[Federal Register Volume 81, Number 90 (Tuesday, May 10, 2016)]
[Notices]
[Pages 28891-28905]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-10949]
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NUCLEAR REGULATORY COMMISSION
[NRC-2016-0093]
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 April 12 to April 25, 2016. The last
biweekly notice was published on April 26, 2016 (81 FR 24659).
DATES: Comments must be filed by June 9, 2016. A request for a hearing
must be filed by July 11, 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-2016-0093. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: 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: Shirley Rohrer, Office of Nuclear
[[Page 28892]]
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-5411, email: [email protected].
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2016-0093 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-2016-0093.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in the
SUPPLEMENTARY INFORMATION section.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2016-0093, 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 will post all comment submissions at http://www.regulations.gov as well as enter 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 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, (2) create the possibility of a new or different
kind of accident from any accident previously evaluated, or (3) involve
a significant reduction in a margin of safety. The basis for this
proposed determination for each amendment request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity to Request a Hearing and Petition for Leave to Intervene
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license or
combined license. Requests for a hearing and a petition for leave to
intervene shall be filed in accordance with the Commission's ``Agency
Rules of Practice and Procedure'' in 10 CFR part 2. Interested
person(s) should consult a current copy of 10 CFR 2.309, which is
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The
NRC's regulations are accessible electronically from the NRC Library on
the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is
filed 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 the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also 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
[[Page 28893]]
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 July
11, 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
July 11, 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
ten 10 days prior to the filing deadline, the participant should
contact the Office of the Secretary by email at [email protected],
or by telephone at 301-415-1677, to request (1) a digital
identification (ID) certificate, which allows the participant (or its
counsel or representative) to digitally sign documents and access the
E-Submittal server for any proceeding in which it is participating; and
(2) advise the Secretary that the participant will be submitting a
request or petition for hearing (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic
[[Page 28894]]
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.
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.
Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power
Station, Unit No. 1 (MPS1), New London County, Connecticut
Date of amendment request: March 28, 2014. A publicly-available
version is in the ADAMS under Accession No. ML14093A028.
Description of amendment request: The amendment would make changes
to the MPS1 Permanently Defueled Technical Specifications (PDTSs) by
deleting the Table of Contents section and making administrative
changes to the PDTSs.
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 are administrative in nature. The proposed
changes remove the PDTS Table of Contents section and make two other
administrative changes to the PDTSs. Furthermore, MPS1 has
permanently ceased operation and is being maintained in a defueled
condition. Therefore, the only credible design basis accident is a
fuel handling accident. The administrative changes proposed herein
are not initiators of any fuel handling accident previously
evaluated, and, consequently, the probability and consequences of a
fuel handling accident previously evaluated is not significantly
increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2) Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are administrative in nature, therefore no
new or different accidents result from the proposed changes. The
changes do not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed), a change in
the method of plant operation, or new operator actions. The changes
do not alter assumptions made in the safety analysis.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3) Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
The proposed administrative changes do not involve a change in
the method of plant operation, do not affect any accident analyses,
and do not relax any safety system settings.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
NRC Branch Chief: Bruce A. Watson.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: February 18, 2016. A publicly-available
version is in ADAMS under Accession No. ML16076A413.
Description of amendment request: The amendment would allow a one-
time extension to the 10-year frequency of the McGuire Nuclear Station,
Units 1
[[Page 28895]]
and 2, containment leakage rate tests. The change would extend the
period from 10 years to 10.5 years for each unit.
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 McGuire Nuclear Station (MNS) Type A
containment integrated leak rate test interval to 10.5 years. The
current Type A test interval of 120 months (10 years) would be
extended on a one-time basis to no longer than 10.5 years from the
last Type A test. This extension is bounded by the 15 month
extension, permissible only for non-routine emergent conditions,
allowed in accordance with NEI [Nuclear Energy Institute] 94-01
revision 0. The proposed extension also does not change the test
method or procedure. 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 10 years to 10.5 years, measured, as an increase to
the total integrated plant risk for those accident sequences
influenced by Type A testing, is 0.023 person-rem/year. EPRI
[Electric Power Research Institute] 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 ILRT [integrated leak rate test] intervals.
