[Federal Register Volume 78, Number 42 (Monday, March 4, 2013)]
[Notices]
[Pages 14126-14141]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-04885]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0045]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined 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 or
combined license, as applicable, 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 February 7, 2013, to February 20, 2013. The
last biweekly notice was published on February 19, 2013 (78 FR 11688).
ADDRESSES: You may access information and comment submissions related
to this document, which the NRC possesses and is publicly available, by
searching on http://www.regulations.gov under Docket ID . You may submit comments by the following methods:
Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID . Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected].
Mail comments to: Cindy Bladey, 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.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID when contacting the NRC
about the availability of information regarding this document. You may
access information related to this document, which the NRC possesses
and is publicly available, by the following methods:
Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID .
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly available documents online in the NRC
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in
ADAMS by performing a search on the document date and docket number.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID in the subject line of
your comment submission, in order to ensure that the NRC is able to
make your comment submission available to the public in this docket.
The NRC cautions you not to include identifying or contact
information in comment submissions that you do not want to be publicly
disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS, and the NRC does not edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information in their comment submissions
that they do not want to be publicly disclosed. Your request should
state that the NRC will not edit comment submissions to remove such
information before making the comment submissions available to the
public or entering the comment submissions into ADAMS.
[[Page 14127]]
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Title 10 of the Code of Federal
Regulations (10 CFR), Section 50.92, this means that operation of the
facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
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 or
combined 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 email at [email protected], or by
telephone at 301-415-1677, to request (1) a digital information (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign
[[Page 14128]]
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 the 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 the 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 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 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 email notice confirming receipt of the document. The
E-Filing system also distributes an email 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 email 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 the
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.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. 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. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the following three
factors in 10 CFR 2.309(c)(1): (i) The information upon which the
filing is based was not previously available; (ii) the information upon
which the filing is based is materially different from information
previously available; and (iii) the filing has been submitted in a
timely fashion based on the availability of the subsequent information.
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 email to [email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: December 12, 2012.
[[Page 14129]]
Description of amendment request: The amendments would change the
Technical Specifications (TSs) by replacing the current limits on
primary coolant gross specific activity with limits on primary coolant
noble gas activity. The noble gas activity would be based on DOSE
EQUIVALENT XE-133 and would take into account only the noble gas
activity in the primary coolant. The changes are consistent with NRC-
approved Industry/Technical Specification Task Force (TSTF) Standard
Technical Specification Change Traveler, TSTF-490, Revision 0,
``Deletion of E-Bar Definition and Revision to RCS [Reactor Coolant
System] Specific Activity Technical Specifications,'' with deviations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The license concluded that the no significant hazards
consideration determination published in the Federal Register on March
19, 2007 (72 FR 12838), is applicable, and is presented below:
1. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated
Response: Reactor coolant specific activity is not an initiator
for any accident previously evaluated. The Completion Time when
primary coolant gross activity is not within limit is not an
initiator for any accident previously evaluated. The current
variable limit on primary coolant iodine concentration is not an
initiator to any accident previously evaluated. As a result, the
proposed change does not significantly increase the probability of
an accident. The proposed change will limit primary coolant noble
gases to concentrations consistent with the accident analyses. The
proposed change to the Completion Time has no impact on the
consequences of any design basis accident since the consequences of
an accident during the extended Completion Time are the same as the
consequences of an accident during the Completion Time. As a result,
the consequences of any accident previously evaluated are not
significantly increased.
2. The Proposed Change Does Not Create the Possibility of a New
or Different Kind of Accident from any Accident Previously Evaluated
Response: The proposed change in specific activity limits does
not alter any physical part of the plant nor does it affect any
plant operating parameter. The change does not create the potential
for a new or different kind of accident from any previously
calculated.
3. The Proposed Change Does Not Involve a Significant Reduction
in the Margin of Safety
Response: The proposed change revises the limits on noble gase
[sic] radioactivity in the primary coolant. The proposed change is
consistent with the assumptions in the safety analyses and will
ensure the monitored values protect the initial assumptions in the
safety analyses.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three 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: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: Michael T. Markley.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendment request: December 26, 2012.
Description of amendment request: The amendments would adopt
Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision
2, ``DC Electrical Rewrite--Update to TSTF-360,'' with one variation.
The amendments would revise the TS requirements related to direct
current (DC) electrical systems in TS Limiting Condition for Operation
(LCO) 3.8.4, ``DC Sources--Operating,'' LCO 3.8.5, ``DC Sources--
Shutdown,'' and LCO 3.8.6, ``Battery Parameters.'' In addition, new TS
5.5.19, ``Battery Monitoring and Maintenance Program,'' is being
proposed for Section 5.5, ``Administrative Controls--Programs and
Manuals.''
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 any accident previously
evaluated?
Response: No.
The proposed changes restructure the Technical Specifications
(TS) for the direct current (DC) electrical power system and are
consistent with TSTF-500, Revision 2. The proposed changes modify TS
Actions relating to battery and battery charger inoperability. The
DC electrical power system, including associated battery chargers,
is not an initiator of any accident sequence analyzed in the Updated
Final Safety Analysis Report (UFSAR). Rather, the DC electrical
power system supports equipment used to mitigate accidents. The
proposed changes to restructure TS and change surveillances for
batteries and chargers to incorporate the updates included in TSTF-
500, Revision 2, will maintain the same level of equipment
performance required for mitigating accidents assumed in the UFSAR.
Operation in accordance with the proposed TS would ensure that the
DC electrical power system is capable of performing its specified
safety function as described in the UFSAR. Therefore, the mitigating
functions supported by the DC electrical power system will continue
to provide the protection assumed by the analysis. The relocation of
preventive maintenance surveillances, and certain operating limits
and actions, to a licensee-controlled Battery Monitoring and
Maintenance Program will not challenge the ability of the DC
electrical power system to perform its design function. Appropriate
monitoring and maintenance that are consistent with industry
standards will continue to be performed. In addition, the DC
electrical power system is within the scope of 10 CFR 50.65,
Requirements for monitoring the effectiveness of maintenance at
nuclear power plants, which will ensure the control of maintenance
activities associated with the DC electrical power system.
The integrity of fission product barriers, plant configuration,
and operating procedures as described in the UFSAR will not be
affected by the proposed changes. Therefore, the consequences of
previously analyzed accidents will not increase by implementing
these changes. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed changes involve restructuring the TS for the DC
electrical power system. The DC electrical power system, including
associated battery chargers, is not an initiator to any accident
sequence analyzed in the UFSAR. Rather, the DC electrical power
system supports equipment used to mitigate accidents. The proposed
changes to restructure the TS and change surveillances for batteries
and chargers to incorporate the updates included in TSTF-500,
Revision 2, will maintain the same level of equipment performance
required for mitigating accidents assumed in the UFSAR.
Administrative and mechanical controls are in place to ensure the
design and operation of the DC systems continues to meet the plant
design basis described in the UFSAR. Therefore, operation of the
facility in accordance with this proposed change will 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
the margin of safety?
[[Page 14130]]
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The equipment margins will be maintained in
accordance with the plant-specific design bases as a result of the
proposed changes. The proposed changes will not adversely affect
operation of plant equipment. These changes will not result in a
change to the setpoints at which protective actions are initiated.
