no more than 150 gallons per day in any one
SG, and that the reactor coolant activity levels of Dose Equivalent
1-131 are at the TS values before the accident. The proposed change
to the SG inspection program does not affect the design of the SGs,
their method of operation, operational leakage limits, or primary
coolant chemistry controls. The proposed change does not adversely
impact any other previously evaluated DBA. In addition, the proposed
changes do not affect the consequences of a main steam line break,
rod ejection, a reactor coolant pump locked rotor event, or other
previously evaluated accident.
Therefore, the proposed change does not affect the consequences
of a[n] SG tube rupture accident and the probability of such an
accident is unchanged.
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 license amendment does not affect the method of
operation of the SGs, or the primary or secondary coolant chemistry
controls. In addition, the proposed amendment does not impact any
other plant system or component. The change modifies existing SG
inspection requirements based on the RSG design and the properties
and experience associated with their improved materials. The revised
inspection requirements will result in the same outcome that SG tube
integrity will continue to be maintained.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of a[n] SG is
maintained by ensuring the integrity of its tubes. SG tube integrity
is a function of the design, environment, and the physical condition
of the tube. The proposed change to the SG inspection program does
not affect tube design or operating environment. The existing SG
Program is maintained in this change. The repair criteria that are
being removed are specific to the existing SGs and are not
applicable to the RSGs. If tube defects are detected that exceed
limits in the RSGs, then the tube will be removed from service. The
effective tube plugging percentage will continue to be tracked for
all plugging in each SG in accordance with TS Section 6.9.1.16.1 to
ensure the heat transfer function of the SGs is not adversely
affected. For the above reasons, the margin of safety is not changed
and overall plant safety will be enhanced by the proposed change to
the TS.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, Tennessee 37902.
NRC Branch Chief: Douglas A. Broaddus.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Exelon Generation Company, LLC, and PSEG Nuclear, LLC, Docket No. 50-
278, Peach Bottom Atomic Power Station (PBAPS), Unit 3, York and
Lancaster Counties, Pennsylvania
Date of application for amendments: June 28, 2011.
Brief description of amendment request: The proposed amendment
would modify the PBAPS, Unit 3, Technical Specification Section 2.1.1
to revise Safety Limit Minimum Critical Power Ratio values.
Date of publication of individual notice in Federal Register:
August 22, 2011 (76 FR 52357).
Expiration date of individual notice: September 21, 2011 (public
comments) and October 21, 2011 (hearing requests).
[[Page 55133]]
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the NRC's Public Document Room (PDR), located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through the Agencywide Documents Access and
Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit 1 and 2, Calvert County,
Maryland
Date of application for amendments: March 22, 2011.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to define a new time limit for restoring
inoperable reactor coolant system (RCS) leakage detection
instrumentation to operable status. The proposed TS changes are
consistent with TS Task Force (TSTF)-513, ``Revise PWR [pressurized-
water reactor] Operability Requirements and Actions for RCS Leakage
Instrumentation.''
Date of issuance: August 24, 2011.
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment Nos.: 299 and 276.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: April 19, 2011 (76 FR
21920).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated August 24, 2011.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, LLC,
Calvert Cliffs Nuclear Power Plant, Unit 1 and 2 (CCNPP),
Docket Nos. 50-317, 50-318,
Calvert County, Maryland,
Nine Mile Point Nuclear Station, LLC,
Nine Mile Point Nuclear Station, Unit 1 and 2 (NMPNS),
Docket Nos. 50-220, 50-410,
Oswego County, New York, and
R. E. Ginna Nuclear Power Plant, LLC,
R. E. Ginna Nuclear Power Plant (Ginna),
Docket No. 50-244, Wayne County, New York
Date of amendment request: July 16, 2010, as supplemented by
letters dated April 4, and July 1, 2011.
Brief description of amendments: The amendments to the Renewed
Facility Operating Licenses (FOLs) includes: (1) The U.S. Nuclear
Regulatory Commission (NRC)-approved Cyber Security Plan (CSP) for
CCNPP, NMPNS, and Ginna, (2) the CSP implementation schedule, and (3)
the license condition added to the existing physical protection license
condition for CCNPP, NMPNS, and Ginna, requiring the licensee to fully
implement and maintain in effect all provisions of the NRC-approved CSP
for CCNPP, NMPNS, and Ginna, as required by Title 10 of the Code of
Federal Regulations (10 CFR) 73.54 ``Protection of digital computer and
communication systems and networks.'' A Federal Register notice dated
March 27, 2009, issued the final rule that amended 10 CFR 73.54. The
regulations in 10 CFR 73.54, establish the requirements for a CSP. This
regulation specifically requires each licensee currently licensed to
operate a nuclear power plant under Part 50 of this chapter to submit a
CSP that satisfies the requirements of the Rule. Each submittal must
include a proposed implementation schedule and implementation of the
licensee's CSP must be consistent with the approved schedule. The
background for this application is addressed by the NRC Notice of
Availability, Federal Register Notice, Final Rule 10 CFR part 73, Power
Reactor Security Requirements, published on March 27, 2009, 74 FR
13926.
