[Federal Register Volume 81, Number 50 (Tuesday, March 15, 2016)]
[Notices]
[Pages 13837-13849]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-05470]
[[Page 13837]]
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NUCLEAR REGULATORY COMMISSION
[NRC-2016-0050]
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 13, 2016, to February 29, 2016.
The last biweekly notice was published on March 1, 2016.
DATES: Comments must be filed by April 14, 2016. A request for a
hearing must be filed by May 16, 2016.
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-2016-0050. 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: Lynn Ronewicz, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-1927, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2016-0050 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-2016-0050.
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 of this document.
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-2016-0050, 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
[[Page 13838]]
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 within 60 days, the
Commission or a presiding officer designated by the Commission or by
the Chief Administrative Judge of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the 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 with respect to resolution of that person's admitted
contentions, including the opportunity to present evidence and to
submit a cross-examination plan for cross-examination of witnesses,
consistent with NRC regulations, policies and procedures.
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).
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, 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.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission by May
16, 2016. The petition must be filed in accordance with the filing
instructions in the ``Electronic Submissions (E-Filing)'' section of
this document, and should meet the requirements for petitions for leave
to intervene set forth in this section, except that under Sec.
2.309(h)(2) a State, local governmental body, or Federally-recognized
Indian Tribe, or agency thereof does not need to address the standing
requirements in 10 CFR 2.309(d) if the facility is located within its
boundaries. A State, local governmental body, Federally-recognized
Indian Tribe, or agency thereof may also have the opportunity to
participate under 10 CFR 2.315(c).
If a hearing is granted, any person who does not wish, or is not
qualified, to become a party to the proceeding may, in the discretion
of the presiding officer, be permitted to make a limited appearance
pursuant to the provisions of 10 CFR 2.315(a). A person making a
limited appearance may make an oral or written statement of position on
the issues, but may not otherwise participate in the proceeding. A
limited appearance may be made at any session of the hearing or at any
prehearing conference, subject to the limits and conditions as may be
imposed by the presiding officer. Persons desiring to make a limited
appearance are requested to inform the Secretary of the Commission by
May 16, 2016.
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 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at
[[Page 13839]]
[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, in some instances, 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 Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station (CNS), Units 1 and 2, York County, South Carolina
Date of amendment request: January 18, 2016. A publicly-available
version is in ADAMS under Accession No. ML16026A048.
Description of amendment request: The proposed amendments would
modify the Renewed Facility Operating Licenses and Technical
Specifications (TS) for CNS, Units 1 and 2. Specifically, the proposed
amendments request to revise TS 5.5.2, ``Containment Leakage Rate
Testing Program,'' to allow an increase in the existing Type A
Integrated Leakage Rate Test (ILRT) program test interval from 10 years
to 15
[[Page 13840]]
years in accordance with Nuclear Energy Institute (NEI) Topical Report
NE1 94-01, Revision 3-A, ``Industry Guideline for Implementing
Performance-Based Option of 10 CFR part 50, appendix J,'' and the
conditions and limitations specified in NEI 94-01, Revision 2-A;
adoption of an extension of the containment isolation valve leakage
testing (Type C) frequency from the 60 months currently permitted by 10
CFR part 50, appendix J, Option B, to a 75-month frequency for Type C
leakage rate testing of selected components, in accordance with NEI 94-
01, Revision 3-A; adoption of the use of ANSI/ANS 56.8-2002,
``Containment System Leakage Testing Requirements''; and adoption of a
more conservative grace interval of 9 months for Type A, Type B, and
Type C leakage tests in accordance with NEI 94-01, Revision 3-A. The
proposed amendments also request the following administrative changes:
Deletion of the information regarding the performance of containment
visual inspections as required by Regulatory Position C.3, as the
containment inspections are addressed in TS Surveillance Requirement
3.6.1.1, deletion of the information regarding the performance of the
next CNS, Unit 1, Type A test no later than November 13, 2015, and the
next CNS, Unit 2, Type A test no later than February 6, 2008, as both
Type A tests have already occurred.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with NRC edits in square
brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment to the Technical Specifications (TS)
involves the extension of the Catawba Nuclear Station (CNS) Type A
containment integrated leak rate test interval to 15 years and the
extension of the Type C test interval to 75 months for selected
components. The current Type A test interval of 120 months (10
years) would be extended on a permanent basis to no longer than 15
years from the last Type A test. The current Type C test interval of
60 months for selected components would be extended on a performance
basis to no longer than 75 months. Extensions of up to nine months
(total maximum interval of 84 months for Type C tests) are
permissible only for non-routine emergent conditions. The proposed
extension does not involve either a physical change to the plant or
a change in the manner in which the plant is operated or controlled.
The containment is designed to provide an essentially leak tight
barrier against the uncontrolled release of radioactivity to the
environment for postulated accidents. The containment and the
testing requirements invoked 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 the
prevention or identification of any precursors of an accident. The
change in dose risk for changing the Type A test frequency from
three-per-ten years to once-per-fifteen years, measured, as an
increase to the total integrated plant risk for those accident
sequences influenced by Type A testing, is 0.026 person-rem/year.
