[Federal Register Volume 79, Number 130 (Tuesday, July 8, 2014)]
[Notices]
[Pages 38585-38597]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-15770]
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NUCLEAR REGULATORY COMMISSION
[NRC-2014-0159]
Biweekly Notice, Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission
(NRC) is publishing this regular biweekly notice. The Act requires the
Commission to 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 June 12, to June 25, 2014. The last biweekly
notice was published on June 24, 2014.
DATES: Comments must be filed by August 7, 2014. A request for a
hearing must be filed by September 8, 2014.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0159. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: 3WFN-06-A44M, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Sandra Figueroa, Office, U.S. Nuclear
Regulatory Commission, Washington DC 20555-0001; telephone: 301-415-
1262, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2014-0159 when contacting the NRC
about the availability of information for this action. You may obtain
publicly-available information related to this action by any of the
following methods:
Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0159.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection 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]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in the
SUPPLEMENTARY INFORMATION section.
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 NRC-2014-0159 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 in your comment submission. The NRC will post all comment
submissions at http://www.regulations.gov as well as enter the comment
submissions into ADAMS, and the NRC does not routinely 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 that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of Title 10 of the Code of
Federal Regulations (10 CFR), 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
[[Page 38586]]
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.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
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 ``Agency
Rules of Practice and Procedure'' 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. The
NRC's regulations are accessible electronically from the NRC Library on
the NRC's 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.
B. Electronic Submissions (E-Filing)
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's 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
ten 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
identification (ID) certificate, which allows the participant (or its
counsel or representative) to digitally sign documents and access the
E-Submittal server for any proceeding in which it is participating; and
(2) advise the Secretary that the participant will be submitting a
request or petition for hearing (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.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.
[[Page 38587]]
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
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's 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's 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 NRC'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's public
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to
[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 three factors in 10
CFR 2.309(c)(1)(i)-(iii).
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on obtaining information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: February 4, 2014. A publicly available
version is in ADAMS under Accession No. ML14050A383.
Description of amendment request: The amendment would revise
Technical Specification 5.5.15, ``Containment Leakage Rate Testing
Program,'' to extend the frequency of the Type A, or the Containment
Integrated Leak Rate Test, from 10 to 15 years on a permanent 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.
The proposed amendment involves changes to the IP3 [Indian Point
Unit No. 3] containment leakage rate testing program. The proposed
amendment does not involve a physical change to the plant or a
change in the manner in which the plant is operated or controlled.
The primary containment function is to provide an essentially leak
tight barrier against the uncontrolled release of radioactivity to
the environment for postulated accidents. As such, the containment
itself and the testing requirements to periodically demonstrate the
integrity of the containment exist to ensure the plant's ability to
mitigate the consequences of an accident do not involve any accident
precursors or initiators.
Therefore, the probability of occurrence of an accident
previously evaluated is not significantly increased by the proposed
amendment.
The proposed amendment adopts the NRC [Nuclear Regulatory
Commission] accepted guidelines of [Nuclear Energy Institute] NEI
94-01, Revision 3-A, for development of the IP3 performance-based
testing program for the Type A testing. Implementation of these
guidelines continues to provide adequate assurance that during
design basis accidents, the primary containment and its components
would limit leakage rates to less than the
[[Page 38588]]
values assumed in the plant safety analyses. The potential
consequences of extending the ILRT [integrated leak rate test]
interval to 15 years have been evaluated by analyzing the resulting
changes in risk. The increase in risk in terms of person-rem per
year within 50 miles resulting from design basis accidents was
estimated to be acceptably small and determined to be within the
guidelines published in [Regulatory Guide] RG 1.174. Additionally,
the proposed change maintains defense-in-depth by preserving a
reasonable balance among prevention of core damage, prevention of
containment failure, and consequence mitigation. Entergy has
determined that the increase in conditional containment failure
probability due to the proposed change would be very small.
Therefore, it is concluded that the proposed amendment does not
significantly increase the 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 amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, for the development of the IP3 performance-
based leakage testing program, and establishes a 15-year interval
for the performance of the containment ILRT. The containment and the
testing requirements to periodically demonstrate the integrity of
the containment exist to ensure the plant's ability to mitigate the
consequences of an accident do not involve any accident precursors
or initiators. The proposed change does not involve a physical
change to the plant (i.e., no new or different type of equipment
will be installed) or a change to the manner in which the plant is
operated or controlled.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, for the development of the IP3 performance-
based leakage testing program, and establishes a 15-year interval
for the performance of the containment ILRT. This amendment does not
alter the manner in which safety limits, limiting safety system
setpoints, or limiting conditions for operation are determined. The
specific requirements and conditions of the containment leakage rate
testing program, as defined in the TS [technical specifications],
ensure that the degree of primary containment structural integrity
and leak-tightness that is considered in the plant's safety analysis
is maintained. The overall containment leakage rate limit specified
by the TS is maintained, and the Type A, Type B, and Type C
containment leakage tests would be performed at the frequencies
established in accordance with the NRC-accepted guidelines of NEI
94-01, Revision 3-A.
Containment inspections performed in accordance with other plant
programs serve to provide a high degree of assurance that the
containment would not degrade in a manner that is not detectable by
an ILRT. A risk assessment using the current IP3 PSA [probabilistic
safety assessment] model concluded that extending the ILRT test
interval from ten years to 15 years results in a very small change
to the risk profile.
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: Ms. Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Benjamin G. Beasley.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New
York
Date of amendment request: April 1, 2014. A publicly-available
version is in ADAMS under Accession No. ML14099A227.
