[Federal Register Volume 80, Number 51 (Tuesday, March 17, 2015)]
[Notices]
[Pages 13902-13920]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-05994]
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NUCLEAR REGULATORY COMMISSION
[NRC-2015-0055]
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 February 19, 2015 to March 4, 2015. The last
[[Page 13903]]
biweekly notice was published on March 3, 2015.
DATES: Comments must be filed by April 16, 2015. A request for a
hearing must be filed by May 18, 2015.
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-2015-0055. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected].
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: OWFN-12-H08, 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: Beverly A. Clayton, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-3475, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2015-0055 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-2015-0055.
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-2015-0055, facility name, unit
number(s), application date, and subject in your comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS. 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 submissions 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 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
[[Page 13904]]
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 unless the
Commission finds an imminent danger to the health or safety of the
public, in which case it will issue an appropriate order or rule under
10 CFR part 2.
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.
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
[[Page 13905]]
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 accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal
River, Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida
Date of amendment request: November 7, 2014. A publicly-available
version is in ADAMS under Accession No. ML14321A450.
Description of amendment request: The amendment would reflect the
transfer of ownership, held by eight minority co-owners, in CR-3 to
DEF. The transfer of ownership will take place pursuant to the
Settlement, Release and Acquisition Agreement, dated September 26,
2014, wherein DEF will purchase the 6.52 percent combined ownership
share in CR-3 held by these minority co-owners, leaving DEF and
Seminole Electric Cooperative, Inc., as the remaining licensees for CR-
3.
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 do not involve a significant increase in
the probability of any accident previously evaluated because no
accident initiators or assumptions are affected. The proposed
license transfers are administrative in nature and have no direct
effect on any plant system, plant personnel qualifications, or the
operation and maintenance of CR-3.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated because no
new accident initiators or assumptions are introduced by the
proposed changes. The proposed license transfers are administrative
in nature and have no direct effect on any plant system, plant
personnel qualifications, or operation and maintenance of CR-3.
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 because the proposed changes do not involve changes
to the initial conditions contributing to accident severity or
consequences, or reduce response or mitigation capabilities. The
proposed license transfers are administrative in nature and have no
direct effect on any plant system, plant personnel qualifications,
or operation and maintenance of CR-3.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, 550 South Tryon Street,
Charlotte NC 28202.
NRC Branch Chief: Douglas A. Broaddus.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating, Unit 2, Westchester County, New York
Date of amendment request: December 9, 2014. A publicly-available
version is in ADAMS under Accession No. ML14353A015.
Description of amendment request: The amendment would revise
Technical Specification 5.5.14, ``Containment Leakage Rate Testing
Program,'' to extend the frequency of the Containment Integrated Leak
Rate Test or Type A Test from once every 10 years to once every 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 IP2 [Indian Point
Unit No. 2] 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
[[Page 13906]]
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 accepted guidelines of NEI
94-01, Revision 2A, for development of the IP2 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 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 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 2A, for the development of the IP2 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?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 2A, for the development of the IP2 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 containment leakage tests would
be performed at the frequencies established in accordance with the
NRC-accepted guidelines of NEI 94-01, Revision 2A with no change to
the 60 month frequencies of Type B, and Type C tests.
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 IP2 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.
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2 (NMP2), Oswego County, New York
Date of amendment request: November 19, 2014. A publicly-available
version is in ADAMS under Accession No. ML14329A353.
Description of amendment request: The proposed amendment would
modify the Nine Mile Point (NMP) Nuclear Station, Unit 2 Technical
Specifications (TS) by relocating specific surveillance frequencies to
a licensee-controlled program with the adoption of Technical
Specification Task Force (TSTF)-425, Revision 3, ``Relocate
Surveillance Frequencies to Licensee Control--Risk Informed Technical
Specification Task Force (RITSTF) Initiative 5b.'' The licensee's
application dated November 19, 2014, Attachment 1, section 2.2, has
identified some variations or deviations from the TSTF-425.
Additionally, the change would add a new program, the Surveillance
Frequency Control Program, to TS section 5, Administrative Controls.
The NRC staff issued a notice of opportunity for comment in the Federal
Register on December 5, 2008, 73 FR 74202, on possible amendments to
revise the plant specific TS, to Relocate Surveillance Frequencies to
Licensee Control--RITSTF Initiative 5b. The Notice included a model
safety evaluation and model No Significant Hazards Consideration (NSHC)
determination, using the consolidated line-item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on July 6, 2009 (74 FR 31996). The licensee affirmed the
applicability of the model NSHC determination in its application dated
November 19, 2014, which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes relocate the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program (SFCP). Surveillance
frequencies are not an initiator to any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The systems and components
required by the technical specifications for which the surveillance
frequencies are relocated are still required to be operable, meet
the acceptance criteria for the surveillance requirements, and be
capable of performing any mitigation function assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
changes. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of
[[Page 13907]]
equipment will be installed) or a change in the methods governing
normal plant operation. In addition, the changes do not impose any
new or different requirements. The changes do not alter assumptions
made in the safety analysis. The proposed changes are consistent
with the safety analysis assumptions and current plant operating
practice.
Therefore, the proposed 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 design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the final safety analysis report and bases to TS),
since these are not affected by changes to the surveillance
frequencies. Similarly, there is no impact to safety analysis
acceptance criteria as described in plant licensing basis. To
evaluate a change in the relocated surveillance frequency, Exelon
will perform a probabilistic risk evaluation using the guidance
contained in NRC approved NEI 04-10, Rev. 1 in accordance with the
TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable
acceptance guidelines and methods for evaluating the risk increase
of proposed changes to surveillance frequencies consistent with
Regulatory Guide 1.177.
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: J. Bradley Fewell, Senior Vice President,
Regulatory Affairs, Nuclear, and General Counsel, Exelon Generation
Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company LLC (), Docket Nos. STN 50-456 and STN 50-
457, Braidwood Station, Units 1 and 2, Will County, Illinois
Docket Nos. STN 50-454 and STN 50-455, Byron Station, Units 1 and 2,
Ogle County, Illinois
Date of amendment request: December 18, 2014. A publicly-available
version is in ADAMS under Accession No. ML14352A204.
Description of amendment request: The proposed amendment would
increase the voltage limit for the diesel generator (DG) full load
rejection test specified by technical specification (TS) Surveillance
Requirement (SR) 3.8.1.10. Additionally, the proposed amendment would
add Note 3 to TS SR 3.8.1.10 for alignment with the Standard Technical
Specifications documented in NUREG-1431, April 2012 (ADAMS Accession
No. ML12100A222).