Therefore, this proposed extension does not involve a significant
increase in the probability of an accident previously evaluated.
As documented in NUREG-1493, Performance-Based Containment Leak-
Test Program, 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 MNS 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 as previously discussed. 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 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.
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 amendment 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
MNS Type A containment integrated leak rate test interval from 10
years to 10.5 years. The current Type A test interval of 120 months
(10 years) would be extended on a one-time basis to 10.5 years from
the last Type A test. The containment and the testing requirements
to periodically demonstrate the integrity of the containment exist
to ensure the plant's ability to mitigate the consequences of an
accident do not involve any accident precursors or initiators. The
proposed change does not involve a physical change to the plant
(i.e., no new or different type of equipment will be installed) or a
change to the manner in which the plant is operated or controlled.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Response: No.
The proposed amendment to TS 5.5.2 involves the extension of the
MNS Type A containment integrated leak rate test interval to 10.5
years. The current Type A test interval of 120 months (10 years)
would be extended on a one-time basis to no longer than 10.5 years
from the last Type A test. 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 for MNS. The proposed
surveillance interval extension is bounded by the 15-year ILRT
interval currently authorized within NEI 94-01, Revisions 2-A and 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 XI, and TS 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 Type A, B, and C
containment leakage tests specified in applicable codes and
standards would continue to be met, with the approval of this
proposed change, since these are not affected by changes to the Type
A test intervals.
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, 526 South Church Street--EC07H, Charlotte, NC
28202.
NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station (CPS), Unit No. 1, DeWitt County, Illinois
Date of amendment request: January 25, 2016, as supplemented by
letter dated March 31, 2016. A publicly-available version is in ADAMS
under Accession Nos. ML16025A182 and ML16076A077.
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) to allow a permanent
extension of the Type ``A'' integrated leak rate testing and Type ``C''
leak rate testing frequencies. This request also proposes to delete
information in TS 5.5.13 regarding a completed requirement to perform
Type ``C'' testing in 2008.
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 activity involves the extension of the Clinton
Power Station (CPS), Unit 1, Type A containment test interval to
[[Page 28896]]
15 years, and the extension of the Type C test interval to 75
months. 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. As such, 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 Integrated Leak
Rate Test (ILRT) interval from three-per-ten years to once-per-
fifteen-years, measured as an increase to the total integrated dose
risk for all accident sequences, is 3.80E-03 person-rem/yr using the
EPRI [Electric Power Research Institute] guidance with the base case
corrosion included. This change meets both of the related acceptance
criteria for change in population dose of less than 1.0 person-rem/
yr or less than 1% person-rem/yr. The change in dose risk drops to
9.37E-04 person-rem/yr when using the EPRI Expert Elicitation
methodology. The change in dose risk meets both of the related
acceptance for change in population dose of less than 1.0 person-
rem/yr or less than 1% person-rem/yr. Therefore, this proposed
extension does not involve a significant increase in the probability
of an accident previously evaluated.
In addition, as documented in NUREG-1493, Types 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 CPS, Unit 1 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 American Society of Mechanical Engineers (ASME)
Section XI, and Technical Specifications (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
CPS. This exception was for an activity that has already taken
place; therefore, this deletion is solely an administrative action
that does not result in any change in how CPS 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 5.5.13, ``Primary Containment
Leakage Rate Testing Program,'' involves the extension of the CPS,
Unit 1 Type A containment test interval to 15 years and the
extension of the Type C test interval to 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.
The proposed change does not involve a physical change to the
plant (i.e., no new or different type of equipment will be
installed) nor does it alter the design, configuration, or change
the manner in which the plant is operated or controlled beyond the
standard functional capabilities of the equipment.