EPRI Report No. 1009325, Revision 2-A states that a very small
population dose is defined as an increase of [less than or equal to]
1.0 person-rem per year, or [less than or equal to] 1% of the total
population dose, whichever is less restrictive for the risk impact
assessment of the extended ILRT intervals. Therefore, this proposed
extension does not involve a significant increase in the probability
of an accident previously evaluated.
As documented in NUREG-1493, Type B and C tests have identified
a very large percentage of containment leakage paths, and the
percentage of containment leakage paths that are detected only by
Type A testing is very small. The CNS Type A test history supports
this conclusion.
The integrity of the containment is subject to two types of
failure mechanisms that can be categorized as: (1) Activity based,
and; (2) time based. Activity based failure mechanisms are defined
as degradation due to system and/or component modifications or
maintenance. Local leak rate test requirements and administrative
controls such as configuration management and procedural
requirements for system restoration ensure that containment
integrity is not degraded by plant modifications or maintenance
activities. The design and construction requirements of the
containment combined with the containment inspections performed in
accordance with ASME Section Xl, the Maintenance Rule, and TS
requirements serve to provide a high degree of assurance that the
containment would not degrade in a manner that is detectable only by
a Type A test. Based on the above, the proposed extensions do not
significantly increase the consequences of an accident previously
evaluated.
The proposed amendment also deletes an exception previously
granted to allow one-time extensions of the Unit 1 and Unit 2 ILRT
test frequency for CNS. This exception was for activities that have
already taken place; therefore, their deletion is solely an
administrative action that has no effect on any component and no
impact on how the units are operated.
Therefore, the proposed change does not result in 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 to the TS involves the extension of the
CNS Type A containment integrated leak rate test interval to 15
years and the extension of the Type C test interval to 75 months for
selected components.
The current Type A test interval of 120 months (10 years) would
be extended on a permanent basis to no longer than 15 years from the
last Type A test. The current Type C test interval of 60 months for
selected components would be extended on a performance basis to no
longer than 75 months. 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.
The proposed amendment also deletes an exception previously
granted to allow one-time extensions of the Unit 1 and Unit 2 ILRT
test frequency for CNS. This exception was for activities that have
already taken; therefore, their deletion is solely an administrative
action that does not result in any change in how the units are
operated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The proposed amendment to TS 5.5.2 involves the extension of the
CNS Type A containment integrated leak rate test interval to 15
years and the extension of the Type C test interval to 75 months for
selected components. The current Type A test interval of 120 months
(10 years) would be extended on a permanent basis to no longer than
15 years from the last Type A test. The current Type C test interval
of 60 months for selected components would be extended on a
performance basis to no longer than 75 months. This amendment does
not alter the manner in which safety limits, limiting safety system
set points, or limiting conditions for operation are determined. The
specific requirements and conditions of the TS Containment Leak Rate
Testing Program exist to ensure that the degree of containment
structural integrity and leak tightness that is considered in the
plant safety analysis is maintained. The overall containment leak
rate limit specified by TS is maintained.
The proposed change involves only the extension of the interval
between Type A containment leak rate tests, and Type C tests for
CNS. The proposed surveillance interval extension is bounded by the
15-year ILRT interval, and the 75-month Type C test interval
currently authorized within NEI 94-01, Revision 3-A. Industry
experience supports the conclusion that Type B and C testing detects
a large percentage of containment leakage paths and that the
percentage of containment leakage paths that are detected only by
Type A testing is small. The containment inspections performed in
accordance with ASME Section Xl, TS and the Maintenance Rule serve
to provide a high
[[Page 13841]]
degree of assurance that the containment would not degrade in a
manner that is detectable only by Type A testing. The combination of
these factors ensures that the margin of safety in the plant safety
analysis is maintained. The design, operation, testing methods and
acceptance criteria for Type A, B, and C containment leakage tests
specified in applicable codes and standards would continue to be
met, with the acceptance of this proposed change, since these are
not affected by changes to the Type A, and Type C test intervals.
The proposed amendment also deletes an exception previously
granted to allow one-time extensions of the Unit 1 and Unit 2 ILRT
test frequency for CNS. This exception was for activities that have
already taken place; therefore, their deletion is solely an
administrative action and does not change how the units are operated
and maintained. Thus, there is no reduction in any margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Associate General Counsel,
Duke Energy Corporation, 526 South Church Street-EC07H, Charlotte, NC
28202.
NRC Branch Chief: Michael T. Markley.
Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: November 19, 2015. A publicly-available
version is in ADAMS under Accession No. ML15323A085.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to allow the extension of the
Type A containment test interval to 15 years and the extension of the
Type B and Type C test intervals for selected components to 120 months
and 75 months, respectively. The proposed amendment also deletes from
the TSs an already implemented one-time extension of the Type A test
frequency.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment to the Technical Specifications (TS)
involves the extension of the H. B. Robinson Steam Electric Plant
Unit No. 2 (HBRSEP2) Type A containment test interval to 15 years,
the extension of the Type B test intervals to 120 months for
selected components, and the extension of the Type C test interval
to 75 months for selected components. The current Type A test
interval of 120 months (10 years) would be extended on a permanent
basis to no longer than 15 years from the last Type A test. The
current Type B test interval of each reactor shutdown for refueling
but in no case at intervals greater than 2 years would be extended
on a performance basis to no longer than 120 months. The current
Type C test interval of each reactor shutdown for refueling but in
no case at intervals greater than 2 years would be extended on a
performance basis to no longer than 75 months. Extensions of up to
nine months (total maximum interval of 84 months for Type C tests)
are permissible only for non-routine emergent conditions. The
proposed extensions do not involve either a physical change to the
plant or a change in the manner in which the plant is operated or
controlled. The containment is designed to provide an essentially
leak tight barrier against the uncontrolled release of radioactivity
to the environment for postulated accidents. The containment and the
testing requirements invoked 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 the
prevention or identification of any precursors of an accident. The
change in dose risk for changing the Type A test frequency from
three-per-ten years to once-per-fifteen years, measured, as an
increase to the total integrated plant risk for those accident
sequences influenced by Type A testing, is 0.020 person-rem
[roentgen equivalent man]/year. The Electric Power Research
Institute (EPRI) Report No. 1009325, Revision 2-A, states that a
very small population dose is defined as an increase of <=1.0
person-rem per year, or <=1% of the total population dose, whichever
is less restrictive for the risk impact assessment of the extended
integrated leak rate test (ILRT) intervals. Therefore, this proposed
extension does not involve a significant increase in the probability
of an accident previously evaluated.
As documented in NUREG-1493, Type B and C tests have identified
a very large percentage of containment leakage paths, and the
percentage of containment leakage paths that are detected only by
Type A testing is very small. The HBRSEP2 Type A test history
supports this conclusion.
The integrity of the containment is subject to two types of
failure mechanisms that can be categorized as: (1) Activity based,
and (2) time based. Activity based failure mechanisms are defined as
degradation due to system and/or component modifications or
maintenance. Local leak rate test requirements and administrative
controls such as configuration management and procedural
requirements for system restoration ensure that containment
integrity is not degraded by plant modifications or maintenance
activities. The design and construction requirements of the
containment combined with the containment inspections performed in
accordance with the American Society of Mechanical Engineers (ASME)
Section XI, the Maintenance Rule, and TS requirements serve to
provide a high degree of assurance that the containment would not
degrade in a manner that is detectable only by a Type A test. Based
on the above, the proposed extensions do not significantly increase
the consequences of an accident previously evaluated.
The proposed amendment also deletes an exception previously
granted to allow one-time extension of the ILRT test frequency for
HBRSEP2. This exception was for an activity that has already taken
place so the deletion is solely an administrative action that has no
effect on any component and no impact on how the unit is operated.
Therefore, the proposed change does not result in 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 to the TS involves the extension of the
HBRSEP2 Type A containment test interval to 15 years, the Type B
test interval to 120 months for selected components and the
extension of the Type C test interval to 75 months for selected
components. 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.
The proposed amendment also deletes an exception previously
granted to allow one-time extension of the ILRT test frequency for
HBRSEP2. This exception was for an activity that has already taken
place so the deletion is solely an administrative action that has no
effect on any component and no impact on how the unit is operated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment to TS 5.5.16 involves the extension of
the HBRSEP2 Type A containment test interval to 15 years, the Type B
test interval to 120 months for selected components and the
extension of the Type C test interval to 75 months for selected
components. This amendment does not alter the manner in which safety
limits, limiting safety system set points, or limiting conditions
for operation are determined. The specific requirements and
conditions of the
[[Page 13842]]
TS Containment Leak Rate Testing Program exist to ensure that the
degree of containment structural integrity and leak tightness that
is considered in the plant safety analysis is maintained. The
overall containment leak rate limit specified by TS is maintained.
The proposed change involves only the extension of the interval
between Type A containment leak rate tests, Type B tests and Type C
tests for HBRSEP2. The proposed surveillance interval extension is
bounded by the 15-year ILRT interval, the 120-month Type B interval
and the 75-month Type C test interval currently authorized within
NEI 94-01, Revision 3-A. Industry experience supports the conclusion
that Types B and C testing detects a large percentage of containment
leakage paths and that the percentage of containment leakage paths
that are detected only by Type A testing is small. The containment
inspections performed in accordance with ASME Section XI, TS and the
Maintenance Rule serve to provide a high degree of assurance that
the containment would not degrade in a manner that is detectable
only by Type A testing. The combination of these factors ensures
that the margin of safety in the plant safety analysis is
maintained. The design, operation, testing methods and acceptance
criteria for Types A, B, and C containment leakage tests specified
in applicable codes and standards would continue to be met, with the
acceptance of this proposed change, since these are not affected by
changes to the Type A, Type B and Type C test intervals.