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:
EGC [Exelon Generation Company] has evaluated the proposed
change for Braidwood Station and Byron Station, using the criteria
in 10 CFR 50.92, and has determined that the proposed change does
not involve a significant hazards consideration. The following
information is provided to support a finding of no significant
hazards consideration.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The DGs design function is to mitigate an accident and there are
no analyzed scenarios where the DGs are initiators of any previously
evaluated accident. Since DGs do not initiate accidents, this change
does not increase the probability of occurrence of a previously
evaluated accident. The proposed change to the testing approach of
the DGs is consistent with the original design of the DGs. The
proposed change is in accordance with RG [Regulatory Guide] 1.9
Revision 3, and this change to the testing approach does not impact
the DGs ability to mitigate accidents. The DGs will continue to
operate within the parameters and conditions assumed within the
accident analysis. This change does not result in an increase in the
likelihood of malfunction of the DGs or their supported equipment.
Since the DGs will continue to perform its required function, there
is no increase in the consequences of previously evaluated
accidents.
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 amendment does not change the DGs operation or
ability to perform its design function. The proposed change to TS SR
3.8.1.10 at increased voltage will ensure the DGs ability to perform
at rated power factor while meeting its requirements. The change to
TS SR 3.8.1.10 does not result in DG operation that would create a
new failure mode of the DGs that could create a new initiator of an
accident. This is because the DGs ability to perform its design
function is maintained in the same manner as originally designed.
The proposed change does not change the single failure capabilities
of the electrical power system or create a potential for loss of
power since the design operation of the DGs is maintained.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the
plant structures, systems, and components, the parameters within
which the plant is operated, and the setpoints for the actuation of
equipment relied upon to respond to an event. The proposed change
does not modify the safety limits or setpoints at which protective
actions are initiated. The proposed change increases the voltage
limit for the DG full load rejection test which results in new test
acceptance criterion that is more restrictive than the existing
acceptance criteria. The proposed change ensures the availability
and operability of safety-related DGs.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above evaluation, EGC concludes that the proposed
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92, paragraph (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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Travis L. Tate.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant (PNPP), Unit 1, Perry, Ohio
Date of amendment request: November 24, 2014. A publicly-available
version is in ADAMS under Accession No. ML14328A665.
Description of amendment request: The proposed amendment is
intended to revise the battery capacity testing surveillance
requirements in the technical specifications to reflect test
requirements when the battery is near end of life.
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:
[[Page 13908]]
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not change the design function of
the Class 1 E divisional battery systems and does not change the way
the plant is maintained or operated when performing battery
surveillance testing. The proposed amendment does not affect any
accident mitigating feature or increase the likelihood of
malfunction for plant structures, systems and components.
The proposed amendment does not affect the operability
requirements of the Class 1 E divisional battery systems.
Verification of operating the plant within prescribed limits will
continue to be performed, as currently required. Compliance with and
continued verification of the prescribed limits support the
capability of the Class 1 E divisional battery systems to perform
their required design functions during all plant operating,
accident, and station blackout conditions, consistent with the plant
safety analyses.
The proposed amendment will not change any of the analyses
associated with the PNPP Updated Safety Analysis Report Chapter 15
accidents because plant operation, plant structures, systems,
components, accident initiators, and accident mitigation functions
remain unchanged.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not change the design function of
the Class 1 E divisional battery systems, and does not change the
way the plant is operated or maintained. The proposed amendment does
not create a credible failure mechanism, malfunction or accident
initiator not already considered in the design and licensing basis.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Safety margins are applied to design and licensing basis
functions and to the controlling values of parameters to account for
various uncertainties and to avoid exceeding regulatory or licensing
limits. The proposed amendment does not involve a physical change to
the plant, does not change methods of plant operation within
prescribed limits, or affect design and licensing basis functions or
controlling values of parameters for plant systems, structures, and
components.
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 W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Travis L. Tate.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida
Date of amendment request: August 7, 2014 (ADAMS Accession No.
ML14225A630).
Description of amendment request: The amendment would revise the
Technical Specifications to add a short Allowed Outage Time to restore
an inoperable system for conditions under which the existing
specifications require a plant shutdown. The proposed amendment is
consistent with an NRC-approved change identified as Technical
Specifications Task Force (TSTF) Traveler TSTF-426, Revision 5,
``Revise or Add Actions to Preclude Entry into LCO [Limiting Condition
for Operation] 3.0.3--RITSTF [Risk-Informed TSTF] Initiatives 6b & 6c''
(see 78 FR 32476, May 30, 2013). The Allowed Outage Time would be added
to specifications governing the boron injection flow paths of the
reactivity control systems, pressurizer heaters, containment spray
trains, shield building ventilation systems, and control room emergency
air cleanup systems.
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 reproduced 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 provides a short Allowed Outage Time to
restore an inoperable system for conditions under which the existing
Technical Specifications require a plant shutdown to begin within
one hour in accordance with Limiting Condition for Operation (LCO)
3.0.3. Entering into Technical Specification Actions is not an
initiator of any accident previously evaluated. As a result, the
probability of an accident previously evaluated is not significantly
increased. The consequences of any accident previously evaluated
that may occur during the proposed Allowed Outage Times are no
different from the consequences of the same accident during the
existing one-hour allowance. As a result, the consequences of any
accident previously evaluated are not significantly increased.
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.
No new or different accidents [would] result from utilizing the
proposed change. The changes [to the TSs] do not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
[any] safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change increases the time the plant may operate
without the ability to perform an assumed safety function. The
analyses in [the NRC-approved topical report] WCAP-16125-NP-A,
``Justification for Risk-Informed Modifications to Selected
Technical Specifications for Conditions Leading to Exigent Plant
Shutdown,'' Revision 2, August 2010, demonstrated that there is an
acceptably small increase in risk due to a limited period of
continued operation in these conditions and that this risk is
balanced by avoiding the risks associated with a plant shutdown. As
a result, the change to the margin of safety provided by requiring a
plant shutdown within one hour is not significant.
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 determines
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore,
the NRC staff proposes to determine that the proposed amendment
involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB,
Juno Beach, FL 33408-0420.
NRC Branch Chief: Shana R. Helton.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida
Date of amendment request: December 5, 2014 (ADAMS Accession No.
ML14353A016).
Description of amendment request: The proposed amendment will
modify the Technical Specification (TS)
[[Page 13909]]
requirements related to Completion Times for Required Actions to
provide the option to calculate longer, risk-informed Completion Times.
The proposed amendment will also add a new program, the Risk Informed
Completion Time Program, to TS section 6.0, ``Administrative
Controls.'' The methodology for using the Risk Informed Completion Time
Program is described in Nuclear Energy Institute topical report NEI 06-
09, ``Risk-Informed Technical Specifications Initiative 4b, Risk-
Managed Technical Specifications (RMTS) Guidelines,'' Revision 0-A,
which was approved by the NRC on May 17, 2007. The proposed amendment
is consistent with the NRC-approved industry-proposed Technical
Specification Task Force-505, Revision 1, ``Provide Risk-Informed
Extended Completion Times--RITSTF Initiative 4b.''