Sufficient DC capacity to support operation of mitigation equipment
is ensured. The changes associated with the new Battery Maintenance
and Monitoring Program will ensure that the station batteries are
maintained in a highly reliable manner. The equipment fed by the DC
electrical sources will continue to provide adequate power to
safety-related loads in accordance with analysis assumptions.
TS changes made in accordance with TSTF-500, Revision 2,
maintain the same level of equipment performance stated in the UFSAR
and the current TSs. Therefore, the proposed changes do not involve
a significant reduction of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three 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: Michael G. Green, Senior Regulatory Counsel,
Pinnacle West Capital Corporation, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: Michael T. Markley.
Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: October 2, 2012, as supplemented by
letter dated November 26, 2012.
Description of amendments request: The amendments would revise
Technical Specification (TS) 3.8.3 ``Diesel Fuel Oil'' by relocating
the current stored diesel fuel oil numerical volume requirements from
the TS to the TS Bases and TS 3.8.1 ``AC Sources-Operating'' by
relocating the specific numerical value for the day tank fuel oil
volume from the TS to the TS Bases. The changes would be consistent
with Nuclear Regulatory Commission (NRC)-approved Industry Technical
Specification Task Force Standard Technical Specification Change
Traveler, TSTF-501-A, Revision 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
No.
The proposed change relocates the volume of diesel fuel oil
required to support 7-day operation of an onsite diesel generator,
and the volume equivalent to a 6-day supply, to licensee control.
The specific volume of fuel oil equivalent to a 7- and 6-day supply
is calculated using the limiting energy content of the fuel, the
required diesel generator output and the corresponding fuel oil
consumption rate. Because the requirement to maintain a 7-day supply
of diesel fuel oil is not changed and is consistent with the
assumptions in the accident analysis, and the actions taken with the
volume of fuel oil is less than a 6-day supply have not changed,
neither the probability nor the consequences of any accident
previously evaluated will be affected.
The proposed change also relocates the volume of diesel fuel oil
required to support one hour of diesel generator operation at full
load in the day tank. The specific volume and time is not changed
and is consistent with the existing plant design basis to support a
diesel generator under accident load conditions.
Therefore, operation of the facility in accordance with the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different type of accident
from any accident previously evaluated; or
No.
The 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. The change
does not alter assumptions made in the safety analysis but ensures
that the diesel generator operates as assumed in the accident
analysis. The proposed change is consistent with the safety analysis
assumptions.
The proposed change also relocates the volume of diesel fuel oil
required to support one hour of diesel generator operation at full
load in the day tank. The change does not alter assumptions made in
the safety analysis but ensures that the diesel generator operates
as assumed in the accident analysis. The proposed change is
consistent with the safety analysis assumptions. Therefore, the
proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
No.
The proposed change relocates the volume of diesel fuel oil
required to support 7-day operation of an onsite diesel generator,
and the volume equivalent to a 6-day supply, and one hour day tank
supply to licensee control. As the basis for the existing limits on
diesel fuel oil are not changed, no change is made to the accident
analysis assumptions and no margin of safety is reduced as part of
this change.
The proposed change also relocates the volume of diesel fuel oil
required to support one hour of diesel generator operation at full
load in the day tank. As the basis for the existing limits on diesel
fuel oil are not changed, no change is made to the accident analysis
assumptions and no margin of safety is reduced as part of this
change.
Therefore, the proposed amendment would not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Steven L. Miller, General Counsel,
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite
200c, Baltimore, MD 21202.
NRC Branch Chief: George Wilson.
Calvert Cliffs Nuclear Power Plant, LLC, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: October 16, 2012.
Description of amendments request: The amendments would revise
Surveillance Requirements (SRs) 3.8.1.8, 3.8.1.11, and 3.8.2.1 and add
SR 3.8.1.17 of Technical Specification (TS) 3.8.1 ``AC Sources--
Operating.''
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 proposes to add or modify certain [TS
SRs] for the diesel generators. This proposed amendment will provide
additional assurance that the AC Sources relied upon to ensure the
availability of necessary power to the Engineered Safety Features
systems are capable of performing their specified safety function if
needed. The diesel generators and their associated emergency loads
are accident mitigating features, not accident initiators. This
proposed amendment does not change the design function of the diesel
generators or any of their required loads, and does not change the
way the systems and plant are operated or maintained. This proposed
amendment does not impact any plant systems that are accident
initiators and does not adversely impact any accident mitigating
systems.
[[Page 14131]]
The proposed amendment does not affect the operability
requirements for the diesel generators, as verification of such
operability will continue to be performed as required. Continued
verification of operability supports the capability of the diesel
generators to perform their required design functions of providing
emergency power to the Engineered Safety Features systems,
consistent with the plant safety analyses as described in the
Updated Final Safety Analysis Report (UFSAR).
Adding or modifying [TS SRs] for the diesel generators will not
significantly increase the probability of an accident previously
evaluated because the diesel generators and their emergency loads
are accident mitigation features, not accident initiators. Adding or
modifying [TS SRs] for the diesel generators will not change any of
the dose analyses associated with the UFSAR Chapter 14 accidents
because accident mitigation functions and requirements remain
unchanged.
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 proposes to add or modify certain [TSs
SRs] for the diesel generators. This proposed amendment does not
change the design function of the diesel generators or any required
loads, and does not change the way the systems and plant are
operated or maintained. This proposed amendment does not impact any
plant systems that are accident initiators and does not adversely
impact any accident mitigating systems. Performance of these
surveillances tests will provide additional assurance that the AC
Sources relied upon to ensure the availability of necessary power to
the Engineered Safety Features systems are capable of performing
their specified safety function if needed.
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 the margin of safety?
Response: No.
This amendment request proposes to add or modify certain [TS
SRs] for the diesel generators. This proposed amendment will provide
additional assurance that the AC Sources relied upon to ensure the
availability of necessary power to the Engineered Safety Features
systems are capable of performing their specified safety function if
needed. Margin of safety is related to the ability of the fission
product barriers (fuel cladding, reactor coolant system, and primary
containment) to perform their design functions during and following
postulated accidents. This proposed amendment does not involve or
affect fuel cladding, the reactor coolant system, or the primary
containment. Performance of these surveillances tests will provide
continued assurance that the AC Sources relied upon to ensure the
availability of necessary power to the Engineered Safety Features
systems are capable of performing their specified safety function if
needed.
Therefore, the proposed amendment 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: Steven L. Miller, General Counsel,
Constellation Energy Nuclear Group, LLC, 100 Constellation Way, Suite
200c, Baltimore, MD 21202.
NRC Branch Chief: George Wilson.
Detroit Edison, Docket No. 50-341, Fermi 2, Monroe County, Michigan
Date of amendment request: December 21, 2012.
Description of amendment request: The proposed amendment would
revise the Fermi 2 operating license to change its name on the license
to ``DTE Electric Company.'' This name change is purely administrative
in nature. Detroit Edison is a wholly owned subsidiary of DTE Energy
Company, and this name change is part of a set of name changes of DTE
Energy subsidiaries to conform their names to the ``DTE'' brand name.