The proposed amendment also deletes an exception previously
granted to allow one-time extension of the ILRT test frequency for
HBRSEP2. This exception was for an activity that has already taken
place so the deletion is solely an administrative action that has no
effect on any component and no impact on how the unit is operated.
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: Lara S. Nichols, Deputy General Counsel,
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A,
Charlotte, NC 28202.
NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station (LGS), Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: January 15, 2016. A publicly-available
version is in ADAMS under Accession No. ML16015A316.
Description of amendment request: The amendments would reduce the
reactor vessel steam dome pressure associated with the Technical
Specification (TS) Safety Limits (SLs) specified in TS 2.1.1 and TS
2.1.2. The amendments would also revise the setpoint and allowable
value for the main steam line low pressure isolation function in TS
Table 3.3.2-2. The proposed changes address a 10 CFR part 21 issue
concerning the potential to violate the SLs limits during a pressure
regulator failure maximum demand (open) transient.
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 do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because decreasing the reactor vessel steam dome pressure in TS
Safety Limits 2.1.1 and 2.1.2 for reactor thermal power ranges and
increasing the trip setpoint and allowable value for the main steam
line low pressure isolation effectively expands the validity range
for GEXL critical power correlation and the calculation of the
minimum critical power ratio. The critical power ratio rises during
the pressure reduction following the scram that terminates the
Pressure Regulator Failure Maximum Demand (Open) (PRFO) transient.
The reduction in the reactor vessel steam dome pressure value in the
SL and the increase in the trip setpoint and the allowable value for
the main steam line low pressure isolation provides adequate margin
to accommodate the pressure reduction during the PRFO transient
within the revised TS limit.
The proposed changes do not alter the use of the analytical
methods used to determine the safety limits that have been
previously reviewed and approved by the NRC. The proposed changes
are in accordance with an NRC approved critical power correlation
methodology and do not adversely affect accident initiators or
precursors.
The proposed changes do not alter or prevent the ability of
structures, systems, and components from performing their intended
function to mitigate the consequences of an initiating event within
the applicable acceptance limits. The proposed changes are
consistent with the safety analysis and resultant consequences.
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.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the proposed reduction in the reactor vessel steam dome
pressure value in the safety limit in conjunction with the increase
in the trip setpoint and the allowable value for the main steam line
low pressure isolation reflects a wider range of applicability for
the GEXL critical power correlation which is approved by the NRC for
both GE14 and GNF2 fuel types in [the] LGS reactor cores.
In addition, no new failure modes are being introduced. There
are no changes in the method by which any plant systems perform a
safety function. No new accident scenarios, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
changes.
The proposed changes do not introduce any new accident
precursors, nor do they involve any changes in the methods governing
normal plant operation. The proposed changes do not alter the
outcome of the safety analysis.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the
plant structures, systems, and components, and through the
parameters for safe operation and setpoints for the actuation of
equipment relied upon to respond to transients and design basis
accidents. Evaluation of the 10 CFR part 21 condition by General
Electric determined that, since the critical power ratio improves
during the PRFO transient, there is no impact on the fuel safety
margin, and therefore, there is no challenge to fuel cladding
integrity. The proposed changes do not change the requirements
governing operation or availability of safety equipment assumed to
operate to preserve the margin of safety.
The proposed changes are consistent with the applicable NRC
approved critical power correlation for the fuel designs in use at
LGS. The proposed changes do not alter the manner in which the
safety limits are determined.
The reduction in value of the reactor vessel steam dome pressure
safety limit and the increase in the trip setpoint and allowable
value for the main steam line low pressure isolation provides
adequate margin to accommodate the pressure reduction during the
PRFO transient within the revised TS limit.
Therefore, the proposed changes do not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 13843]]
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: Douglas A. Broaddus.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station (DBNPS), Unit No. 1, Ottawa County,
Ohio
Date of amendment request: February 17, 2016. A publicly-available
version is in ADAMS under Accession No. ML16049A513.
Description of amendment request: The licensee proposes to change
the emergency plan for DBNPS, Unit No. 1, by revising the emergency
action level (EAL) scheme based on the Nuclear Energy institute's
(NEl's) guidance in NEI 99-01, Revision 6, ``Development of Emergency
Action Levels for Non-Passive Reactors.'' The NEI 99-01, Revision 6,
was endorsed by the NRC by letter dated March 28, 2013 (ADAMS Accession
No. ML12346A463).
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 to DBNPS's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any
physical changes to plant systems or equipment. The proposed changes
do not alter any of the requirements of the technical
specifications. The proposed changes do not modify any plant
equipment and do not impact any failure modes that could lead to an
accident. Additionally, the proposed changes do not impact the
ability of structures, systems, or components (SSCs) to perform
their intended safety functions in mitigating the consequences of an
initiating event within the assumed acceptance limits.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to DBNPS's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any
physical changes to plant systems or equipment. The proposed changes
do not involve the addition of any new plant equipment. The proposed
changes will not alter the design configuration, or method of
operation of plant equipment beyond its normal functional
capabilities. DBNPS functions will continue to be performed as
required. The proposed changes do not create any new credible
failure mechanisms, malfunctions, or accident initiators.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to DBNPS's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any
physical changes to plant systems or equipment. Margins of safety
are unaffected by the proposed changes. There are no changes being
made to safety analysis assumptions, safety limits, or limiting
safety system settings that would adversely affect plant safety as a
result of the proposed EAL scheme change. The proposed change does
not affect the technical specifications. There are no changes to
environmental conditions of any of the SSC or the manner in which
any SSC is operated. The applicable requirements of 10 CFR 50.47 and
10 CFR part 50, appendix E will continue to be met.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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.