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 reproduced 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 permits the extension of Completion Times
provided the associated risk is assessed and managed in accordance
with the NRC[-]approved Risk Informed Completion Time Program. The
proposed change does not involve a significant increase in the
probability of an accident previously evaluated because the change
involves no change to the plant or its modes of operation. The
proposed change does not increase the consequences of an accident
because the design-basis mitigation function of the affected systems
is not changed and the consequences of an accident [occurring]
during the extended Completion Time are no different from those
[occurring] during the existing Completion Time.
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 change the design, configuration,
or method of operation of the plant. The proposed change does not
involve a physical alteration of the plant (no new or different kind
of equipment will be installed).
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change permits the extension of Completion Times
provided risk is assessed and managed in accordance with the NRC[-
]approved Risk Informed Completion Time Program. The proposed change
implements a risk-informed configuration management program to
assure that adequate margins of safety are maintained. Application
of these new specifications and the configuration management program
considers cumulative effects of multiple systems or components being
out of service and does so more effectively than the current TS.
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 determines
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore,
the NRC staff proposes to determine that the proposed amendment
involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB,
Juno Beach, FL 33408-0420.
NRC Branch Chief: Shana R. Helton.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: December 17, 2014. A publicly-available
version is in ADAMS under Accession No. ML14356A022.
Description of amendment request: The proposed amendment would
amend the Appendix A technical specifications to Facility Operating
Licenses DPR-58 and DPR-74, to modify the notes to TS 3.8.1, ``AC
Sources--Operating,'' to allow surveillance testing of the onsite
standby emergency diesel generators (DGs) during modes in which it is
currently prohibited. Specifically, the license amendment request
proposes removing the mode restrictions for the following Surveillance
Requirements (SRs): 3.8.1.10 (DG single largest load rejection test),
3.8.1.11 (DG full load rejection test), and 3.8.1.15 (DG endurance
run).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
The design of plant equipment is not being modified by the
proposed changes. In addition, the DGs and their associated
emergency loads are accident mitigating features. As such, testing
of the DGs themselves is not associated with any potential accident-
initiating mechanism.
Therefore, there will be no significant impact on any accident
probabilities by the approval of the requested changes.
The changes include an increase in the time that a DG under test
will be paralleled to the grid while the unit is in Modes 1 or 2. As
such, the ability of the tested DG to respond to a DBA [design-basis
accident] could be minimally adversely impacted by the proposed
changes. However, the impacts are not considered significant based,
in part, on the ability of the remaining DG to mitigate a DBA or
provide safe shutdown. Experience shows that testing for these SRs
typically does not perturb the electrical distribution system. In
addition, operating experience supports the conclusion that the
proposed changes do not involve any significant increases in the
likelihood of a safety-related bus blackout or damage to plant
loads.
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 capability to synchronize a DG to the offsite source (via
the associated plant bus) and test the DG in such a configuration is
a design feature of the DGs, including the test mode override in
response to a safety injection signal. Paralleling the DG for longer
periods of time during plant operation may slightly increase the
probability of incurring an adverse effect from the offsite source,
but this increase in probability is judged to be still quite small
and such a possibility is not a new or previously unrecognized
consideration.
The proposed change does not introduce a new mode of plant
operation and does not involve physical modification to the plant.
The change does not introduce new accident initiators or impact
assumptions made in the safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not exceed or alter a design basis or
safety limit, so there is no significant reduction in the margin of
safety. The margin of safety is related to the confidence in the
ability of the fission product barriers to perform their design
functions during and following an accident situation. These barriers
include the fuel cladding, the reactor coolant system, and the
containment system. The proposed changes do not directly affect
these barriers, nor do they involve any significantly adverse impact
on the DGs which serve to support these barriers in the event of an
accident concurrent with a LOOP [loss of offsight power]. The
proposed changes to the testing requirements for the plant DGs do
not affect the OPERABILITY requirements for the DGs,
[[Page 13910]]
as verification of such OPERABILITY will continue to be performed as
required (except during different allowed modes). The changes have
an insignificant impact on DG availability, as the DGs remain
available to perform their required function of providing emergency
power to plant equipment that supports or constitutes the fission
product barriers. Only one DG is to be tested at a time, so that the
remaining DG will be available to safety shut down the plant if
required. Consequently, performance of the fission product barriers
will not be impacted by implementation of the proposed amendment.
In addition, the proposed changes involve no changes to
setpoints or limits established or assumed by the accident analysis.
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 requests involve no significant hazards consideration.
Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
NRC Branch Chief: David L. Pelton.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: January 15, 2015. A publicly-available
version is in ADAMS under Accession No. ML15021A127.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to add a limiting condition
for operation, applicability, required actions, completion times, and
surveillance requirements for the residual heat removal (RHR)
containment spray system consistent with the guidance in NUREG-1433,
Revision 4, ``Standard Technical Specifications General Electric BWR
[Boiling Water Reactor]/4 Plants,'' dated April 2012 (ADAMS Accession
No. ML12104A192). New TS section 3.6.1.9, ``Residual Heat Removal (RHR)
Containment Spray,'' would be added to reflect the reliance on
containment spray to maintain the drywell within design temperature
limits during a small steam line break. In addition, the ``Drywell
Pressure--High'' function that serves as an interlock permissive to
allow RHR containment spray mode alignment would be relocated from the
Technical Requirements Manual (TRM) to TS 3.3.5.1, ``Emergency Core
Cooling System (ECCS) Instrumentation.''
The requirements for the RHR containment spray function and
``Drywell Pressure--High'' function are currently contained in TRM
sections T3.6.1, ``RHR Containment Spray,'' and T3.3.2, ``ECCS and
Reactor Core Isolation Cooling Instrumentation,'' respectively. These
TRM sections established specific guidance and criteria related to the
applicability, operation, and testing for the RHR containment spray
system. The TRM requirements for the RHR containment spray system would
be removed once the TS requirements are approved.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to establish the RHR Containment Spray
requirement in TS does not introduce new equipment or new equipment
operating modes, nor do the proposed changes alter existing system
relationships. The proposed change does not affect plant operation,
design function, or any analysis that verifies the capability of a
structure, system, or component (SSC) to perform a design function.
There are no changes or modifications to the RHR system. The RHR
system will continue to function as designed in all modes of
operation, including the Containment Spray function. There are no
significant changes to procedures or training related to the
operation of the Containment Spray function. Primary containment
integrity is not adversely impacted and radiological consequences
from the accidents analyzed in the Updated Safety Analysis Report
(USAR) are not increased. Containment parameters are not increased
beyond those previously evaluated and the potential for failure of
the containment is not increased.