The proposed amendment also deletes an exception previously
granted to allow one-time extension of the ILRT test frequency for
CPS. This exception was for an activity that has already taken
place; therefore, this deletion is solely an administrative action
that does not result in any change in how CPS 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 5.5.13 involves the extension of
the CPS, Unit 1 Type A containment test interval to 15 years 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 TS Containment Leak Rate Testing Program exist to
ensure that the degree of containment structural integrity and
leaktightness 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 the extension of the interval
between Type A containment leak rate tests and Type C tests for CPS,
Unit 1. 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 [Nuclear Energy Institute] 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, and TS
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 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 exceptions previously
granted to allow one time extensions of the ILRT test frequency for
CPS, Unit 1. This exception was for an activity that has taken
place; therefore, the deletion is solely an administrative action
and does not change how CPS is operated and maintained. Thus, there
is no reduction in any margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Acting Branch Chief: Justin C. Poole.
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: February 23, 2016. A publicly-available
version is in ADAMS under Accession No. ML16054A359.
Description of amendment request: The amendment would revise the
Technical Specifications to incorporate previously NRC-approved
Industry/Technical Specification Task Force 439 (TSTF-439), Revision 2,
``Eliminate Second Completion Times Limiting Time From Discovery of
Failure To Meet an LCO.''
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
[[Page 28897]]
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates certain Completion Times from the
Technical Specifications. Completion Times are not an initiator to
any accident previously evaluated. As a result, the probability of
an accident previously evaluated is not affected. The consequences
of an accident during the revised Completion Time are no different
than the consequences of the same accident during the existing
Completion Times. As a result, the consequences of an accident
previously evaluated are not affected by this change. The proposed
change does not alter or prevent the ability of SSCs [systems,
structures, and components] from performing their intended function
to mitigate the consequences of an initiating event within the
assumed acceptance limits. The proposed change does not affect the
source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. Further, the proposed change does not
increase the types or amounts of radioactive effluent that may be
released offsite, nor significantly increase individual or
cumulative occupational/public radiation exposures. The proposed
change is consistent with the safety analysis assumptions and
resultant consequences. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. The proposed change does not alter any assumptions made
in the safety analysis. Therefore, the proposed change does not
create the possibility of anew 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 to delete the second Completion Time does
not alter the manner in which safety limits, limiting safety system
settings or limiting conditions for operation are determined. The
safety analysis acceptance criteria are not affected by this change.
The proposed change will not result in plant operation in a
configuration outside of the design basis. 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: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine
Mile Point Nuclear Station (NMPNS), Units 1 and 2, Oswego County, New
York
Date of amendment request: March 18, 2016. A publicly-available
version is in ADAMS under Accession No. ML16078A065.
Description of amendment request: The amendment would revise the
Technical Specifications (TS) concerning a change to the method of
calculating core reactivity for the purpose of performing the
Reactivity Anomalies surveillance at NMPNS, Units 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed TS changes do not affect any plant systems,
structures, or components designed for the prevention or mitigation
of previously evaluated accidents. The amendment would only change
how the Reactivity Anomalies surveillance is performed. Verifying
that the core reactivity is consistent with predicted values ensures
that accident and transient safety analyses remain valid. This
amendment changes the TS requirements such that, rather than
performing the surveillance by comparing predicted to actual control
rod density, the surveillance is performed by a direct comparison of
keff.
Therefore, since the Reactivity Anomalies surveillance will
continue to be performed by a viable method, the proposed amendment
does not involve a significant increase in the probability or
consequence of a previously evaluated accident.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This TS amendment request does not involve any changes to the
operation, testing, or maintenance of any safety-related, or
otherwise important to safety systems. All systems important to
safety will continue to be operated and maintained within their
design bases. The proposed changes to the Reactivity Anomalies
surveillance will only provide a new, more efficient method of
detecting an unexpected change in core reactivity.