Acting NRC Branch Chief: Justin C. Poole.
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: January 29, 2016. A publicly-available
version is in ADAMS under Accession No. ML16034A032.
Description of amendment request: The proposed amendment would
modify technical specification (TS) requirements to address Generic
Letter 2008-01, ``Managing Gas Accumulation in Emergency Core Cooling,
Decay Heat Removal, and Containment Spray Systems,'' as described in
the Technical Specification Task Force (TSTF) Traveler TSTF-523,
Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation.''
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 SRs [Surveillance
Requirements] that require verification that the ECCS [Emergency
Core Cooling System], RHR [Residual Heat Removal] System, and the
Containment Spray (CTS) 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 System, and the CTS System are
not rendered inoperable due to accumulated gas and to 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 System, and the CTS 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
[[Page 13844]]
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.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
NRC Branch Chief: David J. Wrona.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment request: January 21, 2106. A publicly-available
version is in ADAMS under Accession No. ML16021A067.
Description of amendment request: The amendments 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.
The amendments would revise Technical Specification (TS) 3.4.6, ``RCS
[Reactor Coolant System] Loops--MODE 4''; TS 3.4.7, ``RCS Loops--MODE
5, Loops Filled''; TS 3.4.8, ``RCS Loops--MODE 5, Loops Not Filled'';
TS 3.5.2, ``ECCS [Emergency Core Cooling System]--Operating''; TS
3.6.6, ``Containment Spray and Cooling Systems''; TS 3.9.5, ``RHR
[Residual Heat Removal] and Coolant Circulation--High Water Level'';
and TS 3.9.6, ``RHR and Containment Circulation--Low Water Level.'' The
proposed amendments would modify TS requirements to address Generic
Letter 2008-01, ``Managing Gas Accumulation in Emergency Core Cooling,
Decay Heat Removal, and Containment Spray Systems,'' as described in
Technical Specification Task Force TSTF-523, Revision 2, ``Generic
Letter 2008-01, Managing Gas Accumulation.''
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, and the
Containment Spray (CS) 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 accident from any accident previously evaluated?
Response: No.
The proposed change revises or adds SRs that require
verification that the ECCS, RHR System, and CS System are not
rendered inoperable due to accumulated gas and to 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 System, and the CS 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.
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: Jennifer Post, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, CA 94120.
NRC Branch Chief: Robert J. Pascarelli.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: January 19, 2016. A publicly-available
version is in ADAMS under Accession No. ML16019A403.
Description of amendment request: The requested amendment proposes
to depart from Tier 2* information in the Updated Final Safety Analysis
Report (which includes the plant-specific design control document Tier
2 information) related to the construction methods used for the
composite floors and roof of the auxiliary building.
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 design functions of the nuclear island structures are to
provide support, protection, and separation for the seismic Category
I mechanical and electrical equipment located in the nuclear island.
The nuclear island structures are structurally designed to meet
seismic Category I requirements as defined in Regulatory Guide 1.29.
The use of ACI 349 and AISC N690 provides criteria for the
design, qualification, fabrication, and inspection of composite
steel beam floors and roof in the auxiliary building. These
structures continue to meet the applicable portions of ACI 349 and
AISC N690. The proposed change does not have an adverse impact on
the response of the nuclear island structures to safe shutdown
earthquake ground motions or loads due to
[[Page 13845]]
anticipated transients or postulated accident conditions. The change
does not impact the support, design, or operation of mechanical and
fluid systems. There is no change to plant systems or the response
of systems to postulated accident conditions. There is no change to
the predicted radioactive releases due to normal operation or
postulated accident conditions. The plant response to previously
evaluated accidents or external events is not adversely affected,
nor does the change described create any new accident precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change revises the description of the construction
of composite steel beam floors and roof in the auxiliary building.
The proposed change does not change the design function, support,
design, or operation of mechanical and fluid systems. The proposed
change does not result in a new failure mechanism for the pertinent
structures or new accident precursors. As a result, the design
function of the structures is not adversely affected by the proposed
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 change is consistent with ACI 349 and AISC N690.
The design and construction of the auxiliary building floors and
roof remain in conformance with the requirements in ACI 349 and AISC
N690.
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.
Acting NRC Branch Chief: John McKirgan.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos.
1, 2, and 3, Maricopa County, Arizona
Date of amendment request: February 27, 2015, as supplemented by
letter dated January 19, 2016.