There is no adverse impact on systems designed to mitigate the
consequences of accidents. The proposed change does not increase
system or component pressures, temperatures, and flowrates for
systems designed to prevent accidents or mitigate the consequences
of an accident. Since these conditions do not change, the likelihood
of failure of SSC is not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to establish the RHR Containment Spray
requirement in TS does not alter the design function or operation of
any SSC. The Containment system will continue to function as
designed in all modes of operation, including RHR Containment Spray
function. There is no new system component being installed, no new
construction, and no performance of a new test or maintenance
function. The proposed TS change does not create the possibility of
a new credible failure mechanism or malfunction. The proposed change
does not modify the design function or operation of any SSC. The
proposed change does not introduce new accident initiators. Primary
containment integrity is not adversely impacted and radiological
consequences from the accident analyzed in the USAR are not
increased. Containment parameters are not increased beyond those
previously evaluated and the potential for failure of the
containment is not increased. The proposed change does not increase
system or component pressures, temperatures, and flowrates for
systems designed to prevent accidents or mitigate the consequences
of an accident. Since these conditions do not change, the likelihood
of failure of an SSC is not increased.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not increase system or component
pressures, temperatures, and flowrates for systems designed to
prevent accidents or mitigate the consequences of an accident.
Containment parameters are not increased beyond those previously
evaluated and the potential for failure of the containment is not
increased.
The proposed change to establish the RHR Containment Spray
requirement in TS is needed in order to reflect the current safety
function of Containment Spray related to the small steam line break
accident. The proposed change does not exceed or alter a design
basis or a safety limit parameter that is described in the USAR.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
Acting NRC Branch Chief: Eric R. Oesterle.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of amendment request: February 20, 2013, as supplemented by
letters dated June 25, 2013; September 15,
[[Page 13911]]
2014; and February 26, 2015. Publicly-available versions are in ADAMS
under Accession Nos. ML13053A199, ML13178A024, ML14258A089, and
ML15057A480, respectively.
Brief description of amendment request: The proposed amendments
would remove the technical specification (TS) 3.5.3 ``ECCS [Emergency
Core Cooling System]-Shutdown,'' Limiting Condition for Operation (LCO)
Note 1 to eliminate information to the plant operators that could cause
non-conservative operation, and would revise the LCO Applicability
statement to apply to all of Mode 4.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration,
which the Commission previously issued in the Federal Register on
August 20, 2013 (78 FR 51229). The licensee revised its analysis of the
issue of no significant hazards consideration, which is presented
below, to consider expansion of the scope of the amendments by revising
the LCO Applicability statement to include all of Mode 4.
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This license amendment request proposes to revise the Technical
Specification for ECCS operability requirements in Mode 4 by
removing the LCO Note which allows the RHR [residual heat removal]
subsystem to be considered operable for ECCS when aligned for
shutdown cooling and revising the Applicability statement to include
all of Mode 4. These changes will require one train of RHR to be
aligned for ECCS operation throughout Mode 4.
The proposed changes do not affect the ECCS and RHR subsystem
design, the interfaces between the RHR subsystem and other plant
systems' operating functions, or the reliability of the RHR
subsystem. The proposed changes do not change or impact the
initiators and assumptions of the analyzed accidents. Therefore, the
ECCS and RHR subsystems will be capable of performing their accident
mitigation functions, and the proposed TS changes do not involve an
increase in the probability of an accident.
The proposed TS changes will require that one train of RHR is
aligned for ECCS operation during Mode 4 which assures that one
train of ECCS is operable to mitigate the consequences of a loss of
coolant accident. Thus the proposed TS changes do not involve a
significant increase in the consequences of an accident.
Therefore, the proposed Technical Specification changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This license amendment request proposes to revise the Technical
Specification for ECCS operability requirements in Mode 4 by
removing the LCO Note which allows the RHR subsystem to be
considered operable for ECCS when aligned for shutdown cooling and
revising the Applicability statement to include all of Mode 4. These
changes will require one train of RHR to be aligned for ECCS
operation throughout Mode 4.
The proposed Technical Specification changes involve changes to
when system trains are operated, but they do not change any system
functions or maintenance activities. The changes do not involve
physical alteration of the plant, that is, no new or different type
of equipment will be installed. The changes do not alter assumptions
made in the safety analyses but ensure that one train of ECCS is
operable to mitigate the consequences of a loss of coolant accident.
These changes do not create new failure modes or mechanisms which
are not identifiable during testing and no new accident precursors
are generated.
Therefore, the proposed Technical Specification changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This license amendment request proposes to revise the Technical
Specification [TS] for ECCS operability requirements in Mode 4 by
removing the LCO Note which allows the RHR subsystem to be
considered operable for ECCS when aligned for shutdown cooling and
revising the Applicability statement to include all of Mode 4. These
changes will require one train of RHR to be aligned for ECCS
operation throughout Mode 4.
This license amendment proposes Technical Specification changes
which assure that the ECCS--Shutdown TS LCO requirements are met if
a Mode 4 LOCA were to occur. With these changes, other TS
requirements for shutdown cooling in Mode 4 will continue to be met.
Based on review of plant operating experience, there is no
discernable change in cooldown rates when utilizing a single train
of RHR for shutdown cooling. Thus, no margin of safety is reduced as
part of this change.
Therefore, the proposed Technical Specification changes do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: David L. Pelton.
Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-321 and
50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, GA
Date of amendment request: January 13, 2015. A publicly-available
version is in ADAMS under Accession No. ML15014A411.
Description of amendment request: The licensee proposes to adopt
Technical Specification Task Force (TSTF) change number 523, revision
2, ``Generic Letter 2008-01, Managing Gas Accumulation,'' for the Hatch
Nuclear Plant, Unit 1 and 2, technical specifications (TS). The
proposed change would revise or add Surveillance Requirements to verify
that the system locations susceptible to gas accumulation are
sufficiently filled with water and to provide allowances which permit
performance of the verification.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises or adds Surveillance Requirement(s)
(SRs) that require verification that the Emergency Core Cooling
System (ECCS), the Residual Heat Removal (RHR) System, the RHR
Shutdown Cooling (SDC) System, the Containment Spray (CS) System,
and the Reactor Core Isolation Cooling (RCIC) System are not
rendered inoperable due to accumulated gas and to provide allowances
which permit performance of the revised verification. Gas
accumulation in the subject systems is not an initiator of any
accident previously evaluated. As a result, the probability of any
accident previously evaluated is not significantly increased. The
proposed SRs ensure that the subject systems continue to be capable
to perform their assumed safety function and are not rendered
inoperable due to gas accumulation. Thus, the consequences of any
accident previously evaluated are not significantly increased.
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 revises or adds SRs that require
verification that the ECCS, the RHR, the RHR SDC System, the CS
System, and the RCIC System are not rendered inoperable due to
accumulated gas and to
[[Page 13912]]
provide allowances which permit performance of the revised
verification. The proposed change does not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operation. In addition, the proposed change does not impose any new
or different requirements that could initiate an accident. The
proposed change does not alter assumptions made in the safety
analysis and is consistent with the safety analysis assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises or adds SRs that require
verification that the ECCS, the RHR, RHR SDC System, the CS System,
and the RCIC System are not rendered inoperable due to accumulated
gas and to provide allowances which permit performance of the
revised verification. The proposed change adds new requirements to
manage gas accumulation in order to ensure the subject systems are
capable of performing their assumed safety functions. The proposed
SRs are more comprehensive than the current SRs and will ensure that
the assumptions of the safety analysis are protected. The proposed
change does not adversely affect any current plant safety margins or
the reliability of the equipment assumed in the safety analysis.