Date of issuance: August 19, 2011.
Effective date: These license amendments are effective as of the
date of its issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee on July 16, 2010, as supplemented by letters dated April 4,
and July 1, 2011, and approved by the NRC staff with this license
amendment. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment Nos.: 298, 275 (CCNPP1 & CCNPP2), 209, 137 (NMPNS1 &
NMPNS2), and 113 (Ginna),.
Renewed Facility Operating License Nos. DPR-53 and DPR-69 (CCNPP1 &
CCNPP2), DPR-63, NPF-69, (NMP1 & NMP2), and DPR-18 (Ginna),: Amendments
revised the Licenses.
Date of initial notice in Federal Register: October 12, 2010 (75 FR
62594). The supplement dated April 4, and July 1, 2011, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated August 19, 2011.
No significant hazards consideration comments received: Yes.
The State of Maryland had no comments. However, the New York State
provided comments. The Safety
[[Page 55134]]
Evaluation dated August 19, 2011, provides the discussion of the
comments received from the New York State.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York
Date of application for amendment: July 15, 2010, as supplemented
by letters dated February 15 and April 4, 2011.
Brief description of amendment: The application for the proposed
amendment to the Renewed Facility Operating License (FOL) includes: (1)
The proposed JAFNPP Cyber Security Plan, (2) an implementation
schedule, and (3) a proposed sentence to be added to the existing
renewed FOL Physical Protection license condition for JAFNPP requiring
Entergy to fully implement and maintain in effect all provisions of the
Commission-approved JAFNPP Cyber Security Plan (CSP) as required by 10
CFR 73.54, ``Protection of digital computer and communication systems
and networks.'' A Federal Register notice dated March 27, 2009, issued
the final rule that amended 10 CFR part 73. The regulations in 10 CFR
73.54, establish the requirements for a cyber security program. This
regulation specifically requires each licensee currently licensed to
operate a nuclear power plant under Part 50 of this chapter to submit a
CSP that satisfies the requirements of the Rule. Each submittal must
include a proposed implementation schedule and implementation of the
licensee's Cyber Security Program must be consistent with the approved
schedule. The background for this application is addressed by the NRC
Notice of Availability, Federal Register Notice, Final Rule 10 CFR part
73, Power Reactor Security Requirements, published on March 27, 2009
(74 FR 13926).
Date of issuance: August 19, 2011.
Effective date: This license amendment is effective as of the date
of its issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee on July 15, 2010, as supplemented by letters dated February 15
and April 4, 2011, and approved by the NRC staff with this license
amendment. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment No.: 300.
Renewed Facility Operating License No. DPR-59: The amendment
revised the License
Date of initial notice in Federal Register: August 20, 2010 (75 FR
51492). The supplements dated February 15, and April 4, 2011, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 19, 2011.
No significant hazards consideration comments received: Yes.
The Safety Evaluation dated August 19, 2011, provides the
discussion of the comments received from the New York State.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Date of application for amendment: September 23, 2010 as
supplemented by letter dated. April 22, 2011.
Brief description of amendment: The amendment revised Technical
Specification (TS) limiting condition for operation 3.7.6, ``Main
Turbine Bypass System (MTBS),'' to control the reactor operational
limits, as specified in the Clinton Power Station Core Operating Limits
Report to compensate for the inoperability of the MTBS.
Date of issuance: August 17, 2011.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 195.
Facility Operating License No. NPF-62: The amendment revised the
TSs and license.
Date of initial notice in Federal Register: February 1, 2011 (76 FR
5618). The April 22, 2011 supplement contained clarifying information
and did not change the NRC staff=s initial proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 17, 2011.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: September 22, 2010, supplemented
by letter dated April 7, 2011.
Brief description of amendment: The changes relocate the list of
pumps, fans, and valves in Technical Specification (TS) 4.5.1.1b,
Sequence and Power Transfer Test, to the TMI-1 Updated Final Safety
Analysis Report. In place of the TS equipment listing there will be a
more general reference to the permanently-connected and automatically-
connected emergency loads which are tested through the load sequencer.
In addition, TS 4.5.1.2b, TS 4.5.2.2a, and TS 4.5.2.2b refer to this
test and are revised to reflect the change to TS 4.5.1.1b.
Date of issuance: August 22, 2011.
Effective date: Immediately, and shall be implemented within 30
days.
Amendment No.: 276.
Renewed Facility Operating License No. DPR-50. Amendment revised
the license and the technical specifications.
Date of initial notice in Federal Register: November 30, 2010 (75
FR 74095). The supplement dated April 7, 2011, modified the application
such that the Federal Register notice was re-issued on May 3, 2011 (76
FR 24928). The revised notice did not change the NRC staff's proposed
no significant hazards determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 22, 2011.
No significant hazards consideration comments received: No.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, Rockingham County, New Hampshire
Date of amendment request: December 29, 2010.