Description of amendment request: The amendments would revise the
technical specifications by implementing Technical Specification Task
Force Traveler 510, Revision 2, ``Revision to Steam Generator Program
Inspection Frequencies and Tube Sample Selection.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Steam Generator (SG) Program to
modify the frequency of verification of SG tube integrity and SG
tube sample selection. A steam generator tube rupture (SGTR) event
is one of the design basis accidents that are analyzed as part of a
plant's licensing basis. The proposed SG tube inspection frequency
and sample selection criteria will continue to ensure that the SG
tubes are inspected such that the probability of a SGTR is not
increased. The consequences of a SGTR are bounded by the
conservative assumptions in the design basis accident analysis. The
proposed change will not cause the consequences of a SGTR to exceed
those assumptions. The proposed change to reporting requirements and
clarifications of the existing requirements have no affect on the
probability or consequences of a SGTR.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to the SG Program will not introduce any
adverse changes to the plant design basis or postulated accidents
resulting from potential tube degradation. The proposed changes do
not affect the design of the SGs or their method of operation. In
addition, the proposed changes do not impact any other plant system
or component.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
also isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a SG
is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change will continue to require monitoring of the physical
condition of the SG tubes such that there will not be a reduction in
the margin of safety compared to the current requirements.
Changes associated with inspection frequency and tube selection
criteria are consistent with TSTF-510 Revision 2 and are based on
recent industry experience and are more effective in managing the
frequency of verification of tube integrity and sample selection
than those required by current TSs [technical specifications].
Therefore, the proposed change does not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 38589]]
Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Benjamin G. Beasley.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: October 8, 2013. A publicly-available
version is in ADAMS under Accession No. ML13282A559.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) requirements to reduce the
reactor pressure associated with the Reactor Core Safety Limit from 785
psig to 685 psig in TS 2.1.1.1 and TS 2.1.1.2. The proposed amendment
would address the potential to not meet the lower pressure TS safety
limit associated with a Pressure Regulator Failure-Maximum Demand
(Open) (PRFO) transient reported by General Electric (GE) in their 10
CFR Part 21 Communication, Potential to Exceed Low Pressure Technical
Specification Safety Limit, SC05-03, dated March 29, 2005.
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.
Decreasing the reactor pressure in TS Safety Limit 2.1.1.1 or
2.1.1.2 for reactor rated thermal power ranges effectively expands
the validity range for GEXL correlation and the calculation of
Minimum Critical Power Ratio Safety Limit (MCPR). The [critical
power ratio] CPR rises during the pressure reduction following the
scram that terminates the PRFO transient. Since the change does not
involve a modification of any plant hardware, the probability and
consequence of the PRFO transient are essentially unchanged. The
reduction in the reactor dome pressure value in the safety limit
from 800 psia (785 psig) to 700 psia (685 psig) provides greater
margin to accommodate the pressure reduction during the transient
within the revised TS limit.
The proposed change will continue to support the validity range
for GEXL correlation and the calculation of MCPR as approved. The
proposed TS revision involves no significant changes to the
operation of any systems or components in normal or accident or
transient operating conditions.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed reduction in the reactor pressure value in the
safety limit from 800 psia (785 psig) to 700 psia (685 psig)
reflects a wider range of applicability for the GEXL correlation for
fuels in use at JAF and does not involve changes to the plant
hardware or its operating characteristics. As a result, no new
failure modes are being introduced.
Therefore, the change does not introduce a new or different kind
of accident from those previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety is established through the design of the
plant structures, systems, and components, and through the
parameters for safe operation and setpoints for the actuation of
equipment relied upon to respond to transients and design basis
accidents. The proposed change in the reactor pressure safety limit
enhances the safety margin, which protects the fuel cladding
integrity during a depressurization transient, but does not change
the requirements governing operation or availability of safety
equipment assumed to operate to preserve the margin of safety. The
change does not alter the behavior of plant equipment, which remains
unchanged. The available pressure range is expanded by the change,
thus offering greater margin for pressure reduction during the
transient.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
Based on the above, Entergy concludes that the proposed
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Benjamin G. Beasley.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: January 29, 2014. A publicly-available
version is in ADAMS under Accession No. ML14029A438.
Description of amendment request: The amendment would revise the
facility operating license and technical specifications to reflect
adoption of a new fire protection licensing basis which complies with
the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance
in NRC Regulatory Guide (RG) 1.205, Revision 1, ``Risk-Informed
Performance-Based Fire Protection for Existing Light-Water Nuclear
Power Plants,'' December 2009 (ADAMS Accession No. ML092730314). The
license amendment request follows Nuclear Energy Institute (NEI) 04-02,
Revision 2, ``Guidance for Implementing a Risk-Informed, Performance-
Based Fire Protection Program under 10 CFR 50.48(c),'' April 2008. The
submittal describes the methodology used to demonstrate compliance
with, and transition to, National Fire Protection Association (NFPA)
805, and includes regulatory evaluations, probabilistic risk
assessment, change evaluations, proposed modifications for non-
compliances, and supporting attachments.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1: The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated
Operation of Arkansas Nuclear One, Unit 1 (ANO-1) in accordance
with the proposed amendment does not result in a significant
increase in the probability or consequences of accidents previously
evaluated. The proposed amendment does not affect accident
initiators or precursors as described in the ANO-1 Safety Analysis
Report (SAR), nor does it adversely alter design assumptions,
conditions, or configurations of the facility, and it does not
adversely impact the ability of structures, systems, or components
(SSCs) to perform their intended function to mitigate the
consequences of accidents described and evaluated in the SAR. The
proposed changes do not physically alter safety-related systems nor
affect the way in which safety-related systems perform their
functions as required by the accident analysis. The SSCs required to
safely shut down the reactor and to maintain it in a safe shutdown
condition will remain capable of performing their design functions.