Therefore, there are no changes being made to any safety analysis
assumptions, safety limits or limiting safety system settings that
would adversely affect plant safety as a result of the proposed
change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed change
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: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, 40 Inverness Center
Parkway, Birmingham, AL 35201.
NRC Branch Chief: Robert J. Pascarelli.
South Carolina Electric and Gas Company, Docket Nos.: 52-027 and 52-
028, Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: January 27, 2015. A publicly-available
version is in ADAMS under Accession No. ML15028A537.
Description of amendment request: The proposed change, if approved,
would revise, in part, the description and scope of human factors
engineering (HFE) operational sequence analysis (OSA) task and delete a
reference to document WCAP-15847, which are both identified as Tier 2*
information in the Updated Final Safety Analysis Report (UFSAR).
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 deletion of WCAP-15847 removes obsolete and
superseded procedures from the licensing basis. The amendment of the
operational sequence analysis (OSA) task alters the automatic
depressurization system (ADS) testing from Mode 1 to Mode 5. The
proposed changes to the procedures do not involve any accident
initiating component/system failure or event, and the change to the
ADS testing mode helps prevent accidents that would occur if the
tests were performed in Mode 1. Thus, the probabilities of the
accidents previously evaluated are not affected. The affected
procedures and requirements do not adversely affect or interact with
safety-related equipment or a radioactive material barrier, and this
activity does not involve the containment of radioactive material.
Thus, the proposed changes would not affect any safety-related
accident mitigating function. The radioactive material source terms
and release paths used in the safety analyses are unchanged, thus
the radiological releases in the Updated Final Safety Analysis
Report accident analyses are not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Removing WCAP-15847 from the UFSAR and amending the OSA task
regarding ADS valve testing does not adversely affect the design or
operation of safety-related equipment or equipment whose failure
could initiate an accident other than what is already described in
the licensing basis. These changes do not adversely affect safety-
related equipment or fission product barriers. No safety analysis or
design basis acceptance limit/criterion is challenged or exceeded by
the requested change.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to remove WCAP-15847 from the UFSAR and
amend the OSA task do not adversely affect any safety-related
equipment, design code compliance, design function, design analysis,
safety analysis input or result, or design/safety margin because
NQA-1 requirements are maintained in other Westinghouse procedures
and testing of the ADS valves is still performed. No safety analysis
or design basis acceptance limit/criterion is challenged or exceeded
by the proposed changes, thus no margin of safety is reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW, Washington, DC 20004-2514.
NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County,
Georgia
Date of amendment request: January 30, 2015. A publicly-available
version is in ADAMS under Accession No. ML15030A505.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-91 and NPF-92 for the Vogtle Electric
Generating Plant (VEGP) Units 3 and 4. The requested amendment proposes
changes to Tier 2* information contained within the Human Factors
Engineering Design Verification, Task Support Verification and
Integrated System Validation (ISV) plans. These documents are
incorporated by reference into the VEGP Units 3 and 4 Updated Final
Safety Analysis Report, and will additionally require changes to be
made to affected Tier 2 information.
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?
[[Page 13913]]
Response: No.
The proposed amendment includes changes to Integrated System
Validation (ISV) activities, which are performed on the AP1000 plant
simulator to validate the adequacy of the AP1000 human system
interface design and confirm that it meets human factors engineering
principles. The proposed changes involve administrative details
related to performance of the ISV, and no plant hardware or
equipment is affected whose failure could initiate an accident, or
that interfaces with a component that could initiate an accident, or
that contains radioactive material. Therefore, these changes have no
effect on any accident initiator in the Updated Final Safety
Analysis Report (UFSAR), nor do they affect the radioactive material
releases in the UFSAR accident analysis.
Therefore, the proposed amendment does not involve an 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 includes changes to ISV activities, which
are performed on the AP1000 plant simulator to validate the adequacy
of the AP1000 human system interface design and confirm that it
meets human factors engineering principles. The proposed changes
involve administrative details related to performance of the ISV,
and no plant hardware or equipment is affected whose failure could
initiate an accident, or that interfaces with a component that could
initiate an accident, or that contains radioactive material.
Although the ISV may identify a need to initiate changes to add,
modify, or remove plant structures, systems, or components, these
changes will not be made directly as part of the ISV.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment includes changes to ISV activities, which
are performed on the AP1000 plant simulator to validate the adequacy
of the AP1000 human system interface design and confirm that it
meets human factors engineering principles. The proposed changes
involve administrative details related to performance of the ISV,
and do not affect any safety-related equipment, design code
compliance, design function, design analysis, safety analysis input
or result, or design/safety margin. No safety analysis or design
basis acceptance limit/criterion is challenged or exceeded by the
proposed changes, thus no margin of safety is reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: September 13, 2012, as supplemented
August 2, 2013, July 3, July 17, November 11, and December 12, 2014.
Publicly-available versions are in ADAMS under Accession Nos.
ML12258A055, ML13217A072, ML14189A554, ML14198A574, ML14315A051 and
ML14346A643, respectively.
Description of amendment request: The proposed amendment would
modify certain Technical Specification (TS) requirements related to
Completion Times for Required Actions to provide the option to
calculate a longer, risk-informed Completion Time. The allowance will
be described in a new program, ``Risk Informed Completion Time Program
(RICT),'' to be approved by NRC and to be added to Chapter 5,
``Administrative Controls,'' of the Technical Specifications. The
methodology for using the RICT Program is described in an industry
document NEI 06-09, ``Risk-Informed Technical Specifications Initiative
4b, Risk-Managed Technical Specifications (RMTS) Guidelines,'' which
was approved by the Nuclear Regulatory Commission (NRC) on May 17,
2007. Adherence to NEI 06-09 is required by the proposed RICT Program.
The proposed amendment is also consistent with the methodologies
presented in an industry initiative identified as TSTF-505, Revision 1,
``Provide Risk-Informed Extended Completion Times--RITSTF Initiative
4b.'' Although the proposed amendment is consistent with TSTF-505, the
licensee is not proposing adoption of TSTF-505 with this proposed
amendment; the proposed amendment is a site-specific action.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change permits the extension of Completion Times
provided risk is assessed and managed within the Risk Informed
Completion Time Program. The proposed change does not involve a
significant increase in the probability of an accident previously
evaluated because the changes involve no change to the plant or its
modes of operation. This proposed change does not increase the
consequences of an accident because the design-basis mitigation
function of the affected systems is not changed and the consequences
of an accident during the extended Completion Time are no different
from those during the existing Completion Time.