Brief description of amendments: The amendments revised Technical
Specification (TS) 1.3, ``Completion Times''; TS 3.7.5, ``Auxiliary
Feedwater (AFW) System''; TS 3.8.1, ``AC [Alternating Current]
Sources--Operating''; and TS 3.8.9, ``Distribution Systems--
Operating''; to remove the second Completion Times. The amendment also
revised Example 1.3-3 in TS 1.3, ``Completion Times,'' by adding a
discussion of administrative controls to combinations of conditions to
ensure that the Completion Times for those conditions are not
inappropriately extended.
The changes are consistent with the NRC-approved Technical
Specification Task Force (TSTF) Traveler TSTF-439-A, Revision 2,
``Eliminate Second Completion Times Limiting Time From Discovery of
Failure to Meet an LCO [Limiting Condition of Operation],'' dated June
20, 2005.
Date of issuance: February 19, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: Unit 1--197; Unit 2--197; Unit 3--197. A publicly-
available version is in ADAMS under Accession No. ML16004A013;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74:
The amendments revised the Operating Licenses and TSs.
Date of initial notice in Federal Register: May 12, 2015 (80 FR
27195). The supplement dated January 19, 2016, 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 19, 2016.
No significant hazards consideration comments received: No.
Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina
Date of amendment request: February 19, 2015, as supplemented by
letter dated November 5, 2015.
Description of amendment request: The amendments revised (1)
technical specifications (TSs) by replacing AREVA Topical Report ANP-
10298PA, ``ACE/ATRIUM 10XM Critical Power Correlation,'' Revision 0,
March 2010, with Revision 1, March 2014, of the same topical report;
and (2) Appendix B, ``Additional Conditions,'' by removing the license
condition issued by Amendment Nos. 262 and 290 for Units 1 and Unit 2,
respectively.
Date of issuance: February 9, 2016.
Effective date: Once approved, the Unit 1 amendment shall be
implemented prior to start-up. from the 2016 Unit 1 refueling outage,
and the Unit 2 amendment shall be implemented prior to start-up from
the 2017 Unit 2 refueling outage.
Amendment Nos.: 269 and 297. A publicly-available version is in
ADAMS under Accession No. ML16019A029;
[[Page 13846]]
documents related to these amendments are listed in the Safety
Evaluation (SE) enclosed with the amendments.
Facility Operating License Nos. DPR-71, and DPR-62: Amendments
revised the renewed facility operating licenses and TSs.
Date of initial notice in Federal Register: April 28, 2015 (80 FR
23603). The supplemental letter dated November 5, 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 an SE dated February 9, 2016.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station (CGS),
Benton County, Washington
Date of amendment request: September 2, 2015.
Brief description of amendment: The amendment revised the Technical
Specification (TS) requirements for unavailable barriers by adding
Limiting Condition for Operation (LCO) 3.0.9. The LCO allows a delay
time for entering a supported system TS, when the inoperability is
solely due to an unavailable barrier, if the risk is assessed and
managed. The change is consistent with NRC-approved Technical
Specification Task Force (TSTF) Standard Technical Specification (STS)
Change TSTF-427, Revision 2, ``Allowance for Non Technical
Specification Barrier Degradation on Supported System OPERABILITY''
(ADAMS Accession No. ML061240055). The availability of this TS
improvement was published in the Federal Register on October 3, 2006
(71 FR 58444), as part of the Consolidated Line Item Improvement
Process.
Additionally, LCO 3.0.8 has been revised to replace the term
``train'' with ``division'' to be consistent with CGS's TS definition
of ``OPERABLE-OPERABILITY'' and the terminology used in Section 1.3,
``Completion Times,'' of the CGS TS.
Date of issuance: February 16, 2016.
Effective date: As of its date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 237. A publicly-available version is in ADAMS under
Accession No. ML16020A031; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-21: The amendment
revised the Facility Operating License and TSs.
Date of initial notice in Federal Register: October 27, 2015 (80 FR
65811).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 16, 2016.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One (ANO), Units 1 and 2, Pope County, Arkansas
Date of amendment request: May 20, 2015.
Brief description of amendments: The amendments revised the full
implementation date (Milestone 8) of the ANO, Units 1 and 2, Cyber
Security Plan, and revised the associated physical protection license
conditions for each renewed facility operating license.
Date of issuance: February 24, 2016.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1--255; Unit 2--303. A publicly-available
version is in ADAMS under Accession No. ML16027A109; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-51 and NPF-6: The
amendments revised the renewed facility operating licenses.
Date of initial notice in Federal Register: June 23, 2015 (80 FR
35982).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 24, 2016.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: December 9, 2014, as supplemented by two
letters dated May 20, 2015, and letters dated June 8, 2015, and June
29, 2015.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.5.14, ``Containment Leakage Rate Testing
Program,'' to extend the frequency of the containment integrated leak
rate test from once every 10 years to once every 15 years on a
permanent basis.