The purpose of this amendment is to permit ANO-1 to adopt a new
risk-informed, performance-based fire protection licensing basis
that complies with the requirements in 10 CFR 50.48(a) and 10 CFR
50.48(c), as well as the guidance contained in Regulatory Guide (RG)
1.205. The NRC considers that NFPA 805 provides an acceptable
methodology and performance criteria for licensees to identify fire
protection requirements that are an acceptable
[[Page 38590]]
alternative to the 10 CFR Part 50, Appendix R, fire protection
features (69 FR 33536; June 16, 2004).
The purpose of the fire protection program is to provide
assurance, through defense-in-depth, that the NRC's fire protection
objectives are satisfied. These objectives are: (1) preventing fires
from starting; (2) rapidly detecting and controlling fires and
promptly extinguishing those fires that do occur, thereby limiting
fire damage; (3) providing an adequate level of fire protection for
SSCs important to safety, so that a fire that is not promptly
extinguished will not prevent essential plant safety functions from
being performed; and (4) ensuring that fires will not significantly
increase the risk of radioactive releases to the environment. In
addition, fire protection systems must be designed such that their
failure or inadvertent operation does not adversely impact the
ability of the SSCs important to safety to perform their safety-
related functions.
NFPA 805, taken as a whole, provides an acceptable alternative
for satisfying General Design Criterion 3 (GDC 3) of Appendix A to
10 CFR Part 50, meets the underlying intent of the NRC's existing
fire protection regulations and guidance, and achieves defense-in-
depth along with the goals, performance objectives, and performance
criteria specified in NFPA 805, Chapter 1. In addition, if there are
any increases in core damage frequency (CDF) or risk as a result of
the transition to NFPA 805, the increase will be small, bounded by
the delta risk requirements of NFPA 805, and consistent with the
intent of the Commission's Safety Goal Policy.
Engineering analyses, which may include engineering evaluations,
probabilistic risk assessments, and fire modeling calculations, have
been performed to demonstrate that the performance-based
requirements of NFPA 805 have been met. The SAR documents the
analyses of design basis accidents (DBAs) at ANO-1. All accident
analysis acceptance criteria will continue to be met with the
proposed amendment. The proposed changes will not affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of any accident
previously evaluated. The proposed changes will not alter any
assumptions or change any mitigation actions for the radiological
consequence evaluations in the ANO-1 SAR. In addition, the
applicable radiological dose acceptance criteria will continue to be
met.
Based on the above, the implementation of this amendment to
transition the Fire Protection Plan (FPP) at ANO-1 to one based on
NFPA 805, in accordance with 10 CFR 50.48(c), does not result in a
significant increase in the probability of any accident previously
evaluated. In addition, all equipment required to mitigate an
accident remains capable of performing the assumed function.
Therefore, the consequences of any accident previously evaluated are
not significantly increased with the implementation of this
amendment.
Criterion 2: The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from Any Accident Previously
Evaluated
Operation of ANO-1 in accordance with the proposed amendment
does not create the possibility of a new or different kind of
accident from any accident previously evaluated. Previously analyzed
accidents with potential offsite dose consequences were included in
the evaluation of the transition to NFPA 805. The proposed amendment
does not impact these accident analyses. The proposed change does
not alter the requirements or functions for systems required during
accident conditions as assumed in the licensing basis analyses and/
or DBA radiological consequences evaluations.
Implementation of the new risk-informed, performance-based fire
protection licensing basis, which complies with the requirements in
10 CFR 50.48(a) and 10 CFR 50.48(c), as well as the guidance
contained in RG 1.205, will not result in new or different kinds of
accidents. The NRC considers that NFPA 805 provides an acceptable
methodology and performance criteria for licensees to identify fire
protection systems and features that are an acceptable alternative
to the 10 CFR 50, Appendix R fire protection features (69 FR 33536,
June 16, 2004). No new modes of operation are introduced by the
proposed amendment, nor will it create any failure mode not bounded
by previously evaluated accidents. Further, the impacts of the
proposed change are not directly assumed in any safety analysis to
initiate an accident sequence.
The requirements in NFPA 805 address only fire protection and
the impacts of fire effects on the plant have been evaluated. The
proposed fire protection program changes do not involve new failure
mechanisms or malfunctions that could initiate a new or different
kind of accident beyond those already analyzed in the SAR. Based on
this, as well as the discussion above, the implementation of this
amendment to transition the FPP at ANO-1 to one based on NFPA 805,
in accordance with 10 CFR 50.48(c), does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Criterion 3: The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
Operation of ANO-1 in accordance with the proposed amendment
does not involve a significant reduction in a margin of safety. The
transition to a new risk-informed, performance-based fire protection
licensing basis that complies with the requirements in 10 CFR
50.48(a) and 10 CFR 50.48(c) does not alter the manner in which
safety limits, limiting safety system settings, or limiting
conditions for operation are determined. The safety analysis
acceptance criteria are not affected by this change. The proposed
amendment does not adversely affect existing plant safety margins or
the reliability of equipment assumed in the SAR to mitigate
accidents. The proposed change does not adversely impact systems
that respond to safely shut down the plant and maintain the plant in
a safe shutdown condition. In addition, the proposed amendment will
not result in plant operation in a configuration outside the design
basis for an unacceptable period of time without implementation of
appropriate compensatory measures.
The risk evaluations for plant changes, in part as they relate
to the potential for reducing a safety margin, were measured
quantitatively for acceptability using the delta risk (i.e.,
[Delta]CDF and [Delta]LERF [large early release frequency]) criteria
from Section 5.3.5, ``Acceptance Criteria,'' of NEI 04-02, as well
as the guidance contained in RG 1.205. Engineering analyses, which
may include engineering evaluations, probabilistic safety
assessments, and fire modeling calculations, have been performed to
demonstrate that the performance-based methods of NFPA 805 do not
result in a significant reduction in the margin of safety. As such,
the proposed changes are evaluated to ensure that risk and safety
margins are kept within acceptable limits. Based on the above, the
implementation of this amendment to transition the FPP at ANO-1 to
one based on NFPA 805, in accordance with 10 CFR 50.48(c), will not
significantly reduce a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, LA 70113.
NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: April 30, 2014. A publicly-available
version is in ADAMS under Accession No. ML14127A435.
Description of amendment request: The proposed amendment would
revise Oyster Creek Nuclear Generating Station (OCNGS) Technical
Specification (TS) 4.5 M., ``Shock Suppressors (Snubbers),'' to conform
the TS to the revised OCNGS Snubber Inspection Program.
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 changes would revise TS 4.5.M to conform the TS to
the revised Snubber Inspection Program. Snubber examination, testing
and service life
[[Page 38591]]
monitoring will continue to meet the requirements of 10 CFR
50.55a(g). Snubber examination, testing and service life monitoring
is not an initiator of any accident previously evaluated.
Therefore, the probability of an accident previously evaluated
is not significantly increased.
Snubbers will continue to be demonstrated OPERABLE by
performance of a program for examination, testing and service life
monitoring in compliance with 10 CFR 50.55a or authorized
alternatives. The proposed changes do not adversely affect plant
operations, design functions or analyses that verify the capability
of systems, structures, and components to perform their design
functions.
Therefore, the consequences of accidents previously evaluated
are not significantly increased.
Based on the above, these proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the amendment change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve any physical alteration of
plant equipment. The proposed changes do not alter the method by
which any safety-related system performs its function. As such, no
new or different types of equipment will be installed, and the basic
operation of installed equipment is unchanged. The methods governing
plant operation and testing remain consistent with current safety
analysis assumptions.
Therefore, it is concluded that these 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 proposed changes ensure snubber examination, testing and
service life monitoring will continue to meet the requirements of 10
CFR 50.55a(g). Snubbers will continue to be demonstrated OPERABLE by
performance of a program for examination, testing and service life
monitoring in compliance with 10 CFR 50.55a or authorized
alternatives.
Therefore, it is concluded that 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: J. Bradley Fewell, VP & Deputy General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Meena Khanna.
Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: April 9, 2014. A publicly-available
version is in ADAMS under Accession No. ML14101A367.
Description of amendment request: The proposed amendment would
revise the Donald C. Cook Nuclear Plant, Units 1 and 2, technical
specification (TS) 3.4.2, ``[Reactor Coolant System (RCS)] Pressure and
Temperature (P/T) Limits,'' to address an issue regarding the
applicability of TS Figures 3.4.3-1 ``Reactor Coolant System Pressure
versus Temperature Limits--Heatup Limit, Criticality Limit, and Leak
Test Limit (Applicable for service period up to 32 [Effective Full
Power Years (EPFY)]'' and 3.4.3-2 ``Reactor Coolant System Pressure
versus Temperature Limits--Various Cooldown Rates Limits (Applicable
for service period up to 32 EFPY)'' during vacuum fill operations of
the RCS.
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 TS changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
There are no physical changes to the plant being introduced by the
proposed changes to the heatup and cooldown limitation curves. The
proposed changes do not modify the RCS pressure boundary. That is,
there are no changes in operating pressure, materials, or seismic
loading. The proposed changes do not adversely affect the integrity
of the RCS pressure boundary such that its function in the control
of radiological consequences is affected.
Therefore, it is concluded that the proposed amendment does not
involve a significant increase in the probability or the
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 TS changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated. No new modes of operation are introduced by the proposed
changes. The proposed changes will not create any failure mode not
bounded by previously evaluated accidents. Further, the proposed
changes to the heatup and cooldown limitation curves do not affect
any activities or equipment other than the RCS pressure boundary and
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
Consequently, 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 change involve a significant reduction in a
margin of safety?
Response: No.
The proposed TS changes do not involve a significant reduction
in the margin of safety. The revised heatup and cooldown limitation
curves and low-temperature overpressure protection limits are
established in accordance with current regulations and the [American
Society of Mechanical Engineers Boiler and Pressure Vessel (ASME
B&PV)] Code 1995 edition with 1996 Addenda. These proposed changes
are acceptable because the ASME B&PV Code maintains the margin of
safety required by [Title 10 of the Code of Federal Regulations (10
CFR)] 50.55(a). Because operation will be within these limits, the
RCS materials will continue to behave in a non-brittle manner
consistent with the original design bases.
Therefore, the proposed amendment does not involve a significant
reduction in 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: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
NRC Branch Chief: Robert D. Carlson.
Northern States Power Company--Minnesota (NSPM), Docket No. 50-263,
Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota
Date of amendment request: October 4, 2013, as supplemented by
letter dated April 29, 2014. Publicly-available versions are in ADAMS
under Accession Nos. ML13281A826 and ML14122A044, respectively.
Description of amendment request: Following completion of an on-
site staffing analysis of the Emergency Response Organization, NSPM
determined that the Radwaste Operator is no longer required to augment
plant staff for performing repairs and corrective actions as prescribed
in the MNGP Emergency Plan. The amendment proposes to remove the
Radwaste Operator position as a 60-minute responder credited within the
MNGP Emergency 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
[[Page 38592]]
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
function of plant structures, systems, or components (SSCs). The
proposed change does not affect accident initiators or precursors,
nor does it alter design assumptions. The proposed change does not
alter or prevent the ability of the Emergency Response Organization
to perform their intended functions to mitigate the consequences of
an accident or event. This proposed change only removes a no longer
credited position from the Emergency Plan.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not 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 does not introduce failure modes that could result in a new
accident, and the change does not alter assumptions made in the
safety analysis. This proposed change only removes a no longer
credited position from the Emergency Plan. The proposed change,
therefore, does not alter or prevent the ability of the Emergency
Response Organization to perform their intended functions to
mitigate the consequences of an accident or event.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed change is
associated with the Emergency Plan staffing and does not impact
operation of the plant or its response to transients or accidents.