Since all systems continue to be operated within their design
bases, no new failure modes are introduced and the possibility of a
new or different kind of accident is not created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This proposed TS amendment proposes to change the method for
performing the Reactivity Anomalies surveillance from a comparison
of predicted to actual control rod density to a comparison of
predicted to monitored keff. The direct comparison of
keff provides a technically superior method of
calculating any differences in the expected core reactivity. The
Reactivity Anomalies surveillance will continue to be performed at
the same frequency as is currently required by the TS, only the
method of performing the surveillance will be changed. Consequently,
core reactivity assumptions made in safety analyses will continue to
be adequately verified.
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: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois and Docket Nos.
STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle
County, Illinois
Date of amendment request: February 23, 2016. A publicly-available
version is in ADAMS under Accession No. ML16055A149.
Description of amendment request: The amendment would (1) revise
Technical Specification (TS) 4.2.1, ``Reactor Core, Fuel Assemblies,''
to add Optimized ZIRLO\TM\, as an approved fuel rod cladding material,
(2) revise TS 5.6.5.b to add the Westinghouse topical reports for
Optimized ZIRLO\TM\ and ZIRLO[supreg], and (3) revise TS 5.6.5.b with a
non-technical change to the Reference 11 title (replace a semicolon
with a period).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 28898]]
EGC [Exelon Generation Company] has evaluated the proposed
changes for Braidwood and Byron, using the criteria in 10 CFR 50.92,
and has determined that the proposed changes do not involve a
significant hazards consideration. The following information is
provided to support a finding of no significant hazards
consideration.
Criteria
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change would allow the use of Optimized ZIRLO\TM\
clad nuclear fuel in the reactors. The NRC approved topical report
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, ``Optimized ZIRLO\TM\
prepared by Westinghouse Electric Company LLC (Westinghouse),
addresses Optimized ZIRLO\TM\ and demonstrates that Optimized
ZIRLO\TM\ has essentially the same properties as currently licensed
ZIRLO[supreg]. The fuel cladding itself is not an accident initiator
and does not affect accident probability. With the approved
exemption, use of Optimized ZIRLO\TM\ fuel cladding will continue to
meet all 10 CFR 50.46 acceptance criteria and, therefore, will not
increase the consequences of an accident. 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.
Use of Optimized ZIRLO\TM\ clad fuel will not result in changes
in the operation or configuration of the facility. Topical Report
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, demonstrated that the
material properties of Optimized ZIRLO\TM\ are similar to those of
standard ZIRLO[supreg]. Therefore, Optimized ZIRLO\TM\ fuel rod
cladding will perform similarly to those fabricated from standard
ZIRLO[supreg] thus precluding the possibility of the fuel cladding
becoming an accident initiator and causing a new or different type
of accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not involve a significant reduction in
the margin of safety. Topical Report WCAP-12610-P-A & CENPD-404-P-A,
Addendum 1-A, demonstrated that the material properties of the
Optimized ZIRLO\TM\ are not significantly different from those of
standard ZIRLO[supreg]. Optimized ZIRLO\TM\ is expected to perform
similarly to standard ZIRLO[supreg] for all normal operating and
accident scenarios, including both loss of coolant accident (LOCA)
and non-LOCA scenarios. For LOCA scenarios, where the slight
difference is Optimized ZIRLO\TM\ material properties relative to
standard ZIRLO[supreg] could have some impact on the overall
accident scenario, plant-specific LOCA analyses using Optimized
ZIRLO\TM\ properties will demonstrate that the acceptance criteria
of 10 CFR 50.46 have been satisfied. Therefore, the proposed change
does not involve a significant reduction in a margin of safety.