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 change the design, configuration,
or method of operation of the plant. The proposed change does not
involve a physical alteration of the plant (no new or different kind
of equipment will be installed).
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety[?]
Response: No.
The proposed change permits the extension of Completion Times
provided risk is assessed and managed within the Risk Informed
Completion Time Program. The proposed change implements a risk-
informed configuration management program to assure that adequate
margins of safety are maintained. Application of these new
specifications and the configuration management program considers
cumulative effects of multiple systems or components being out of
service and does so more effectively than the current TS.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involve no significant hazards consideration.
Attorney for licensee: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, 40 Inverness Center
Parkway, Birmingham, AL 35242.
NRC Branch Chief: Robert J. Pascarelli.
[[Page 13914]]
Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328,
Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: December 2, 2014. A publicly-available
version is in ADAMS under Accession No. ML14339A539.
Description of amendment request: The amendments would revise
Technical Specification (TS) 6.8.4.h, ``Containment Leakage Rate
Testing Program,'' by adopting Nuclear Energy Institute (NEI) 94-01,
Revision 3-A, ``Industry Guideline for Implementing Performance-Based
Option of 10 CFR part 50, Appendix J,'' as the implementation document
for the performance-based Option B of 10 CFR part 50, Appendix J. The
proposed changes would permanently extend the Type A containment
integrated leak rate testing (ILRT) interval from 10 years to 15 years,
and the Type C local leakage rate testing (LLRT) intervals from 60
months to 75 months.
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 consequence of an accident previously evaluated?
Response: No.
The proposed revision to TS 6.8.4.h changes the testing period
to a permanent 15-year interval for Type A testing (10 CFR part 50,
Appendix J, Option B, ILRT) and a 75-month interval for Type C
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current
type A test interval of 10 years would be extended to 15 years from
the last Type A test. The proposed extension to Type A testing does
not involve a significant increase in the consequences of an
accident because research documented in NUREG-1493, ``Performance-
Based Containment System Leakage Testing Requirements [sic]
[Performance-Based Containment Leak-Test Program],'' September 1995,
has found that, generically, very few potential containment leakage
paths are not identified by Type B and C tests. NUREG-1493 concluded
that reducing the Type A testing frequency to one per twenty years
was found to lead to an imperceptible increase in risk. A high
degree of assurance is provided through testing and inspection that
the containment will not degrade in a manner detectable only by Type
A testing. The last Type A test (performed October 27, 2007 for SQN,
Unit 1 and December 30, 2006 for SQN, Unit 2) shows leakage to be
below acceptance criteria, indicating a very leak tight containment.
Inspections required by the ASME [American Society of Mechanical
Engineers] Code section Xl (subsections IWE and IWL) and Maintenance
Rule monitoring (10 CFR 50.65, ``Requirements for Monitoring the
Effectiveness of Maintenance at Nuclear Power Plants''), are
performed in order to identify indications of containment
degradation that could affect that leak tightness. Types B and C
testing required by TSs will identify any containment opening such
as valves that would otherwise be detected by the Type A tests.
These factors show that a Type A test interval extension will not
represent a significant increase in the consequences of an accident.
The proposed amendment involves changes to the SQN, Units 1 and
2, 10 CFR 50 Appendix J Testing Program Plan. The proposed amendment
does not involve a physical change to the plant or a change in the
manner in which the units are 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, and 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-accepted guidelines of NEI
94-01, Revision 3-A, for development of the SQN, Units 1 and 2,
performance-based leakage testing program. Implementation of these
guidelines continues to provide adequate assurance that during
design basis accidents, the primary containment and its components
will limit leakage rates to less than the values assumed in the
plant safety analyses. The potential consequences of extending the
ILRT interval from 10 years to 15 years have been evaluated by
analyzing the resulting changes in risk. The increase in risk in
terms of person-rem per year resulting from design basis accidents
was estimated to be very small, and the increase in the LERF [large
early release frequency] resulting from the proposed change was
determined to be within the guidelines published in NRC RG
[Regulatory Guide] 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. TVA has determined that the increase in CCFP
[conditional containment failure probability] due to the proposed
change would be very small.
Based on the above discussions, the proposed changes do not
involve an 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 revision to TS 6.8.4.h changes the testing period
to a permanent 15-year interval for Type A testing (10 CFR part 50,
Appendix J, Option B, ILRT) and a 75-month interval for Type C
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current
test interval of 10 years, based on past performance, would be
extended to 15 years from the last Type A test (performed October
27, 2007 for SQN, Unit 1 and December 30, 2006 for SQN, Unit 2). The
proposed extension to Type A and Type C test intervals does not
create the possibility of a new or different type of accident
because there are no physical changes being made to the plant and
there are no changes to the operation of the plant that could
introduce a new failure mode creating an accident or affecting the
mitigation of an accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed revision to TS 6.8.4.h changes the testing period
to a permanent 15-year interval for Type A testing (10 CFR part 50,
Appendix J, Option B, ILRT) and a 75-month interval for Type C
testing (10 CFR part 50, Appendix J, Option B, LLRT). The current
test interval of 10 years, based on past performance, would be
extended to 15 years from the last Type A test (performed October
27, 2007 for SQN, Unit 1 and December 30, 2006 for SQN, Unit 2). The
proposed extension to Type A testing will not significantly reduce
the margin of safety. NUREG-1493, ``Performance-Based Containment
System Leakage Testing Requirements [sic] [Performance-Based
Containment Leak-Test Program],'' September 1995, generic study of
the effects of extending containment leakage testing, found that a
20-year extension to Type A leakage testing resulted in an
imperceptible increase in risk to the public. NUREG-1493 found that,
generically, the design containment leakage rate contributes about
0.1% to the individual risk and that the decrease in Type A testing
frequency would have a minimal effect on this risk since 95% of the
potential leakage paths are detected by Type C testing. Regular
inspections required by the ASME Code section Xl (subsections IWE
and IWL) and maintenance rule monitoring (10 CFR 50.65,
``Requirements for Monitoring the Effectiveness of Maintenance at
Nuclear Power Plants'') will further reduce the risk of a
containment leakage path going undetected.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, for development of the SQN, Units 1 and 2,
performance-based leakage testing program, and establishes a 15-year
interval for the performance of the primary containment ILRT and a
75-month interval for Type C testing. The 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 10 CFR part 50, Appendix J
Testing Program Plan, as defined in the TS, ensure that the degree
of primary containment structural integrity and leak-tightness that
is considered in the plant safety analyses is maintained. The
overall containment leakage rate limit specified by the TS is
maintained, and the
[[Page 13915]]
Type A, B, and C containment leakage tests will continue to 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 will not degrade in a manner that is detectable only by
an ILRT. This ensures that evidence of containment structural
degradation is identified in a timely manner. Furthermore, a risk
assessment using the current SQN, Units 1 and 2, PRA model concluded
that extending the ILRT test interval from 10 years to 15 years
results in a very small change to the SQN, Units 1 and 2, risk
profile.