The change does not affect the Technical Specifications. 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. Safety analysis acceptance criteria are not affected by this
proposed change. The revised Emergency Plan will continue to provide
the necessary response staff with the proposed change.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Robert D. Carlson.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: February 10, 2014, as supplemented by
letter dated June 9, 2014. Publicly-available versions are in ADAMS
under Accession Nos. ML14041A408 and ML14163A417, respectively.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) surveillance frequency for the
pressurizer safety valves from a refueling frequency (i.e., 18 months
+25 percent) to be consistent with the Inservice Testing Program. In
addition, the proposed amendment would administratively change the
format of the footnotes in TS Table 3-5, ``Minimum Frequencies for
Equipment Tests.''
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 requested change revises the performance interval of one TS
surveillance requirement to be consistent with the Inservice Testing
Program as stated in 10 CFR 50.55a(g)(5). The performance of the
surveillance, or the failure to perform the surveillance, is not a
precursor to an accident. Performing the surveillance or failing to
perform the surveillance does not affect the probability of an
accident. Even with the requested extension, the period during which
the plant is in Modes 1 or 2 and the valves are required to be
operable will be no longer than a typical operating cycle. Also, the
proposed interval between tests will be consistent with the interval
for this type of valve specified by the American Society of
Mechanical Engineers (ASME) Code for Operation and Maintenance of
Nuclear Power Plants (OM Code), 1998 Edition, through 2000 Addenda,
Appendix I, frequency requirements for testing of pressure relief
valves.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not alter the physical design, safety
limits, or safety analysis assumptions associated with the operation
of the plant. Hence, the proposed change does not introduce any new
accident initiators, nor does it reduce or adversely affect the
capabilities of any plant structure or system in the performance of
their safety function.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change revises the performance interval for one
surveillance requirement to be consistent with the test interval for
this type of valve specified by the ASME OM Code, 1998 Edition,
through 2000 Addenda as required by 10 CFR 50.55a. This change does
not alter any safety margins.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: March 31, 2014. A publicly-available
version is in ADAMS under Accession No. ML14090A417.
Description of amendment request: The proposed amendment would
change Technical Specification 2.5, Auxiliary Feedwater (AFW) system to
allow a 7-day completion time for the turbine-driven AFW pump if the
inoperability occurs following a refueling outage and if MODE 2 had not
been entered.
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
[[Page 38593]]
consequences of an accident previously evaluated?
Response: No.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change to Technical Specification (TS) 2.5 would
allow a seven day Completion Time for the turbine-driven Auxiliary
Feedwater (AFW) pump if the inoperability occurs following a
refueling outage, and if MODE 2 had not been entered. The note
currently in TS 2.5 Applicability addresses the issue of allowing
additional time to perform necessary testing to prove the
operability of the turbine driven AFW pump following refueling as
approved by the NRC in TS Amendment 127. This note does not
specifically state that it is only allowed following refueling and
does not restrict the time the plant can be in this condition. The
proposed change will be more restrictive than the current TS since
it will specifically state when it is allowed (following refueling)
and for how long it is allowed.
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated
because: 1) the proposed amendment does not represent a change to
the system design, 2) the proposed amendment does not prevent the
safety function of the AFW system from being performed, since the
other fully redundant essential train is required to be operable, 3)
the proposed amendment does not alter, degrade, or prevent action
described or assumed in any accident Updated Safety Analysis Report
(USAR) from being performed since the other train of AFW is required
to be operable, 4) the proposed amendment does not alter any
assumptions previously made in evaluating radiological consequences,
and 5) the proposed amendment does not affect the integrity of any
fission product barrier. No other safety related equipment is
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 amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not alter the physical design, safety
limits, or safety analysis assumptions associated with the operation
of the plant. Hence, the proposed change does not introduce any new
accident initiators, nor does it reduce or adversely affect the
capabilities of any plant structure or system in the performance of
their safety function.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not involve a significant reduction in a
margin of safety. The proposed change to TS 2.5 would restrict for
the turbine-driven AFW pump inoperability to a seven day Completion
Time if the inoperability occurs following a refueling outage and
prior to MODE 2 being entered. The current Note in TS 2.5
Applicability does not require the turbine driven AFW pump to be
operable until prior to entering MODE 2; therefore, the proposed
change is more restrictive than current TS.
The proposed change does not involve a significant reduction in
a margin of safety because: (1) during a return to power operations
following a refueling outage, decay heat is at its lowest levels,
(2) the other AFW train is required to be operable, and (3) the
motor-driven AFW train can provide sufficient flow to remove decay
heat and cool the unit to shutdown cooling system entry conditions
from power operations. This change does not alter any safety
margins.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: April 25, 2014. A publicly-available
version is in ADAMS under Accession No. ML14118A435.
Description of amendment request: The proposed amendment would
revise Section 5.11, ``Structures Other Than Containment,'' and
Appendix F, ``Classification of Structures and Equipment and Seismic
Criteria,'' of the Fort Calhoun Station, Unit No. 1, Updated Safety
Analysis Report. The changes would clarify the licensing and design
basis to permit the use of seismic floor response spectra in analysis
and design of seismic Class I structures and structural elements
attached to structures.
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.