No other changes are contained within this request. This request does
not involve a transfer of control over or of an interest in the license
for Fermi 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment changes the name of the owner licensee.
The proposed amendment is purely administrative in nature. The
functions, powers, resources and management of the owner licensee
will not change. Detroit Edison, which will be renamed DTE Electric
Company, will remain the licensee of the facility. The proposed
changes do not adversely affect accident initiators or precursors,
and do not alter the design assumptions, conditions, or
configuration of the plant or the manner in which the plant is
operated and maintained. The ability of structures, systems, and
components to perform their intended safety functions is not altered
or prevented by the proposed changes, and the assumptions used in
determining the radiological consequences of previously evaluated
accidents are not affected.
Therefore, the proposed changes do 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 accident previously
evaluated.
The proposed amendment is purely administrative in nature. The
functions of the owner licensee will not change. These changes do
not involve any physical alteration of the plant (i.e., no new or
different type of equipment will be installed), and installed
equipment is not being operated in a new or different manner. Thus,
no new failure modes are introduced.
Therefore, the proposed changes do 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.
The proposed amendment is a name change to reflect the new name
of the owner licensee. The proposed amendment is purely
administrative in nature. The functions of the owner licensee will
not change. Detroit Edison, which will be renamed DTE Electric
Company, will remain the licensee of the facility, and its functions
will not change. The proposed changes do not alter the manner in
which safety limits, limiting safety system settings, or limiting
conditions for operation are determined. There are no changes to
setpoints at which protective actions are initiated, and the
operability requirements for equipment assumed to operate for
accident mitigation are not affected.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bruce R. Masters, DTE Energy, General
Council--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
NRC Branch Chief: Robert D. Carlson.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County,
Texas
Date of amendment request: December 19, 2012.
Brief description of amendments: The amendments would revise
Technical Specification (TS) 3.8.1, ``AC [Alternating Current]
Sources--Operating,'' to revise the Completion Time (CT) for Required
Action A.3, ``Restore required offsite circuit to OPERABLE status,'' on
one-time basis from 72 hours to 14 days for Comanche
[[Page 14132]]
Peak Nuclear Power Plant (CPNPP), Units 1 and 2. The CT extension from
72 hours to 14 days will be used twice while completing the plant
modification to install alternate startup transformer (ST) XST1A and
will expire on March 31, 2014. After completion of this modification,
if ST XST1 should require maintenance or if failure occurs, the
alternate ST XST1A can be aligned to the Class 1E buses well within the
current CT of 72 hours. Installation of alternate ST will result in
improved plant design and will improve the long-term reliability of the
138 kiloVolt (kV) offsite circuit ST.
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 will revise the CT for the loss of one
offsite source from 72 hours to 14 days to allow two, one-time, 14-
day CTs. The proposed two, one-time extensions of the CT for the
loss of one offsite power circuit does not significantly increase
the probability of an accident previously evaluated. The TS will
continue to require equipment that will power safety related
equipment necessary to perform any required safety function. The
two, one-time extensions of the CT to 14 days does not affect the
design of the STs, the interface of the STs with other plant
systems, the operating characteristic of the STs, or the reliability
of the STs.
The consequence of a LOOP [loss-of-offsite power] event has been
evaluated in the CPNPP Final Safety Analysis Report (Reference 8.1
[of application dated December 19, 2012]) and the Station Blackout
evaluation. Increasing the CT for one offsite power source twice on
a one-time basis from 72 hours to 14 days does not increase the
consequences of a LOOP event nor change the evaluation of LOOP
events.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not result in a change in the manner in
which the electrical distribution subsystems provide plant
protection. The proposed change will only affect the time allowed to
restore the operability of the offsite power source through a ST.
The proposed change does not affect the configuration, or operation
of the plant. The proposed change to the CT will facilitate
installation of a plant modification which will improve plant design
and will eliminate the necessity to shut down both Units if XST1
fails or requires maintenance that goes beyond the current TS CT of
72 hours. This change will improve the long-term reliability of the
138kV offsite circuit ST which is common to both CPNPP Units.
There are no changes to the STs or the supporting systems
operating characteristics or conditions. The change to the CT does
not change any existing accident scenarios, nor create any new or
different accident scenarios. In addition, the change does not
impose any new or different requirements or eliminate any existing
requirements. The change does not alter any of the assumptions made
in the safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not affect the acceptance criteria for
any analyzed event nor is there a change to any safety limit. The
proposed change does not alter the manner in which safety limits,
limiting safety system settings, or limiting conditions for
operation are determined. Neither the safety analyses nor the safety
analysis acceptance criteria are affected by this change. The
proposed change will not result in plant operation in a
configuration outside the current design basis. The proposed
activity only increases, for two, one-time pre-planned occurrences,
the period when the plant may operate with one offsite power source.
The margin of safety is maintained by maintaining the ability to
safely shut down the plant and remove residual heat.
Therefore, the proposed change does not involve a 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: Timothy P. Matthews, Esq., Morgan, Lewis and
Bockius, 1111 Pennsylvania Avenue NW, Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: January 3, 2012.
Description of amendment request: The amendment proposes to revise
License Condition 2.B(6)(d) ``Physical Protection.'' It is proposed to
update the title of the Physical Security Plan, from the ``Maine Yankee
Nuclear Power Station Physical Security Plan'', the ``Maine Yankee
Nuclear Atomic Power Station Guard Training and Qualification Plan'',
and the ``Maine Yankee Nuclear Power Safeguards Contingency Plan'' to
the ``Maine Yankee Independent Spent Fuel Storage Installation Physical
Security 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 proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment is a title change only. There is no
reduction in commitments in the Maine Yankee Independent Spent Fuel
Storage Installation Physical Security Plan 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 amendment is a title change only. There is no
reduction in commitments in the Maine Yankee Independent Spent Fuel
Storage Installation Physical Security Plan 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 amendment is a title change only. There is no
reduction in commitments in the Maine Yankee Independent Spent Fuel
Storage Installation Physical Security Plan 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: Joseph Fay, Maine Yankee Atomic Power
Company, 362 Injun Hollow Road, East Hampton, Connecticut, 06424-3099.
NRC Branch Chief: Michele M. Sampson.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: December 6, 2012.
Description of amendment request: The amendment proposes to revise
the
[[Page 14133]]
Monticello Nuclear Generating Plant (MNGP) Technical Specification (TS)
Limiting Condition for Operation 3.10.1, ``Inservice Leak and
Hydrostatic Testing Operation,'' and the associated Bases, to expand
its scope to include provisions for temperature excursions greater than
212[emsp14][deg]F as a consequence of inservice leak and hydrostatic
testing, and as a consequence of scram time testing initiated in
conjunction with an inservice leak or hydrostatic test, while
considering operational conditions to be in MODE 4. The change is
consistent with NRC-approved Technical Specification Task Force (TSTF)
Improved Standard Technical Specifications Change Traveler, TSTF-484,
Revision 0, ``Use of TS 3.10.1 for Scram Time Testing Activities.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration,
which is provided below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Technical Specifications currently allow for operation at
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. Extending the activities that can apply this allowance will not
adversely impact the probability or consequences of an accident
previously evaluated. 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 previously evaluated?