Based on the above, EGC concludes that the proposed amendment to
allow the use of Optimized ZIRLO\TM\ fuel cladding material does not
involve a significant hazards consideration under the standards set
forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Acting Branch Chief: Justin C. Poole.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: March 15, 2016. A publicly-available
version is in ADAMS under Accession No. ML16075A411.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.6.2.2, ``Suppression Pool Water
Level,'' as well as TS surveillance requirements 3.6.2.4.1 and
3.6.2.4.4 associated with TS 3.6.2.4, ``Suppression Pool Makeup System
(SPMU),'' to allow installation of the reactor well to steam dryer
storage pool gate in the upper containment pool (UCP) in MODES 1, 2,
and 3. The proposed amendment would also create new special operations
TS 3.10.9, ``Suppression Pool Makeup--MODE 3 Upper Containment Pool
Drain-Down,'' to allow draining of the reactor well portion of the UCP
in MODE 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The changes proposed in the license amendment request specify
different water level requirements in the upper containment pool and
suppression pool to permit gate installation in MODES 1, 2, and 3,
and drain-down of the reactor well in MODE 3. The probability of an
accident previously evaluated is unrelated to the water level in
these pools, since they are mitigating systems. The operation or
failure of a mitigating system does not contribute to the occurrence
of an accident. No active or passive failure mechanisms that could
lead to an accident are affected by these proposed changes.
Suppression pool water levels are increased during upper pool
gate installation in MODES 1, 2, and 3 and during reactor well
drain-down in MODE 3, with a potential for an increased probability
of drywell flooding during an inadvertent dump of the upper
containment pool. An inadvertent dump of the upper pool during any
period of operation with a pressurized vessel does not represent, in
and of itself, any significant hazard to the public, the plant
operating personnel, or any plant equipment. The piping components
which would be affected in this event have been analyzed for the
flooding effect, and it has been determined that this event could
not initiate a loss of coolant accident (LOCA).
The changes have no impact on the ability of any of the
emergency core cooling systems (ECCS) to function adequately, since
adequate net positive suction head (NPSH) is maintained. The
increase in suppression pool water level to compensate for the
reduction in UCP volume will provide reasonable assurance that the
minimum post-accident vent coverage is adequate to assure the
pressure suppression function of the suppression pool is
accomplished. The suppression pool water level will be raised above
the current high water level for the proposed reactor well drain-
down activity only after the reactor pressure has been reduced
sufficiently to assure that the hydrodynamic loads from a loss of
coolant accident will not exceed the design values. The reduced
reactor pressure will also ensure that the loads due to main steam
safety relief valve actuation with an elevated pool level are within
the design loads.
Relative to dose rates on the refuel floor, the resultant dose
rates from the reactor in MODES 3 and 4 are the same regardless of a
drain-down of the upper pool reactor well. Relative to a low
pressure LOCA in MODE 3, the reduced post-LOCA containment pressure
and the decay time to reach MODE 3 conditions ensures that post-
accident dose consequences are bounded by the design-basis accident
LOCA.
Therefore, the proposed amendment does not significantly
increase 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 an accident previously evaluated?
Response: No.
The proposed changes specify different water level requirements
in the upper containment pool and suppression pool to permit gate
installation in MODES 1, 2, and 3, and drain-down of the reactor
well in MODE 3. These changes do not affect or alter the ability of
the suppression pool makeup
[[Page 28899]]
(SPMU) system to perform its design function. The proposed change in
the pool water levels will maintain the design function of
mitigating the pressure and temperature increase generated by a
LOCA, and will maintain the required drywell vent coverage during
post-accident ECCS draw down.
The altered water levels in the pools do not create a different
type of accident than presently evaluated. With the reduced pressure
in the reactor coolant system, the GOTHIC computer program
simulations demonstrate that the accident responses at defined
conditions with the reactor well drained in MODE 3 are bounded by
the current design basis accidents.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to the UCP and the suppression pool water
levels do not introduce any new setpoints at which protective or
mitigating actions are initiated. Current instrument setpoints
remain unaltered by this change. Although the water levels are
adjusted for the UCP gate installation and the reactor well drain-
down activity, the design and functioning of the containment
pressure suppression system remains unchanged. The proposed total
water volume is sufficient to provide high confidence that the
pressure suppression and containment systems will be capable of
mitigating large and small break accidents. All analyzed accident
results remain within the design values for the structures and
equipment.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: David J. Wrona.