[T]his change to the Updated Safety Analysis Report (USAR) has
no effect on the consequences of any accident, as it makes no
physical changes to the plant. Since the Alternate Seismic Criteria
and Methodologies (ASCM) floor response spectra (FRS) represent a
refined version of the plant's original design basis, the design
margins for any application utilizing the FRS will be maintained
with respect to the design basis earthquake. Thus, the proposed
amendment does not result in 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.
[T]he change to the USAR does not change any accident analyses,
does not make any physical changes to the plant, and does not change
the way the plant is operated. The only change is to permit the
utilization of the ASCM curves in the design and evaluation of
structural applications. The curves themselves are based on the same
earthquake as the plant's original design. Thus, 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.
[T]he ASCM FRS is based on the same earthquake as the plant's
original design basis. The ASCM FRS are refined curves of the same
design basis and thus, the design margins of any application or
evaluation utilizing the ASCM FRS will be maintained with respect to
the design basis earthquake. Thus, 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: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: May 16, 2014. A publicly-available
version is in ADAMS under Accession No. ML14143A370.
Description of amendment request: The proposed amendment would
revise the Updated Safety Analysis Report (USAR) to allow pipe stress
analysis of non-reactor coolant system safety-related piping to be
performed in accordance with the American Society
[[Page 38594]]
of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,
Section III, 1980 Edition as an alternative to current Code of record.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the current licensing basis (CLB) allows
the use of American Society of Mechanical Engineers (ASME) Boiler
and Pressure Vessel (BPV) Code, Section III, 1980 Edition (no
Addenda) as an alternative to the original Code of Record (i.e.,
United States of America Standards (USAS) B31.7 1968 (DRAFT)
Edition) for the design and analysis of non-reactor coolant system
(RCS) piping. The American National Standards Institute (ANSI) B31
Code Committee has determined that:
``. . . piping that has been designed and constructed in
accordance with Section III of the ASME Boiler and Pressure Vessel
Code including addenda and applicable cases may be accepted as
complying with the requirements of B31.7, 1969 and applicable
addenda for the respective class of construction.''
Although the ANSI B31 Code Committee statement refers to the
B31.7, 1969 Edition, there are no significant differences between it
and the B31.7 1968 (DRAFT) Edition. The change involves the
substitution of one accepted piping Code for another and not a
physical plant change. The Updated Safety Analysis Report (USAR)
accident analysis assumes the proper functioning of safety systems
in demonstrating the adequacy of the plant's design. This change
does not alter the intended function of any plant equipment nor does
it degrade or increase challenges to the performance of safety
systems assumed to function in the accident analysis.
The use of ASME BPV Code, Section III, 1980 Edition (no Addenda)
analytical methods provides acceptable design results with no
reduction in radiological barrier safety margin. Hence, there is no
change in radiological barrier performance that would increase the
dose to personnel onsite (10 CFR 20) or to the public at the site
boundary (10 CFR 100).
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated in the USAR.
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 provides the basis for the use of ASME
BPV Code, Section III, 1980 Edition (no Addenda) for stress analysis
of non-RCS safety-related piping. This approach will not introduce
any methods or analytical techniques that could create the
possibility of a new or different kind of accident. Application of a
Code methodology does not create the possibility of a different kind
of accident.
The application of the ASME BPV Code, Section III, 1980 Edition
(no Addenda) does not create any new unanalyzed interactions between
systems or components. Piping systems will be analyzed in accordance
with the Code, which is one part of the framework to establish the
necessary design, fabrication, construction, testing, and
performance requirements for structures, systems, and components
important to safety. The proposed change to the CLB does not create
a new failure mechanism or new accident initiator. The proposed
amendment does not involve a change in methods governing the
operation of plant systems or components.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated in the USAR.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The Fort Calhoun Station Technical Specifications (TS) ensure
that the plant operates in a manner that will ensure acceptable
levels of protection for the health and safety of the public. The
Technical Specifications ensure that the available equipment and
initial conditions for a Design Basis Accident (DBA) as defined in
the USAR meet the assumptions in the accident analysis contained in
the USAR. The plant safety margins are addressed in the Technical
Specification Bases and the USAR.
This proposed amendment revises the CLB to allow the use of ASME
BPV Code, Section III, 1980 Edition (no Addenda) for stress analysis
of non-RCS safety-related piping. No changes are being made to the
physical plant. The use of the ASME BPV Code, Section III, 1980
Edition (no Addenda) does not change, revise, or otherwise affect
the current Technical Specifications (TS) or TS Bases. Incorporation
of the ASME BPV Code, Section III, 1980 Edition (no Addenda) into
the FCS CLB will not affect the current plant design parameters or
TS Limiting Conditions for Operation (LCO).
The proposed change does not modify, change, revise, or
otherwise affect any current calculations concerning the plant
accident analysis or supporting basis for which the TSs, TS Bases,
or USAR safety margins were established. 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: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
ZionSolutions LLC, Docket Nos. 50-295 and 50-304, Zion Nuclear Power
Station (ZNPS), Units 1 and 2, Lake County, Illinois
Date of amendment request: March 17, 2014. A publicly-available
version is in ADAMS under Accession No. ML14078A049.
Description of amendment request: The proposed amendments would
amend licenses DPR-39 and DPR-48 and revise the Zion Technical
Specifications (TS) to reflect the removal of all the spent fuel from
the Zion spent fuel pool. The proposed changes to both Facility
Operating Licenses modify Section 2.C.(6) to specify the ZNPS
Independent Spent Fuel Storage Installation Physical Security Plan
(ISFSI), eliminate Section 2.C.(7) Spent Fuel Pool Modification, and
eliminate Section 2.C.(16), related to the single-failure proof fuel
building crane. The proposed changes to the TS eliminate provisions of
the specifications applicable to spent fuel stored in the spent fuel
pool and relocate the remaining TS administrative requirements to the
Quality Assurance Project Plan. These changes are proposed pursuant to
the criteria contained in 10 CFR 50.36 and in accordance with the
recommendations contained in the U.S. Nuclear Regulatory Commission's
(NRC) Administrative Letter 95-06. The proposed changes will result in
a TS that will be applicable to the ZNPS once the last spent fuel
assembly has been removed from the spent fuel pool and placed at the
ISFSI.