Date of issuance: February 23, 2016.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 283. A publicly-available version is in ADAMS under
Accession No. ML15349A794; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. DPR-26: The amendment revised the
Facility Operating License and the Technical Specifications.
Date of initial notice in Federal Register: March 17, 2015 (80 FR
13905). The supplemental letters dated May 20, 2015; June 8, 2015; and
June 29, 2015, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the NRC staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 23, 2016.
No significant hazards consideration comments received: Yes. The
comments submitted by the State of New York on November 20, 2015, are
addressed in the NRC staff's Safety Evaluation dated February 23, 2016.
Entergy Operations, Inc.; System Energy Resources, Inc.; South
Mississippi Electric Power Association; and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne
County, Mississippi
Date of amendment request: May 27, 2015, as supplemented by letters
dated October 28, 2015, and December 10, 2015.
Brief description of amendment: The amendment revised the GGNS
Technical Specifications (TSs) to allow for a permanent extension of
the Type C leakage rate testing frequency and reduction of the Type B
and Type C grace intervals that are required by GGNS TS 5.5.12, ``10
CFR part 50, appendix J, Testing Program,'' by including a reference to
Nuclear Energy Institute (NEI) Topical Report, NEI 94-01, Revision 3-A,
``Industry Guideline for Implementing Performance-Based Option of 10
CFR part 50, appendix J,'' dated July 2012. In addition, the amendment
changed Surveillance Requirement (SR) 3.6.5.1.1 by deleting the
information regarding the performance of the last Type A test that has
already occurred. This amendment
[[Page 13847]]
does not alter the Type A testing frequencies nor any other
requirements as specified in the existing GGNS TS.
Date of issuance: February 17, 2016.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No: 209. A publicly-available version is in ADAMS under
Accession No. ML16011A247; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-29: The amendment revised the
Facility Operating License and TSs.
Date of initial notice in Federal Register: September 29, 2015 (80
FR 58516). The supplemental letters dated October 28, 2015, and
December 10, 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 17, 2016.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: March 25, 2014, as supplemented by
letters dated October 7, 2014, and August 24, 2015.
Brief description of amendment: The amendment modifies the
Technical Specifications (TSs) by relocating certain surveillance
frequencies to a licensee-controlled program, the Surveillance
Frequency Control Program, using probabilistic risk guidelines
contained in NRC-approved Nuclear Energy Institute (NEI) 04-10,
Revision 1, ``Risk-Informed Technical Specifications Initiative 5b,
Risk-Informed Method for Control of Surveillance Frequencies.'' The
changes are consistent with the approved Technical Specification Task
Force (TSTF) Traveler TSTF-425, Revision 3, ``Relocate Surveillance
Frequencies to Licensee Control-RITSTF Initiative 5b.''
Date of issuance: February 23, 2016.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 171. A publicly-available version is in ADAMS under
Accession No. ML15307A349; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-58: Amendment revised the
Facility Operating License and TSs.
Date of initial notice in Federal Register: September 16, 2014 (79
FR 55512). The supplemental letters dated October 7, 2014, and August
24, 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 23, 2016.
No significant hazards consideration comments received: No.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of amendment request: October 12, 2015.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) related to facility staff qualifications
for licensed operators.
Date of issuance: February 25, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos: 268 and 263. A publicly-available version is in
ADAMS under Accession No. ML16008B072; documents related to these
amendments are listed in the Safety Evaluation (SE) enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: December 22, 2015 (80
FR 79620).
The Commission's related evaluation of the amendments is contained
in an SE dated February 25, 2016.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: August 20, 2015, as supplemented by
letter dated January 27, 2016.
Brief description of amendment: The amendment made administrative
changes to update personnel and committee titles in the Technical
Specifications (TSs), deleted outdated or completed additional actions
contained in Appendix B, Additional Conditions, of the license, and
relocated the definition of Process Control Program from the TSs to the
Updated Safety Analysis Report.
Date of issuance: February 23, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 286. A publicly-available version is in ADAMS under
Accession No. ML15307A013; 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, TSs, and Appendix B to the license.
Date of initial notice in Federal Register: October 13, 2015 (80 FR
61486). The supplemental letter dated January 27, 2016, 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, 2016.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company (PG&E), Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo
County, California
Date of amendment request: February 25, 2015, as supplemented by
letter dated July 8, 2015.
Brief description of amendments: The amendments incorporated into
the licensing basis an analysis of pressurizer reaching a water-solid
(filled) condition associated with the main feedwater pipe rupture
accident summarized in the Updated Final Safety Analysis Report
(UFSAR), Section 15.4.2.2. Further, the amendments involved the
addition of time critical operator actions and modifications of the
PG&E Design Class I backup nitrogen accumulators, which are credited in
the new pressurizer filling analysis.
Date of issuance: February 19, 2016.
Effective date: As of its date of issuance and shall be implemented
within 90 days following PG&E implementation of Design Class 1 backup
nitrogen accumulator modifications, planned for the nineteenth
refueling outage 2R19 for Unit No. 2.