Pacific Gas and Electric Company (PG&E), Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo
County, California
Date of amendment request: March 23, 2016. A publicly-available
version is in ADAMS under Accession No. ML16084A588.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.4.12, ``Low Temperature
Overpressure Protection (LTOP) System,'' to reflect the mass input
transient analysis that assumes an emergency core cooling system (ECCS)
centrifugal charging pump (CCP) and the normal charging pump (NCP)
capable of simultaneously injecting into the reactor coolant system
(RCS) during TS 3.4.12 applicability.
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 TS 3.4.12 to allow an ECCS CCP and
the NCP aligned to LTOP orifice to be capable of injecting into the
RCS during low RCS pressures and temperatures. The LCO [Limiting
Condition for Operation] provides RCS overpressure protection by
having a minimum coolant input capability and have adequate pressure
relief capability. Analyses have demonstrated that one power
operated relief valve (PORV) or an RCS vent of at least 2.07 square
inches is capable of limiting the RCS pressure excursions below the
10 CFR 50, Appendix G limits for the design basis LTOP limits.
The proposed change does not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained. The proposed change does not
adversely affect the ability of structures, systems, and components
to perform their intended safety function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change does not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of any accident previously
evaluated. Further, the proposed change does not increase the types
and amounts of radioactive effluent that may be released offsite,
nor significantly increase individual or cumulative occupational/
public radiation exposure.
The NRC has previously evaluated a similar LAR [license
amendment request] related to Wolf Creek Generating Station. In
Amendment No. 207, the NRC concluded that the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated [ADAMS Accession
No. ML13282A534].
In 2007, PG&E replaced the Unit 1 non-safety-related PDP
[positive displacement pump] with a non-safety-related CCP, called
the NCP, in order to alleviate operational issues associated with
the PDP. In 2008, PG&E performed the replacement on Unit 2. PG&E
also designed, tested, and installed an FCO [flow choking orifice]
called the LTOP orifice to be used during LTOP operation to ensure
that the total maximum mass injection capability with the NCP
remained bounded by the LTOP mass injection analysis. These changes
were implemented under 10 CFR 50.59. However, no physical changes
are being made to the plant as a result of the proposed license
amendment.
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 TS 3.4.12 to allow an ECCS CCP and
the NCP aligned to LTOP orifice to be capable of simultaneously
injecting into the RCS during low RCS pressures and temperatures.
The LCO provides RCS overpressure protection by having a minimum
coolant input capability and have adequate pressure relief
capability. Analyses have demonstrated that one PORV or an RCS vent
of at least 2.07 square inches is capable of limiting the RCS
pressure excursions below the 10 CFR 50, Appendix G limits for the
design basis LTOP limits.
The proposed change will not physically alter the plant (no new
or different type of equipment will be installed) or change the
methods governing normal plant operation. The proposed change does
not introduce new accident initiators or impact assumptions made in
the safety analysis. Testing requirements continue to demonstrate
that the LCOs are met and the system components are functional.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by this change. The proposed change will not result
in plant operation in a configuration outside the design basis.
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.
[[Page 28900]]
South Carolina Electric and Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: March 4, 2016. A publicly-available
version is in ADAMS under Accession No. ML16067A145.