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?
The proposed changes (deletion of operational requirements and
certain design requirements) reflect the complete transfer of the
spent fuel from the spent fuel pool to the ISFSI. Design basis
accidents related to the spent fuel pool are discussed in the ZNPS
Defueled Safety Analysis Report (DSAR) Chapter 5. These postulated
accidents are predicated on spent fuel being stored in the spent
fuel pool. With the removal of the spent fuel from the spent fuel
pool, there are no remaining spent fuel assemblies to be monitored
and there are no credible accidents that require the actions of a
Certified Fuel Handler, Shift Supervisor, or a Non-certified
Operator to prevent occurrence or mitigate the consequences of an
accident.
In addition, the ZNPS DSAR Chapter 5 also provides analyses of
accidents as result of
[[Page 38595]]
decommissioning with the bounding consequences resulting from the
failure of a High Integrity Container (HIC) containing dewatered
radioactive demineralizer resin.
The proposed changes do not have an adverse impact on the
remaining decommissioning activities or any decommissioning related
postulated accident consequences.
The proposed changes related to the relocation of certain
administrative requirements do not affect operating procedures or
administrative controls that have the function of preventing or
mitigating any remaining decommissioning design basis accidents. In
addition, these proposed changes are consistent with the guidance of
the NRC's Administrative Letter 95-06.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
(2) Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes eliminate the operational requirements and
certain design requirements associated with the storage of the spent
fuel in the spent fuel pool, and relocate certain administrative
controls to the Quality Assurance Program Plan.
With the complete removal of the spent fuel from the spent fuel
pool and transfer to the ISFSI, there are no spent fuel assemblies
that remain at the plant and the potential for fuel related
accidents is removed. The proposed changes do not introduce any new
failure modes. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) Does the change involve a significant reduction in a margin
of safety?
The design basis and accident assumptions within the ZNPS DSAR
and the TS relating to spent fuel are no longer applicable. The
proposed changes do not affect remaining plant operations, systems,
or components supporting decommissioning activities. In addition,
the proposed changes do not result in a change in initial
conditions, system response time, or in any other parameter
affecting the course of the remaining decommissioning activity
accident analysis. 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: Russ Workman, Deputy General Counsel,
EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT
84101.
NRC Branch Chief: Bruce Watson.
III. 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 can be accessed as described in the
``Accessing Information and Submitting Comments'' section of this
document.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendment: December 26, 2012, as
supplemented by letter dated August 26, 2013.
Brief description of amendment: The amendments adopt Technical
Specifications Task Force (TSTF) change traveler TSTF-500, Revision 2,
``DC Electrical Rewrite--Update to TSTF-360.'' The amendments revised
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.'' A new ``Battery Monitoring and Maintenance
Program'' was added to Section 5.5, ``Programs and Manuals.''
Date of issuance: June 25, 2014.
Effective date: As of the date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment No.: Unit 1--193; Unit 2--193; Unit 3--193. A publicly-
available version is in ADAMS under Accession No. ML14115A045;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74:
The amendments revised the Operating Licenses and Technical
Specifications.
Date of initial notice in Federal Register: March 4, 2013 (78 FR
14129). The supplement dated August 26, 2013, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 25, 2014.
No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc. (DEK), Docket No. 50-305, Kewaunee Power
Station (KPS), Kewaunee County, Wisconsin
Date of application for amendment request: April 16, 2013, as
supplemented by letters dated September 5, 2013, October 14, 2013, and
March 19, 2014.
Brief description of amendment: The amendment revised the Renewed
Facility Operating License by deleting a license condition associated
with license renewal and adding a license condition related to spent
fuel pool storage rack boron absorber surveillance.
Date of issuance: June 23, 2014.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 213. A publicly-available version is in ADAMS under
Accession No. ML14008A297; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-43: Amendment revised
the Renewed Facility Operating License.
[[Page 38596]]
Date of initial notice in Federal Register: August 20, 2013 (78 FR
51223). The supplemental letters dated September 5, 2013, October 14,
2013, and March 19, 2014, 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 June 23, 2014.
No significant hazards consideration comments received: No.
Dominion Energy Kewaunee, Inc. (DEK), Docket No. 50-305, Kewaunee Power
Station (KPS), Kewaunee County, Wisconsin
Date of application for amendment request: May 29, 2013, as
supplemented by letters dated September 23, October 15, October 17,
October 31, and November 7, 2013, and letters dated January 7, 2014,
and March 13, 2014.
Brief description of amendment: The amendment revised the Renewed
Facility Operating License Technical Specifications (TSs) to permit
fuel handling activities consistent with the permanently shutdown and
defueled condition of the facility. Specifically, in its March 13,
2014, supplemental letter DEK stated that it had accelerated the
schedule to transfer spent fuel from the spent fuel pool to the
independent spent fuel storage installation (ISFSI). Under its new
schedule, DEK plans to begin activities to support spent fuel transfer
to the ISFSI by July 1, 2014. Based on its new schedule, DEK requested
expedited review and partial approval of the deletion of certain TSs
currently required for movement of irradiated fuel assemblies. If not
amended, the affected TSs would require restoring operability of
certain equipment during spent fuel handling activities that are no
longer needed for accident mitigation.