Description of amendment request: The proposed change deletes the
Seabrook Technical Specification (TS) 3.4.10, ``Structural Integrity,''
while relocating the requirements of Surveillance Requirement 4.4.10 to
TS 6.7.6.m.
Date of issuance: August 22, 2011.
Effective date: As of its date of issuance and shall be implemented
within 60 days.
Amendment No.: 126.
Facility Operating License No. NPF-86: The amendment revised the TS
and the License.
Date of initial notice in Federal Register: May 31, 2011 (76 FR
31375).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 22, 2011.
No significant hazards consideration comments received: No.
[[Page 55135]]
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, Rockingham County, New Hampshire
Date of amendment request: July 26, 2010, as supplemented by
letters dated September 28, 2010, March 31, June 23, and August 4,
2011.
Description of amendment request: This amendment approves the
NextEra Seabrook LLC, cyber security plan (CSP) for Seabrook Station,
Unit 1. Additionally, the amendment adds a license condition requiring
that the licensee fully implement and maintain in effect all provisions
of the approved plan.
Date of issuance: August 23, 2011.
Effective date: The license amendment is effective as of its date
of issuance. The implementation of the CSP, including key intermediate
milestone dates and the full implementation date, shall be in
accordance with the implementation schedule submitted by the licensee
by letter dated March 31, 2011, and approved by the NRC staff with this
license amendment. All subsequent changes to the NRC-approved CSP
implementation schedule will require prior NRC approval pursuant to 10
CFR 50.90.
Amendment No.: 127.
Facility Operating License No. NPF-86: The amendment revised the
License.
Date of initial notice in Federal Register: May 10, 2011 (76 FR
27097).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 23, 2011.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of application for amendments: March 18, 2011, as supplemented
by letters dated May 4 and June 2, 2011.
Brief description of amendments: The amendments modified the
Security Plan, Training and Qualification Plan, Safeguards Contingency
Plan, and Independent Spent Fuel Storage Installation Security Program.
Date of issuance: August 16, 2011.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 203, 190.
Facility Operating License Nos. DPR-42 and DPR-60: The amendments
revised the Operating Licenses for both units.
Date of initial notice in Federal Register: May 10, 2011 (76 FR
27098). The supplemental letters contained clarifying information and
did not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 16, 2011.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1,
Fairfield County, South Carolina
Date of application for amendment: August 5, as supplemented
September 27, and November 30, 2010 and March 28, 2011.
Brief description of amendment: The amendments revised Paragraph
2.E of the renewed facility operating license to provide a license
condition to require the licensee to fully implement and maintain in
effect all provisions of the NRC-approved Cyber Security Plan and
associated implementation schedule.
Date of issuance: August 24, 2011.
Effective date: This license amendment is effective as of its date
of issuance. The implementation of the CSP, including the key
intermediate milestone dates and the full implementation date, shall be
in accordance with the implementation schedule submitted by the
licensee on March 28, 2011, and approved by the Nuclear Regulatory
Commission (NRC) staff with this license amendment. All subsequent
changes to the NRC-approved CSP implementation schedule will require
prior NRC approval pursuant to 10 CFR 50.90.
Amendment No.: 184 .
Renewed Facility Operating License No. NPF-12: Amendment revised
the license.
Date of initial notice in Federal Register: April 12, 2011 (76 FR
20380). The September 27, 2010, and March 28, 2011, supplements
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change NRC staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 24, 2011.
No significant hazards consideration comments received: No
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: August 12, 2010, as supplemented
by letters dated September 27, November 29, and December 30, 2010, and
April 1, June 14, and June 29, 2011.
Brief description of amendment: The amendment approved the Callaway
Plant, Unit 1, Cyber Security Plan and associated implementation
schedule, and revised Paragraph 2.E of Facility Operating License No.
NPF-30 to provide a license condition to require the licensee to fully
implement and maintain in effect all provisions of the NRC-approved
Cyber Security Plan. The proposed change is generally consistent with
Nuclear Energy Institute (NEI) 08-09, Revision 6, ``Cyber Security Plan
for Nuclear Power Reactors.''
Date of issuance: August 17, 2011.
Effective date: This license amendment is effective as of the date
of its issuance. The implementation of the cyber security plan (CSP),
including the key intermediate milestone dates and the full
implementation date, shall be in accordance with the revised
implementation schedule submitted by the licensee on June 29, 2011, and
approved by the NRC staff with this license amendment. All subsequent
changes to the NRC-approved CSP implementation schedule will require
prior NRC approval pursuant to 10 CFR 50.90.
Amendment No.: 203.
Facility Operating License No. NPF-30: The amendment revised the
Operating License.
Date of initial notice in Federal Register: November 9, 2010 (75 FR
68837). The supplemental letters dated September 27, November 29, and
December 30, 2010, and April 1, June 14, and June 29, 2011, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 17, 2011.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 25th day of August 2011.
[[Page 55136]]
For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2011-22541 Filed 9-2-11; 8:45 am]
BILLING CODE 7590-01-P