Response: No.
Technical Specifications currently allow for operation at
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. No new operational conditions beyond those currently allowed by
LCO 3.10.1 are introduced. 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. In addition, the changes do not impose any new or
different requirements or eliminate any existing requirements. The
changes do not alter assumptions made in the safety analysis. The
proposed changes are consistent with the safety analysis assumptions
and current plant operating practice. 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
the margin of safety?
Response: No.
Technical Specifications currently allow for operation at
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. Extending the activities that can apply this allowance will not
adversely impact any margin of safety. Allowing completion of
inspections and testing and supporting completion of scram time
testing initiated in conjunction with an inservice leak or
hydrostatic test prior to power operation results in enhanced safe
operations by eliminating unnecessary maneuvers to control reactor
temperature and pressure. 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: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Robert D. Carlson.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: December 21, 2012.
Description of amendment request: The amendment proposes to revise
the Monticello Nuclear Generating Plant (MNGP) Emergency Plan by
revising the Emergency Action Level (EAL) setpoint for the Turbine
Building Normal Waste Sump (TBNWS) Monitor. The proposed change reduces
the classification of a liquid effluent release via the TBNWS pathway
to approximately 48 times the Offsite Does Calculation Manual (ODCM)
limit from the current 200 times the ODCM limit, thus establishing a
value within the indication capability of the radiation monitor.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration,
which is provided below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the emergency plan 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. No operating procedures or
administrative controls that function to prevent or mitigate
accidents are 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. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed change does 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 change in
the method of plant operation, or new operator actions. 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. The proposed change revises an emergency action
level (EAL), which establishes the threshold for placing the plant
in an emergency classification. EALs are not initiators of any
accidents.
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
the margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation does to the public. The proposed change is
associated with the EALs and does not impact operation of the plant
or its response to transients or accidents. The change does not
affect the technical specifications or the operating license. The
proposed change does not involve a change in the method of plant
operation, and no accident analyses will be affected by the proposed
change. Additionally, the proposed change 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 shutdown the plant and to maintain the plant in a safe
shutdown condition.
The revised EAL provides more appropriate and accurate criteria
for determining protective measures that should be considered within
and outside the site boundary to protect public health and safety.
The emergency plan will continue to activate an emergency response
commensurate with the extent of degradation of plant 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
[[Page 14134]]
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: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Robert D. Carlson.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello
Nuclear Generating Plant (MNGP), Wright County, Minnesota
Date of amendment request: January 4, 2013.
Description of amendment request: The licensee proposed to revise
the MNGP Technical Specifications (TS) 3.6.4.3, ``Standby Gas Treatment
(SGT) System,'' TS 3.7.4, ``Control Room Emergency Filtration (CREF)
System,'' and TS 5.5.6, ``Ventilation Filter Testing Program (VFTP).''
The licensee proposed to modify the TS requirements to operate
ventilation systems with charcoal filters from 10 hours each month to
15 minutes in accordance with Technical Specifications Task Force
(TSTF) Traveler TSTF-522, Revision 0, ``Revise Ventilation System
Surveillance Requirements to Operate for 10 hours per Month.''
Specifically, the licensee proposed to revise the surveillance
requirements STET which currently require testing of SGT and CREF
Systems, with heaters operating, for a continuous 10 hour period every
31 days without the heaters operating. The associated SRs are proposed
to be revised to require operation of these systems for 15 continuous
minutes every 31 days. Additionally, the licensee proposed to remove
Specification 5.5.6, Item e, under the VFTP, concerning operation of
the SGT and CREF Systems heaters.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration,
which is provided below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change replaces existing SRs to operate the SGT
System and CREF System equipped with electric heaters for a
continuous 10 hour period every 31 days with a requirement to
operate the systems for 15 continuous minutes (without the heaters
operating) and removes a no longer required SR under the VFTP.
These systems are not accident initiators and, therefore, these
changes do not involve a significant increase in the probability of
an accident. The proposed system and filter testing changes are
consistent with current regulatory guidance for these systems and
will continue to assure that these systems perform their design
function which may include mitigating accidents. Thus, the changes
do not involve a significant increase in the consequences of an
accident.
Therefore, it is concluded that these 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.
The proposed changes replaces existing SRs to operate the SGT
System and CREF System equipped with electric heaters for a
continuous 10 hour period every 31 days with a requirement to
operate the systems for 15 continuous minutes (without the heaters
operating) and removes a no longer required SR under the VFTP.
The change proposed for these ventilation systems does not
change any systems operations or maintenance activities. Testing
requirements will be revised and will continue to demonstrate that
the Limiting Conditions for Operation (LCO) are met and the system
components are capable of performing their intended safety
functions. The changes do not create new failure modes or mechanisms
and no new accident precursors are generated.
Therefore, it is concluded that these changes do 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 replaces existing SRs to operate the SGT
System and CREF System equipped with electric heaters for a
continuous 10 hour period every 31 days with a requirement to
operate the systems for 15 continuous minutes (without the heaters
operating) and removes a no longer required SR under the VFTP.
Testing requirements will be revised and will continue to
demonstrate that the LCOs are met and the system components are
capable of performing their intended safety functions.
The proposed changes are consistent with regulatory guidance.
Therefore, it is concluded that these changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for the licensee: Peter M. Glass, Assistant General
Counsel, Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN
55401
NRC Branch Chief: Robert D. Carlson.
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 amendment request: December 13, 2012.
Description of amendment request: The proposed amendments would
revise the Prairie Island Nuclear Generating Plant Emergency Plan by
revising certain emergency action levels described in the 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 proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This license amendment request proposes to revise Emergency Plan
emergency action levels for classification of liquid effluent
releases and determining fuel clad barrier loss. These changes
propose to use installed plant radiation monitors differently but do
not involve any physical plant changes.
The Emergency Plan emergency action levels and installed plant
radiation monitors are not accident initiators and therefore the
proposed changes do not involve an increase in the probability of an
accident. The proposed emergency action level changes do not affect
the capability of any structures, system or components to mitigate a
design basis accident. Thus the proposed changes do not involve a
significant increase in the consequences of an accident.
Therefore, the proposed Emergency Plan emergency action level
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 to revise Emergency Plan
emergency action levels for classification of liquid effluent
releases and determining fuel clad barrier loss. These changes
propose to use installed plant radiation monitors differently but do
not involve any physical plant changes.
The proposed Emergency Plan emergency action level changes do
not change any system operations or maintenance activities. The
changes do not involve physical alteration of the plant, that is, no
new or different type of equipment will be installed. The changes do
not alter assumptions made in the safety analyses but ensures that
the plant Emergency Plan is effectively and consistently
implemented. These changes do not create new failure modes or
mechanisms which are not identifiable during testing and no new
accident precursors are generated.
[[Page 14135]]
Therefore, the proposed Emergency Plan emergency action level
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 to revise Emergency Plan
emergency action levels for classification of liquid effluent
releases and determining fuel clad barrier loss. These changes
propose to use installed plant radiation monitors differently but do
not involve any physical plant changes.