Accordingly, 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Shana R. Helton.
III. Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses and Combined Licenses,
Proposed No Significant Hazards Consideration Determination, and
Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit 2, Westchester County, New York
Date of amendment request: February 12, 2015. A publicly-available
version is in ADAMS under Accession No. ML15044A471.
Brief description of amendment request: The proposed amendment
would allow a revision to the acceptance criteria for the Surveillance
Requirement 3.1.4.2 for Control Rod G-3. During the last two
performances of this Surveillance on September 18, 2014, and December
11, 2014, Control Rod G-3 misalignment occurred with Shutdown Bank B
group movement as displayed by Individual Rod Position Indication and
Plant Instrument Computer System. The proposed change is to defer
subsequent testing of the Control Rod G-3 until repaired during the
next refuel outage (March 2016) or forced outage long enough to repair
the Control Rod.
Date of publication of individual notice in Federal Register: March
2, 2015 (80 FR 11236).
Expiration date of individual notice: April 1, 2015 (public
comments); May 1, 2015 (hearing requests).
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323 for
Diablo Canyon Nuclear Power Plant (DCPP), Units 1 and 2, Docket No. 72-
26 for Diablo Canyon Independent Spent Fuel Storage Installation
(ISFSI), San Luis Obispo County, California
Date of amendment request: September 24, 2013, as supplemented by
letters dated December 18, 2013 (security-related), and May 15, 2014.
Publicly-available versions of the letters dated September 24, 2013,
and May 15, 2014, are in ADAMS under Accession Nos. ML13268A398 and
ML14135A379, respectively.
Brief description of amendment request: The proposed amendments
would modify the licenses to reflect a grant of section 161A of the
Atomic Energy Act, to authorize the licensee the authority to possess
and use certain firearms, ammunition, and other devices such as large-
capacity ammunition feeding devices, to implement the NRC-approved
security plan for DCPP, Unit Nos. 1 and 2, and the Diablo Canyon ISFSI.
Date of publication of individual notice in Federal Register:
February 18, 2015 (80 FR 8706).
Expiration date of individual notice: March 20, 2015 (public
comments); April 19, 2015 (hearing requests).
Southern California Edison Company, et al., Docket Nos. 50-361, 50-362,
and 72-41, San Onofre Nuclear Generating Station, Units 2 and 3, and
Independent Spent Fuel Storage Installation, San Diego County,
California
Date of amendment request: August 28, 2013, as supplemented by
letters dated December 31, 2013, May 15, 2014, and February 10, 2015.
Publicly-available versions are in ADAMS under Accession Nos.
ML13242A277, ML14007A496, ML14139A424, and ML15044A047, respectively.
Brief description of amendment request: The licensee is requesting
that the Commission grant it preemption authority consistent with the
Commission's authority under section 161A of the Atomic Energy Act of
1954, as amended, to authorize the security personnel of designated
classes of licensees to possess, use, and access covered weapons for
the physical security of SONGS, Units 2 and 3, and the Independent
Spent Fuel Storage Installation, notwithstanding Federal, State, or
local laws prohibiting such possession or use. If the amendment request
is granted, the licenses would be modified to reflect the Commission's
granting of section 161A preemption authority.
Date of publication of individual notice in Federal Register:
February 18, 2015 (80 FR 8701).
Expiration date of individual notice: March 20, 2015 (public
comments); April 20, 2015 (hearing requests).
IV. Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses and Final Determination of No Significant Hazards
Consideration and Opportunity for a Hearing (Exigent Public
Announcement or Emergency Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual notice of
consideration of issuance of amendment, proposed no significant hazards
consideration determination, and opportunity for a hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the
[[Page 13916]]
Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License or Combined License, as applicable, 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
``Obtaining Information and Submitting Comments'' section of this
document.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, any person(s) whose interest may be
affected by this action may file a request for a hearing and a petition
to intervene with respect to issuance of the amendment to the subject
facility operating license 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, and electronically on the Internet at the NRC's Web
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR's Reference staff
at 1-800-397-4209, 301-415-4737, or by email to [email protected].
If a request for a hearing or petition for leave to intervene is filed
by the above date, the Commission or a presiding officer designated by
the Commission or by the Chief Administrative Judge of the Atomic
Safety and Licensing Board Panel, will rule on the request and/or
petition; and the Secretary or the Chief Administrative Judge of the
Atomic Safety and Licensing Board will issue a 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 petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or fact.
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A 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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
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
[[Page 13917]]
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.
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.
Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating
Station, Unit 2, Montgomery County, Pennsylvania
Date of amendment request: February 12, 2015.
Description of amendment request: The amendment extends the
implementation period for Amendment No. 174, ``Leak Detection System
Setpoint and Allowable Value Changes,'' which was issued on December
29, 2014. Amendment No. 174 was effective as of the date of issuance
(i.e., on December 29, 2014) and was required to be implemented within
60 days (i.e., by February 27, 2015). Amendment No. 177 extends the
implementation period for Amendment No. 174 from 60 days to prior to
startup from the spring 2015 refueling outage.
Date of issuance: February 25, 2015.
Effective date: As of its date of issuance and shall be implemented
prior to startup from the Spring 2015 Unit 2 Refueling Outage.
Amendment No.: 177. A publicly-available version is in ADAMS under
Accession No. ML15049A084; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
[[Page 13918]]
Renewed Facility Operating License Nos. NPF-85: Amendment revised
the Renewed Facility Operating License to extend the implementation
date of Amendment No. 174, issued on December 29, 2014, to prior to
startup from the Spring 2015 Unit 2 Refueling Outage.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. Public notice of the proposed amendment was
published in The Pottstown Mercury, located in in Pottstown,
Pennsylvania, on February 15, and February 16, 2015. The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. Comments were received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, public comments, and final
NSHC determination are contained in a safety evaluation dated February
25, 2015.
Attorney for licensee: J. Bradley Fewell, Esquire, Vice President
and Deputy General Counsel, Exelon Generation Company, LLC, 200 Exelon
Way, Kennett Square, PA 19348.
NRC Branch Chief: Douglas A. Broaddus.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Respectively,
Limestone County, Alabama
Date of amendment request: February 12, 2015.
Brief description of amendment request: The amendments revised
Technical Specification (TS) 5.6.5, ``Core Operating Limits Report
(COLR),'' to add the date of a previously issued NRC safety evaluation
(SE) that stated it was acceptable for the licensee to use new
analytical methods supporting the use of ATRIUM 10XM (10XM) fuel. In
its letter dated February 12, 2015, the licensee stated BFN, Unit 2, is
entering an outage on March 14, 2015, and is scheduled to commence
loading 10XM fuel on March 17, 2015. Because the TSs do not reference
the aforementioned NRC evaluation, the licensee would not be able to
issue a COLR for the Unit 2 transition cycle unless the notation to the
latest NRC SE is added. Therefore, the licensee requested that NRC
process the license amendment request under exigent circumstances in
accordance with 10 CFR 50.91(a)(6). The NRC staff determined that the
provisions of 10 CFR 50.91(a)(6) were applicable for processing the
licensee's request under exigent circumstances.