The NRC staff has issued a partial approval of the original May 29,
2013, amendment request as supplemented, to permit fuel handling
activities in accordance with DEK's request in its March 13, 2014,
submittal. The staff continues to review the remaining license
condition and technical specification changes requested in DEK's May
29, 2013, submittal as supplemented, that were not addressed in this
amendment.
Date of issuance: June 9, 2014.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 212. A publicly-available version is in ADAMS under
Accession No. ML14111A234; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-43: Amendment revised
the Renewed Facility Operating License.
Date of initial notice in Federal Register: August 20, 2013 (78 FR
51224). The supplemental letters dated September 23, October 15,
October 17, October 31, and November 7, 2013, January 7, 2014, and
March 13, 2014, 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 June 9, 2014.
No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: February 7, 2013, as supplemented by
letter dated January 16, 2014.
Brief description of amendment: The amendment revised the River
Bend Station, Unit 1 (RBS) Technical Specification (TS) 3.8.4, ``DC
[Direct Current] Sources--Operating,'' Surveillance Requirements
3.8.4.2 and 3.8.4.5. The change is the result of the licensee's
determination that the total battery capacity would possibly be
insufficient to supply the required load to the DC system if each of
the battery-to-battery connections were to reach the individual
resistance limits. The changes to the Surveillance Requirements added
new acceptance criteria to address the possible non-conservative
conditions when the battery connection resistances are at maximum TS
values.
Date of issuance: June 18, 2014.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 181. A publicly-available version is in ADAMS under
Accession No. ML14136A008; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 30, 2013 (78 FR
25312). The supplemental letter dated January 16, 2014, 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 June 18, 2014.
No significant hazards consideration comments received: No.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of application for amendment: September 14, 2012, as
supplemented by letters dated January 29, February 14, May 30, and
October 22, 2013, and March 11, 2014.
Brief description of amendment: The amendments revised the
operating licenses and Technical Specifications (TSs) to remove
completed and satisfied license conditions, revised TS 5.5.1 to remove
related conditions, corrected inadvertent errors, updated references to
the Physical Security Plan, and made editorial changes to the operating
licenses and TSs.
Date of issuance: June 13, 2014.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 260 and 255. A publicly-available version is in
ADAMS under Accession No. ML13329A092; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the licenses and the TSs.
Date of initial notice in Federal Register: January 8, 2013 (78 FR
1271), and April 16, 2013 78 FR 22569). The submittal dated January 29,
2013, expanded the scope of the application dated September 14, 2012,
and the application was renoticed April 16, 2013. The supplements dated
February 14, May 30, and October 22, 2013, and March 11, 2014, provided
additional information that clarified the application, did not expand
the scope of the submittal dated January 29, 2013, as noticed, and did
not change the staff's proposed no significant hazards consideration
determinations published on January 8, 2013, and April 16, 2013. The
supplement dated March 11, 2014, limited the scope of the supplement
dated January 29, 2013, by deleting the
[[Page 38597]]
proposed change to TS Figure 3.1-2, ``Boric Acid Tank Minimum Volume.''
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 13, 2014.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of application for amendment: September 28, 2011, as
supplemented by letters dated December 19 and December 22, 2011; March
20, July 24, August 24, and September 27, 2012; April 23, May 21, July
29, September 12, October 11, November 4, November 11, and December 18,
2013; and January 24, February 28, April 10, and June 11, 2014.
Brief description of amendment: The amendment transitions the Fort
Calhoun Station fire protection program to a risk-informed,
performance-based program based on National Fire Protection Association
(NFPA) 805, in accordance with 10 CFR 50.48(c). NFPA 805 allows the use
of performance-based methods such as fire modeling and risk-informed
methods such as fire probabilistic risk assessment to demonstrate
compliance with the nuclear safety performance criteria.
Date of issuance: June 16, 2014.
Effective date: As of its date of issuance and shall be implemented
by 12 months from the date of issuance.
Amendment No.: 275. A publicly-available version is in ADAMS under
Accession No. ML14098A092; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 10, 2012 (77 FR
21598). The supplements dated March 20, July 24, August 24, and
September 27, 2012; April 23, May 21, July 29, September 12, October
11, November 4, November 11, and December 18, 2013; and January 24,
February 28, April 10, and June 11, 2014, 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 June 16, 2014.
No significant hazards consideration comments received: No.
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: December 20, 2013.
Brief description of amendment: The amendment revises the plant's
emergency plan. In conjunction with the new license condition, the
amendment complies with the established regulatory changes set forth in
``Enhancements to Emergency Preparedness Regulations,'' published in
the Federal Register on November 23, 2011 (76 FR 72560). Specifically,
the license amendment changes on-shift staffing analysis and the
changes to the emergency plan address evacuation time estimates. The
design, construction and operation of the plant are not affected by
this license amendment and license condition.
Date of issuance: May 30, 2014.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 20. A publicly-available version is in ADAMS under
Accession No. ML14118A252; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised
the Facility Combined Licenses.
Date of initial notice in Federal Register: February 4, 2014, (79
FR 6643).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 30, 2014.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: December 13, 2012, as
supplemented by letters dated June 11, 2013, and January 16 and April
9, 2014.
Brief description of amendment: The amendment revised 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 is added to verify that the
UHS cooling tower fans respond appropriately to automatic start
signals.
Date of issuance: June 17, 2014.
Effective date: As of its date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment No.: 208. A publicly-available version is in ADAMS under
Accession No. ML14149A164; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 4, 2013 (78 FR
14138). The supplements dated June 11, 2013, and January 16 and April
9, 2014, 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 June 17, 2014.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 27th day of June, 2014.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2014-15770 Filed 7-7-14; 8:45 am]
BILLING CODE 7590-01-P