Margin of safety is provided by the ability of accident
mitigation structures systems or components to perform at their
analyzed capability. The changes proposed in this license amendment
request do not affect the capability of any equipment to perform its
accident mitigation function. Thus, no margin of safety is reduced
as part of this change.
Therefore, the proposed Emergency Plan emergency action level
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: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Robert D. Carlson.
South Carolina Electric and Gas Docket Nos.: 52-027 and 52-028, Virgil
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: February 7, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear
Station (VCSNS) Units 2 and 3 in regard to the Primary Sampling System
(PSS) by: (1) Replacing containment air return check valve PSS-PL-V024
with a solenoid-operated valve, and (2) redesigning the PSS inside-
containment header and adding a PSS containment penetration.
Because, this proposed change requires a departure from Tier 1
information in the Westinghouse Advanced Passive 1000 design control
document (DCD), the licensee also requested an exemption from the
requirements of the Generic DCD Tier 1 in accordance with 52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Primary Sampling System (PSS) provides the safety-related
function of preserving containment integrity by isolation of the PSS
lines penetrating containment. The proposed amendment will enhance
the ability of the PSS to perform its nonsafety-related function of
providing the capability to obtain reactor coolant and containment
atmosphere samples, while maintaining the ability of the PSS to
perform its safety-related containment isolation function. The
replacement of a check valve with a solenoid-operated containment
isolation valve and the redesigned inside-containment header does
not affect the safety-related function of isolating the PSS lines
for containment isolation. The components added by this proposed
activity, including tubing and the solenoid-operated containment
isolation valve, are designed to the same codes and standards as
other components addressed in the certified design that perform
similar functions. The additional PSS containment penetration is a
passive extension of containment and is identical in form, fit, and
function to other PSS sampling containment penetrations currently
addressed in the certified AP1000 plant design. The addition of a
new PSS containment penetration will not change the maximum
allowable leakage rate allowed by Technical Specifications and
verified periodically in accordance with regulations. Furthermore,
the proposed PSS configuration changes will neither impact any
accident source term parameter or fission product barrier nor affect
radiological dose consequence analysis.
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 additional containment penetration is similar in form, fit,
and function to the PSS penetrations that are currently described in
the Updated Final Safety Analysis Report. Because the PSS changes
use valve types, piping, and a containment penetration consistent
with those already described in the Updated Final Safety Analysis
Report, no new failure modes or equipment failure initiators are
introduced by these changes. Accordingly, the proposed changes do
not create any new malfunctions, failure mechanisms, or accident
initiators.
Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The containment isolation function is not changed by this
activity and is bounded by the existing design. The proposed PSS
containment penetration is similar in form, fit, and function to
other containment penetrations in similar applications in the
current certified AP1000 plant design. The additional PSS
containment penetration is an extension of containment, and,
therefore, does not affect containment or its ability to perform its
design function. The addition of PSS components, including the
solenoid-operated containment isolation valve, the additional PSS
containment penetration, and the associated tubing, do not exceed or
alter a design basis or safety limit. Because the containment
isolation function, containment leakage rate limit, potential
containment leakage, and protective shielding are not changed by
this activity and are bounded by the existing design, there is no
change to any current margin of safety.
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: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart, Acting.
South Carolina Electric and Gas Docket Nos.: 52-027 and 52-028, Virgil
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: February 14, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear
Station (VCSNS) Units 2 and 3 in regard to the structural module stud
size and spacing by increasing the carbon steel vertical stud spacing,
decreasing the stainless steel stud diameter, and decreasing the
stainless steel vertical and horizontal stud spacing in accordance with
the design basis.
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.
[[Page 14136]]
The design function of the containment modules is to support the
reactor coolant system components and related piping systems and
equipment. The design functions of the affected structural module in
the auxiliary building are to provide support and protection for new
and spent fuel and the equipment needed to support fuel handling,
cooling, and storage in the spent fuel racks, and to provide
support, protection, and separation for the seismic Category I
mechanical and electrical equipment located outside the containment
building. The design function of the shear studs it to transfer
loads into the concrete of the structural modules. The proposed
change corrects a drawing note regarding shear stud size and spacing
for structural wall modules to be consistent with the underlying
design basis calculations, which are more conservative. The
thickness, geometry, and strength of the structures are not
adversely altered. The properties of the concrete included in the
modules are not altered. As a result, the design function of the
structural modules is not adversely affected by the proposed change.
There is no change to plant, systems or the response of systems to
postulated accident conditions. There is no change to the predicted
radioactive releases due to normal operation or postulated accident
conditions. The plant response to previously evaluated accidents or
external events is not adversely affected, nor does the change
described create any new accident precursors.
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 change corrects a drawing note regarding shear stud
size and spacing for structural wall modules to be consistent with
the underlying design basis calculations. Stud spacing and sizing
are updated such that stud loadings are within acceptable limits and
that the structural module acts in a composite manner. The
thickness, geometry, and strength of the structures are not
adversely altered. The properties of the concrete included in the
modules are not altered. The change to the internal design of the
structural modules does not create any new accident precursors. As a
result, the design function of the modules is not adversely affected
by the proposed change.
Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The criteria and requirements of AISC-N690 provide a margin of
safety to structural failure. The design of the shear studs for the
structural wall modules conforms to criteria and requirements in
AISC-N690 and therefore maintains the margin of safety. The proposed
change corrects a drawing note regarding shear stud size and spacing
for the structural wall modules so as to be consistent with the
underlying design basis calculations. There was no change to the
method of evaluation from that used in the design basis
calculations.
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: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart, Acting.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: February 7, 2013 and revised on February
14, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos.: NPF-93 and NPF-94 for Virgil C. Summer Nuclear
Station (VCSNS) Units 2 and 3 to allow the use of concentrically and
eccentrically braced frames in the turbine building main area and
modify the applicable design code.
Because, this proposed change requires a departure from Tier 1
information in the Westinghouse Advanced Passive 1000 design control
document (DCD), the licensee also requested an exemption from the
requirements of the Generic DCD Tier1 in accordance with 52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The turbine building bracing design is changed to a mixed
bracing system which uses special concentric and eccentric bracing.
The turbine building does not contain safety-related systems or
components. The main area of the turbine building continues to meet
its design function of preventing a turbine building collapse from
impairing the integrity of seismic Category I structures, systems,
or components. The first bay of the turbine building is designed to
prevent the collapse of the main area of the Turbine Building onto
the Nuclear Island during a seismic event. The proposed changes do
not affect or impact this design capability. Therefore, the response
of the safety related systems, structures, and components in the
Nuclear Island to earthquakes and postulated accidents are not
affected by the bracing of the turbine building. Based on the above,
there is no change in the probability of an accident previously
evaluated. The activity does not introduce a new fission product
release path, result in a new fission product barrier failure mode,
or create a new sequence of events that result in significant fuel
cladding failures. Accordingly, there is no change in the
consequences of an 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 turbine building bracing design is changed to a mixed
bracing system which uses Special Concentrically Braced Framing
(SCBF) and Eccentrically Braced Framing (EBF). The main area of the
turbine building continues to meet its design function of preventing
a turbine building collapse from impairing the integrity of seismic
Category I structures, systems, or components. The design function
of the turbine building first bay to provide the intended
limitations to a potential collapse onto the nuclear island during a
seismic event is retained. The turbine building structure does not
involve any accident initiating component and therefore, changes to
use SCBF and EBF would not introduce new accident components or
faults.