Date of issuance: February 26, 2015.
Effective date: As of the date of issuance and shall be implemented
during the refueling outages in fall of 2016 for Unit 1, in spring of
2015 for Unit 2, and in spring of 2016 for Unit 3.
Amendment Nos.: 288, 313, and 272, which are available in ADAMS
under Accession No. ML15051A337. Documents related to these amendments
are listed in the SE enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the TSs.
Public comments requested as to proposed no significant hazards
consideration (NSHC): The public notice was published in ``The
Huntsville Times,'' located in Huntsville, Alabama, on February 18 and
20, 2015. The notice provided an opportunity to submit comments on the
Commission's proposed NSHC determination. No comments have been
received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated February 26, 2015.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, TN 37902.
NRC Branch Chief: Shana R. Helton.
V. 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
``Obtaining Information and Submitting Comments'' section of this
document.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of application for amendment: December 12, 2012, as
supplemented by letters dated February 21, September 30, October 24,
and December 2, 2013; April 2, May 7, June 17, August 14, November 4,
and December 18, 2014.
Brief description of amendment: The amendment authorizes the
transition of the Palisades Nuclear Plant 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: February 27, 2015.
Effective date: As of its date of issuance and shall be implemented
by six months from the date of issuance.
Amendment No.: 254. A publicly-available version is in ADAMS under
Accession No. ML15007A191; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-20: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: February 27, 2014 (79
FR 11148). The supplements dated April 2, May 7, June 17, August 14,
November 4, and December 18, 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.
[[Page 13919]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 27, 2015.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
1, Pope County, Arkansas
Date of amendment request: March 26, 2013, as supplemented by
letters dated November 14, 2013, and August 18, October 22, and
December 5, 2014.
Brief description of amendment: The amendment revised the Technical
Specification (TS) requirements for end states associated with the
implementation of the NRC-approved Topical Report BAW-2441-A, Revision
2, ``Risk-Informed Justification for LCO End-State Changes,'' as well
as Required Actions revised by a specific Note in TS Task Force (TSTF)
change traveler TSTF-431, Revision 3, ``Change in Technical
Specifications End States (BAW-2441).''
Date of issuance: March 3, 2015.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 253. A publicly-available version is in ADAMS under
Accession No. ML15023A147; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-51: Amendment revised
the TSs/license.
Date of initial notice in Federal Register: July 23, 2013 (78 FR
44170). The supplemental letters dated November 14, 2013, and August
19, October 22, and December 5, 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 March 3, 2015.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 9, 2013, as supplemented by
letters dated October 1, 2014, and December 17, 2014.
Brief description of amendment: The amendment revised the Technical
Specifications for the Waterford Steam Electric Station, Unit 3 to
improve clarity, correct administrative and typographical errors, or
establish consistency with NUREG-1432, ``Standard Technical
Specifications--Combustion Engineering Plants,'' Revision 4.0.
Date of issuance: February 23, 2015.
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 242. A publicly-available version is in ADAMS under
Accession No. ML15005A126; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-38: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 5, 2014 (79 FR
45475). The supplements dated October 1, 2014, and December 17, 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 February 23, 2015.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida
Date of amendment request: February 26, 2014, as supplemented by
letters dated May 29 and July 25, 2014.
Brief description of amendment: The amendments revised the
Technical Specifications (TSs), modifying requirements for mode change
limitations in Limiting Condition for Operation 3.0.4 and Surveillance
Requirement (SR) 4.0.4 to adopt the provisions of Industry/TS Task
Force (TSTF)-359, Rev. 9, ``Increase Flexibility in MODE Restraints.''
The language of SR 4.0.1 is revised to conform to the language of
NUREG-1432, ``Standard Technical Specifications for Combustion
Engineering Plants,'' to resolve language incongruences and ensure
conservative implementation of the TSTF-359, Rev. 9, changes.
Date of issuance: February 27, 2015.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 220 and 170. A publicly-available version is in
ADAMS under Accession No. ML14343A918; documents related to these
amendments are listed in the Safety Evaluation (SE) enclosed with the
amendments.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: May 27, 2014 (79 FR
30187). The supplements dated May 29 and July 25, 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 amendments is contained
in a SE dated February 27, 2015.
No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Nuclear Power Plant, Units 1 and 2, Somervell County,
Texas
Date of amendment request: July 1, 2014.
Brief description of amendment: The amendments revised Technical
Specification 3.8.1, ``AC [Alternating Current] Sources--Operating,''
to extend on a one-time basis the Completion Time (CT) of Required
Action A.3, ``Restore required offsite circuit to OPERABLE status,''
from 72 hours to 14 days. The CT extension from 72 hours to 14 days
will be used while completing the plant modification to install
alternate startup transformer XST1A and will expire on March 31, 2017.
Date of issuance: February 24, 2015.
Effective date: As of the date of issuance and shall be implemented
within [licensee requested number] days from the date of issuance.
Amendment Nos.: Unit 1--164; Unit 2--164. A publicly-available
version is in ADAMS under Accession No. ML15008A133; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: October 28, 2014 (79 FR
64226).
The Commission's related evaluation of the amendments is contained
in a
[[Page 13920]]
Safety Evaluation dated February 24, 2015.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit 1, Washington County, Nebraska
Date of amendment request: April 30, 2014, as supplemented by
letter dated January 27, 2015.
Brief description of amendment: The amendment revised Technical
Specification section 3.2, Table 3-5, for Fort Calhoun Station, Unit
No. 1, to add a new surveillance requirement to verify the correct
position of the valves required to restrict flow in the high pressure
safety injection system.
Date of issuance: February 20, 2015.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment No.: 280. A publicly-available version is in ADAMS under
Accession No. ML15015A413; 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 license and Technical Specifications.
Date of initial notice in Federal Register: August 19, 2014 (79 FR
49108). The supplemental letter dated January 27, 2015, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated February 20, 2015.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 14, 2014, as supplemented by
letter dated December 18, 2014.
Brief description of amendments: The amendments revised
Administrative Controls Technical Specification (TS) 6.9.1.6, ``Core
Operating Limits Report (COLR),'' with respect to the analytical
methods used to determine the core operating limits.
Date of issuance: February 27, 2015.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1--204; Unit 2--192. A publicly-available
version is in ADAMS under Accession No. ML15049A129; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: December 2, 2014 (79 FR
71455). The supplemental letter dated December 18, 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 amendments is contained
in a Safety Evaluation dated February 27, 2015.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 9th day of March 2015.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2015-05994 Filed 3-16-15; 8:45 am]
BILLING CODE 7590-01-P