[[Page 13848]]
Amendment Nos.: Unit 1--223; Unit 2--225. A publicly-available
version is in ADAMS under Accession No. ML16032A006; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses and UFSAR.
Date of initial notice in Federal Register: April 28, 2015 (80 FR
23605). The supplemental letter dated July 8, 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 amendments is contained
in a Safety Evaluation dated February 19, 2016
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: May 12, 2015, as supplemented by letters
dated September 15, 2015; November 25, 2015; and January 28, 2016.
Brief description of amendments: The amendments revised and added
Surveillance Requirements to verify that the system locations
susceptible to gas accumulation are sufficiently filled with water and
to provide allowances that permit performance of the verification. The
changes are consistent with Technical Specification Trask Force
Traveler (TSTF)-523, Revision 2, ``Generic Letter 2008-01, Managing Gas
Accumulation.''
Date of issuance: February 26, 2016.
Effective date: As of its date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: Unit 1--200, Unit 2--196. A publicly-available
version is in ADAMS under Accession No. ML15345A131, documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-2 and NPF-8: The amendments
revised the Renewed Facility Operating Licenses and Technical
Specifications.
Date of initial notice in Federal Register: June 23, 2015 (80 FR
35982). The supplemental letters dated September 15, 2015; November 25,
2015; and January 28, 2016, 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 26, 2016.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: May 18, 2015.
Description of amendment: The amendment authorizes changes to the
VCSNS, Units 2 and 3 Updated Final Safety Analysis Report by revising
the Radiation Emergency Plan to expand the plume exposure pathway
emergency planning zone (EPZ) boundary. The Evacuation Time Estimates
Study and Alert and Notification System Design Report have also been
revised to encompass the expanded EPZ boundary.
Date of issuance: February 5, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No.: 41. A publicly-available version is in ADAMS under
Accession No. ML15292A404; documents related to this amendment are
listed in a Safety Evaluation enclosed with the amendment.
Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendment
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: September 29, 2015 (80
FR 585120).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 5, 2016.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: August 21, 2015, as supplemented by
letters dated September 17, 2015, and September 22, 2015.
Brief description of amendment: The amendment authorized changes to
the VEGP, Units 3 and 4, Updated Final Safety Analysis Report in the
form of departures from the incorporated plant-specific Design Control
Document Tier 2* and associated Tier 2 information. The changes are to
demonstrate that the capacity of mechanical couplers welded to
structural steel embed plates required by American Concrete Institute
(ACI) 349-01, ``Code Requirements for Nuclear Safety Related Concrete
Structures,'' is satisfied using American Institute of Steel
Construction (AISC) N690-1994, ``Specification for the Design,
Fabrication, and Erection of Steel Safety-Related Structures for
Nuclear Facilities,'' analysis and testing provisions.
Date of issuance: November 5, 2015.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 40. A publicly-available version is in ADAMS under
Accession No. ML15287A031; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: September 3, 2015 (80
FR 53340). The supplemental letters dated September 17, 2015, and
September 22, 2015, provided additional information that did not change
the scope or the conclusions of the no significant hazards
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 5, 2015.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: May 8, 2015, as supplemented by letter
dated November 9, 2015.
Brief description of amendment: The amendment revised Technical
Specifications (TSs) 2.1.1.1 and 5.6.5 to adopt the NRC-approved
methodologies of Westinghouse Commercial Atomic Power reports (WCAP)-
14483-A, ``Generic Method for Expanded Core Operating Limits Report,''
and WCAP-14565-P-A, Addendum 2-P-A, ``VIPRE-1 Modeling and
Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic
Safety Analysis,'' respectively. The change in TS 2.1.1.1 would provide
the departure from nucleate boiling ratio in a form that reduces the
need for cycle-specific license amendments, and the change in TS 5.6.5
adds an NRC-approved methodology for determining core operating limits.
[[Page 13849]]
Date of issuance: February 29, 2016.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 216. A publicly-available version is in ADAMS under
Accession No. ML16020A516; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-30: The amendment
revised the operating license and TSs.
Date of initial notice in Federal Register: July 7, 2015 (80 FR
38763). The supplemental letter dated November 9, 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 29, 2016.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket No. 50-339, North Anna
Power Station, Unit No. 2, Louisa County, Virginia
Date of amendment request: May 22, 2015. As supplemented by letter
dated October 13, 2015.
Brief description of amendment: The amendment revised the Technical
Specification (TS) 3.8.1, ``AC Sources-Operating,'' to remove the
limitation in Note 1 that the surveillance is only applicable to Unit
1. Revised Surveillance Requirement (SR) 3.8.1.8 is applicable to both
units.
Date of issuance: February 22, 2016.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 260. A publicly-available version is in ADAMS under
Accession No. ML16013A444. Documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-7: Amendment revised the
Facility Operating License and Technical Specification.
Date of initial notice in Federal Register: July 21, 2015 (80 FR
43131). The supplement letter dated October 13, 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 22, 2016.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 2nd day of March 2016.
For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2016-05470 Filed 3-14-16; 8:45 am]
BILLING CODE 7590-01-P