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.
Use of a mixed bracing system and changing the structural code
design for the turbine building main area continue to meet the
design function of preventing a turbine building collapse from
impairing the integrity of seismic Category I Structures, Systems,
and Components. In addition, the first bay of the turbine building
continues to be designed to seismic Category II requirements to
prevent a turbine building collapse from impairing the integrity of
the seismic Category I nuclear island structures, systems and
components. This portion of the turbine building and its design is
unchanged by the proposed amendment. Maintaining the seismic
Category II rating for the turbine building first bay, along with
continuing to meet the design function for the non-safety, non-
seismic design of the turbine building main area preserves the
current structural safety margins.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
[[Page 14137]]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart, Acting.
Southern Nuclear Operating Company Docket Nos.: 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County,
Georgia
Date of amendment request: January 11, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-91 and NPF-92 for Vogtle Electric Generating
Plant (VEGP) Units 3 and 4 in regard to the Chemical and Volume Control
System (CVS) by: (1) Providing a spring-assisted check valve around the
air-operated Reactor coolant System (RCS) Purification Return Line Stop
Check Valve, (2) replacing the CVS zinc addition inboard containment
isolation lift check valve with an air-operated globe valve and a
thermal relief valve and (3) separating the zinc and hydrogen injection
paths and relocate the zinc injection path.
Because, this proposed change requires a departure from Tier 1
information in the Westinghouse Advanced Passive 1000 design control
document (DCD), the licensee also requested an exemption from the
requirements of the Generic DCD Tier 1 in accordance with 52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The changes to provide a spring-assisted check valve located in
the bypass line around the makeup stop check valve would continue to
meet the existing design functions because the ASME Boiler and
Pressure Vessel Code (ASME Code) Section III valves will maintain
the flow isolation design function and preserve the Reactor Coolant
System (RCS) pressure boundary safety function. The replacement of
the Chemical and Volume Control System (CVS) zinc addition inboard
containment isolation lift check valve with an air operated globe
valve and addition of a pressure relief valve would continue to meet
the containment isolation and RCS pressure boundary design functions
because the replacement valves will be designed, analyzed, tested
and qualified, including seismic qualification, to ASME Code Section
III requirements. Separating the zinc and hydrogen injection paths
and relocating the zinc injection point would continue to meet
containment boundary requirements, including containment isolation
and in-service testing, and preserve the RCS pressure boundary
safety functions because the revised containment isolation
configuration is consistent with those described in 10 CFR 50,
Appendix A, General Design Criterion (GDC) 55, and the additional
valves and piping will be qualified to ASME Code Section III.
Because the proposed CVS changes would preserve the CVS safety-
related design functions, the probability of an accident previously
evaluated is not affected.
The CVS safety functions have been preserved, because the
proposed CVS configuration changes, including revised valve types,
will perform the same safety functions as the current design. The
proposed CVS configuration changes would neither impact any accident
source term parameter or fission product barrier nor affect
radiological dose consequence analysis.
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.
The additional containment penetration is similar in form, fit,
and function to the CVS combined zinc/hydrogen containment
penetration that is currently described in the Updated Final Safety
Analysis Report. Because the CVS changes use valve types, piping,
and a containment penetration consistent with those already
described in the Updated Final Safety Analysis Report, no new
failure modes or equipment failure initiators are introduced by
these changes. Accordingly, the proposed changes do not create any
new malfunctions, failure mechanisms, or accident initiators.
Therefore, the proposed changes do 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 containment isolation and pressure relief functions would
not be changed by this activity and are consistent with the existing
design. The proposed CVS containment penetration is similar in form,
fit, and function to existing CVS combined zinc/hydrogen containment
penetration and, therefore, does not affect containment or its
ability to perform its design function. The addition of these CVS
components, including piping, a spring-assisted check valve, an air-
operated containment isolation valve, a thermal relief valve and the
additional CVS containment penetration do not impact a design basis
or safety limit. Because the CVS design functions of controlling the
RCS oxygen concentration, reducing radiation fields, containment
isolation and overpressure protection within existing limits are not
changed by this activity and are bounded by the existing design,
there is no change to any current 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 request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence Burkhart, Acting.
Southern Nuclear Operating Company Docket Nos.: 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County,
Georgia
Date of amendment request: February 7, 2013 and revised on February
15, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos.: NPF-91 and NPF-92 for Vogtle Electric Generating
Plant (VEGP) Units 3 and 4 to allow the use of concentrically and
eccentrically braced frames in the turbine building main area and
modify the applicable design code.
Because this proposed change requires a departure from Tier 1
information in the Westinghouse Advanced Passive 1000 design control
document (DCD), the licensee also requested an exemption from the
requirements of the Generic DCD Tier1 in accordance with 52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The turbine building bracing design is changed to a mixed
bracing system which uses special concentric and eccentric bracing.
The turbine building does not contain safety-related systems or
components. The main area of the turbine building continues to meet
its design function of preventing a turbine building collapse from
impairing the
[[Page 14138]]
integrity of seismic Category I structures, systems, or components.
The first bay of the turbine building is designed to prevent the
collapse of the main area of the Turbine Building onto the Nuclear
Island during a seismic event. The proposed changes do not affect or
impact this design capability. Therefore, the response of the safety
related systems, structures, and components in the Nuclear Island to
earthquakes and postulated accidents are not affected by the bracing
of the turbine building. Based on the above, there is no change in
the probability of an accident previously evaluated. The activity
does not introduce a new fission product release path, result in a
new fission product barrier failure mode, or create a new sequence
of events that result in significant fuel cladding failures.
Accordingly, there is no change in the consequences of an 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 turbine building bracing design is changed to a mixed
bracing system which uses Special Concentrically Braced Framing
(SCBF) and Eccentrically Braced Framing (EBF). The main area of the
turbine building continues to meet its design function of preventing
a turbine building collapse from impairing the integrity of seismic
Category I structures, systems, or components. The design function
of the turbine building first bay to provide the intended
limitations to a potential collapse onto the nuclear island during a
seismic event is retained. The turbine building structure does not
involve any accident initiating component and therefore, changes to
use SCBF and EBF would not introduce new accident components or
faults.
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.
Use of a mixed bracing system and changing the structural code
design for the turbine building main area continue to meet the
design function of preventing a turbine building collapse from
impairing the integrity of seismic Category I Structures, Systems,
and Components. In addition, the first bay of the turbine building
continues to be designed to seismic Category II requirements to
prevent a turbine building collapse from impairing the integrity of
the seismic Category I nuclear island structures, systems and
components. This portion of the turbine building and its design is
unchanged by the proposed amendment. Maintaining the seismic
Category II rating for the turbine building first bay, along with
continuing to meet the design function for the non-safety, non-
seismic design of the turbine building main area preserves the
current structural safety margins.
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: Mr. M. Stanford Blanton, Blach & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence Burkhart, Acting.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: December 13, 2012.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.7.9, ``Ultimate Heat Sink (UHS),'' to
incorporate more restrictive UHS level and pond temperature limits
which are specified in Surveillance Requirements (SRs) 3.7.9.1 and
3.7.9.2, respectively. In addition, new SR 3.7.9.4 would be added to
verify that the UHS cooling tower fans respond appropriately to
automatic start signals.
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.
There are no design changes associated with the proposed
amendment. All design, material, and construction standards that
were applicable prior to this amendment request will continue to be
applicable. The proposed change will not adversely affect accident
initiators or precursors or adversely alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained with respect to such initiators
or precursors. The proposed changes do not affect the way in which
safety-related systems perform their functions.
All accident analysis acceptance criteria will continue to be
met with the proposed changes. The proposed changes will not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. The proposed changes will not alter
any assumptions or change any mitigation actions in the radiological
consequence evaluations in the FSAR [final safety analysis report].
The applicable radiological dose acceptance criteria will continue
to be met.
The intent of the modified UHS water level and temperature
limits for TS 3.7.9, as proposed, is to ensure that the UHS can
perform its specified safety function for accident mitigation,
including consideration of its 30-day mission time. The proposed
surveillance limits are more restrictive and are based on an
analysis that includes credit given to specific operator actions
(with assumed completion times) not previously assumed. However, the
operator actions are reasonable and have been established in
accordance with NRC-approved guidance. Further, they have been
simulator verified and proven to be capable of being met by plant
operators under applicable accident scenarios.
The crediting of these operator actions is consistent with the
plant's current licensing basis which already credits operator
action to provide long-term protection of the UHS following an
accident. These actions, in conjunction with the more restrictive
proposed UHS water temperature and level surveillance limits,
support the plant's existing accident analysis such that there is no
change in analyzed consequences. In light of these considerations,
there is no significant increase in the consequences of any accident
previously evaluated with regard to the assumed operator actions and
revised UHS water level and temperature limits, as proposed. The
proposed change adds additional controls to the Technical
Specifications but does not physically alter safety-related systems
or affect the way in which safety-related systems perform their
functions per the intended plant design.
As such, the proposed change will not alter or prevent the
capability of structures, systems, and components (SSCs) to perform
their intended functions for mitigating the consequences of an
accident and meeting applicable acceptance limits. 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.
With respect to any new or different kind of accident, there are
no proposed design changes nor are there any changes in the method
by which any safety-related plant SSC performs its specified safety
function. The proposed change will not affect the normal method of
plant operation. No new transient precursors will be introduced as a
result of this amendment. The reanalysis discussed herein addresses
new large break LOCA [loss-of-coolant accident] scenarios with
assumptions, including single failures, aimed at maximizing the UHS
temperature and minimizing the UHS inventory.
The proposed change adds requirements to the Technical
Specifications. The change does not involve a physical modification
of the plant. The UHS level and temperature limits within which the
plant is normally operated are being changed in the
[[Page 14139]]
conservative direction. Appropriate changes have been made to the
emergency operating procedures relied upon to mitigate a design
basis event. The change does not have a detrimental impact on the
manner in which plant equipment operates or responds to an actuation
signal. The changes to the ultimate heat sink (UHS) surveillance
limits are in the conservative direction.
The proposed change does not, therefore, create the possibility
of a new or different accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
There will be no effect on those plant systems necessary to
assure the accomplishment of protection functions associated with
reactor operation or the reactor coolant system. There will be no
impact on the overpower limit, departure from nucleate boiling ratio
(DNBR) limits, heat flux hot channel factor (FQ), nuclear
enthalpy rise hot channel factor (F[Delta]H), loss of coolant
accident peak cladding temperature (LOCA PCT), peak local power
density, or any other limit and associated margin of safety.
Required shutdown margins in the COLR [core operating limits report]
will not be changed.
The proposed change does not eliminate any surveillances or
alter the frequency of surveillances required by the Technical
Specifications. The proposed change would add Technical
Specification Surveillance Requirements for assuring the automatic
closure of the UHS cooling tower bypass valves when required and the
automatic start of the UHS cooling tower fans and their transition
from slow speed to fast speed when required. The extent of
Callaway's conformance to NRC Regulatory Guide (RG) 1.27 is
discussed in FSAR Site Addendum Table 9.2-5 (see Attachment 4 to
this Enclosure [to the submittal]). RG 1.27 requires that the UHS be
sized for 30 day post-LOCA operation; however, it does not specify a
margin value above that 30-day requirement. During initial plant
licensing (Callaway Safety Evaluation Report, NUREG-0830, Supplement
4, Section 2.4.4) a UHS level margin of 50% was accepted in lieu of
a more restrictive minimum Technical Specification water level of
834 feet mean sea level (16 feet above the reference pond bottom)
and a thermal and hydrologic analysis of the ESW [essential service
water] and UHS. In this amendment request SR 3.7.9.1 is being
changed to adopt the former and the supporting EF-123 analysis
addresses the latter. The SER [safety evaluation report] Supplement
4 discussion, copied in Section 2.2 of this Evaluation, will no
longer be applicable upon NRC approval of this license amendment
request.
As such, the proposed change does not involve a significant
reduction in a margin of safety as defined in any regulatory
requirement or guidance document.
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: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: December 20, 2012.
Description of amendment request: The amendment would revise a
methodology in the licensing basis as described in the Final Safety
Analysis Report--Standard Plant to include damping values for the
seismic design and analysis of the integrated head assembly that are
consistent with the recommendations of NRC Regulatory Guide 1.61,
``Damping Values for Seismic Design of Nuclear Power Plants,'' Revision
1, March 2007.
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 change would allow use of critical damping values
consistent with the recommendations of RG [Regulatory Guide] 1.61,
``Damping Values for Seismic Design of Nuclear Power Plants,''
Revision 1, dated March 2007, for the seismic design and analysis of
the IHA [integrated head assembly].
The RG 1.61, Revision 1, Table 1 note allowing use of a
``weighted average'' for design-basis SSE [safe shutdown earthquake]
damping values applicable to steel structures of different
connection types, is also applied to determine the IHA design-basis
OBE [operating basis earthquake] damping values. RG 1.61, Revision
1, Table 2 for OBE damping values does not contain the same note
found in Table 1. However use of the note for the determination of
the OBE damping value is consistent with the use of the note for the
determination of the SSE damping values, and a weighted average more
realistically represents the IHA structure. RG 1.61, Revision 1,
specifies the damping values that the NRC staff currently considers
acceptable for complying with the agency's regulations and guidance
for seismic analysis. Revision 1 incorporates the latest data and
information, and reduces unnecessary conservatism in specification
of damping values for seismic design and analysis of SSCs
[structures, systems, and components].
The proposed change does not change the design functions of the
IHA or its response to design-basis events, nor does it affect the
capability of related SSCs to perform their design or safety
functions. The use of the proposed damping values in the seismic
design and analysis of the IHA is related to the ability of the IHA
to function in response to design-basis seismic events, and is
unrelated to the probability of occurrence of those events, or other
previously evaluated accidents. Therefore, the proposed change will
not have any impact on the probability of an accident previously
evaluated.
The proposed damping values are an element of the seismic
analyses performed to confirm the ability of the IHA to function
under postulated seismic events while maintaining resulting stresses
within ASME [American Society of Mechanical Engineers Boiler and
Pressure Vessel Code] Section III allowable values. Therefore, the
use of damping values consistent with the recommendations of RG
1.61, Revision 1 does not result in an increase in the consequences
of accidents previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve changes to any plant SSCs,
nor does it involve changes to any plant operating practice or
procedure. The damping values are an element of the seismic analyses
performed to confirm the ability of the IHA to function under
postulated seismic events while maintaining resulting stresses
within ASME Section III allowable values. Therefore, no credible new
failure mechanisms, malfunctions, or accident initiators not
considered in the design and licensing bases are created that would
create the possibility of a new or different kind of accident.
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 change involve a significant reduction in a margin
of safety?
Response: No.
The design basis of the plant requires structures to be capable
of withstanding normal and accident loads including those from a
design basis earthquake. The proposed change would allow the use of
damping values in the IHA seismic analyses that are, in general,
more realistic and, thus, more accurate than the damping values
recommended in RG 1.61, Revision 0, used in the original analysis
for the SSE, or the plant specific damping values used in the
original analysis for the OBE. The damping values in RG 1.61,
Revision 0, were based on limited data, expert opinion, and other
information available in 1973. NRC and industry research since 1973
shows that the damping values provided in the original version of RG
1.61 may not reflect realistic damping values for SSCs. RG 1.61,
Revision 1, therefore, provides damping values based on the updated
research results that predict
[[Page 14140]]
and estimate damping values for seismic design of SSCs in nuclear
power plants, and similarly should not be regarded as an arbitrary
lowering of the margins 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: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Notice of Issuance of Amendments to Facility Operating Licenses and
Combined 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.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, 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's Reference staff at 1-800-397-4209, 301-415-4737 or by
email to [email protected].
Carolina Power and Light Company, et al., Docket No. 50-261, H.B.
Robinson Steam Electric Plant, Unit No. 2, Darlington County, South
Carolina
Date of application for amendment: March 16, 2012, as supplemented
by letter dated August 16, 2012.
Brief Description of amendment: The amendment revised the Technical
Specifications (TSs) to make corrections in TS Table 3.3.1-1 for
Overtemperature Delta Temperature consistent with NUREG-1431, Revision
3, ``Standard Technical Specifications Westinghouse Plants.''
Date of issuance: February 13, 2013.
Effective date: As of date of issuance and shall be implemented
within 120 days.
Amendment No.: 231.
Renewed Facility Operating License No. DPR-23: Amendment changed
the license and TSs.
Date of initial notice in Federal Register: April 17, 2012 (77 FR
22811). The supplement dated August 16, 2012, 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 February 13, 2013.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and
2, Ogle County, Illinois
Date of application for amendment: June 6, 2012, as supplemented by
letter dated. November 19, 2012.
Brief description of amendment: The proposed amendment modifies
Braidwood and Byron technical specifications (TS) to add a Note to
surveillance requirements (SRs) 3.3.1.7, 3.3.1.8, and 3.3.1.12 in TS
3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' and SRs 3.3.2.2
and 3.3.2.6 in TS 3.3.2, ``Engineered Safety Features Actuation System
(ESFAS) Instrumentation,'' to exclude the Solid State Protection System
input relays from the Channel Operational Test Surveillance for RTS and
ESFAS functions with installed bypass capability which the U.S. Nuclear
Regulatory Commission (NRC) approved by letters dated March 30, and
April 9, 2012.
Date of issuance: February 6, 2013.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 171 for Braidwood Station, Units 1 and 2, and 178
for Byron Station, Unit Nos. 1 and 2, respectively.
Facility Operating License Nos. NPF-72. NPF-77, NPF-37, and NPF-66:
The amendments revised the Technical Specifications and License.
Date of initial notice in Federal Register: September 4, 2012 (77
FR 53927).
The November 19, 2012, 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 amendments is contained
in a Safety Evaluation dated February 6, 2013.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant (BFN), Units 2 and 3, Limestone County, Alabama
Date of application for amendments: February 25, 2011, as
supplemented by letters dated September 15, 2011, July 30, 2012, and
January 24, 2013. The enclosure to the July 30, 2012, letter
superseded, in its entirety, the enclosure to the February 25, 2011,
letter.
Brief description of amendments: The amendments delete the BFN,
Units 2 and 3, Technical Specification (TS) Surveillance Requirement
3.5.1.12, which requires the verification of the capability to
automatically transfer the power supply from the normal source to the
alternate source for each Low-Pressure Coolant Injection subsystem
inboard injection valve and each recirculation pump discharge valve on
a 24-month frequency. In addition, these amendments approve the use of
a modified loss-of-coolant accident
[[Page 14141]]
(LOCA) methodology that requires revising TS 5.6.5.b to include a
reference to the modified LOCA methodology. Also, the amendments revise
TSs 3.3.1.1, 5.6.5.a, and 5.6.5.b to include the modified LOCA
methodology and the oscilliation power range monitor upscale function
period based detection algorithm setpoint limits.
Date of issuance: February 15, 2013.
Effective date: The amendments are effective as of this date of
issuance. For Unit 2, the amendment shall be implemented prior to
entering Mode 3 (i.e., Hot Shutdown) from the spring 2013 refueling
outage. For Unit 3, changes to TSs 5.6.5 and 3.3.1 shall be implemented
within 60 days of issuance. The remaining changes shall be implemented
prior to entering Mode 3 from the spring 2014 refueling outage.
Amendment Nos.: Unit 1--309 and Unit 2--268.
Renewed Facility Operating License Nos. DPR-52 and DPR-68:
Amendments revised the licenses and TSs.
Date of initial notice in Federal Register: The original
application dated February 25, 2011, was noticed on May 3, 2011 (76 FR
24930). The supplement dated July 30, 2012, was noticed on November 5,
2012 (77 FR 66490). The supplement dated January 24, 2013, provided
additional information that clarified the licensee's July 30, 2012,
submittal, did not expand the scope of the application as noticed and
did not change the NRC staff's proposed no significant hazards
consideration determination as published in the FR on November 5, 2012
(77 FR 66490).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 15, 2013.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket No. 50-339, North Anna
Power Station, Unit No. 2, Louisa County, Virginia
Date of application for amendment: May 11, 2012.
Brief Description of amendment: The amendment would revise the
Technical Specification (TS) 3.1.7, ``Rod Position Indication'' to
allow two demand position indicators in one or more banks to be
inoperable for up to 4 hours. This change is proposed as a temporary
change to the TS for the current operating cycle and is proposed as a
footnote to the current TS Limiting Condition for Operation (LCO)
Section 3.1.7, Condition D.
Date of issuance: February 14, 2013.
Effective date: As of the date of issuance and shall be implemented
within the end of operating Cycle 22.
Amendment No.: 251.
Renewed Facility Operating License No. NPF-7: Amendment changes the
license and the TS.
Date of initial notice in Federal Register: June 12, 2012 (77 FR
35077).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 14, 2013.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 25th day of February 2013.
For the Nuclear Regulatory Commission.
Louise Lund,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2013-04885 Filed 3-1-13; 8:45 am]
BILLING CODE 7590-01-P