Description of amendment request: The proposed changes, if
approved, would amend Combined License (COL) No. NPF-93 and NPR-94 for
the VCSNS. The requested amendment proposed changes would depart from
the approved AP1000 Design Control Document (DCD) ``Tier 2'' and ``Tier
2*'' information as currently incorporated into the VCSNS Updated Final
Safety Analysis Report (UFSAR). The changes relate to updating the
UFSAR text and tables; and information incorporated by reference
related to Westinghouse Electric Company's Reports WCAP-16096,
``Software Program Manual for Common QTM Systems,'' (also
known as the Common Q SPM) Revision 4, WCAP-16097, ``Common Qualified
Platform Topical Report,'' (also known as the Common Q Topical Report)
Revision 3, and WCAP-15927, ``Design Process for AP1000 Common Q Safety
Systems,'' Revision 4; and associated documents and references such as
a reference to the NRC's Regulatory Guide 1.152, ``Criteria for Use of
Computers in Safety Systems of Nuclear Power Plants'' (Revision 3, July
2011), and its associated exceptions. The proposed changes also include
removal of Tier 2* WCAP-17201-P, ``AC160 High Speed Link Communication
Compliance to DI&C-ISG-04 Staff Positions 9, 12, 13 and 15 Technical
Report,'' as a UFSAR incorporated by reference document.
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.
WCAP-16096 (Common Q Software Program Manual) was updated to
Revision 4 to reference later NRC endorsed regulatory guides and
standards and update the requirements for the software design and
development processes for the Common Q portion of the AP1000
Protection and Safety Monitoring System (PMS). WCAP-16097 (Common Q
Topical Report) was updated to Revision 3 to describe new Common Q
components and standards currently used for the AP1000 PMS
implementation of the Common Q platform. These two WCAPs have been
reviewed and approved by the NRC in Safety Evaluations dated
February 7, 2013. WCAP-15927 was updated to reference the newest
revisions of WCAP-16096 and WCAP-16097 and for editorial
corrections. The proposed activity adopts the updated versions as
incorporated by reference documents into the Updated Final Safety
Analysis Report. Other proposed document changes support the
implementation of the updated versions of WCAP-16096, WCAP-16097,
and WCAP-15927.
The Common Q platform is an acceptable platform for nuclear
safety-related applications. The Common Q system meets the
requirements of 10 CFR part 50, Appendix A, General Design Criteria
(Criteria 1, 2, 4, 13, 19, 20, 21, 22, 23, 24, and 25), the
Institute of Electrical and Electronics Engineers (IEEE) Standard
603-1991 for the design of safety-related reactor protection
systems, engineered safety features systems and other plant systems,
and the guidelines of Regulatory Guide 1.152 and supporting industry
standards for the design of digital systems.
Because the Common Q platform and the Protection and Safety
Monitoring System (PMS) implementation of the Common Q platform meet
the criteria in the applicable General Design Criteria, the
revisions to these documents do not affect the prevention and
mitigation of abnormal events, such as accidents, anticipated
operational occurrences, earthquakes, floods and turbine missiles,
or their safety or design analyses as described in the licensing
basis. The incorporation of the updated documents does not adversely
affect the interface with any structure, system, or component (SSC)
accident initiator or initiating sequence of events. Thus, the
probabilities of the accidents previously evaluated in the UFSAR are
not affected.
Therefore, the proposed activity 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 adopt the updated WCAP-16096, WCAP-
16097, and WCAP-15927 into the UFSAR do not adversely affect the
design or operation of safety-related equipment or equipment whose
failure could initiate an accident beyond what is already described
in the licensing basis. These changes do not adversely affect
fission product barriers. No safety analysis or design basis
acceptance limit/criterion is challenged or exceeded by the
requested change.
Therefore, this activity 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 adopt the updated WCAP-16096, WCAP-
16097, and WCAP-15927 into the UFSAR do not adversely affect the
design, construction, or operation of any plant SSCs, including any
equipment whose failure could initiate an accident or a failure of a
fission product barrier. No analysis is adversely affected by the
proposed changes. Furthermore, no system function, design function,
or equipment qualification will be adversely affected by the
changes.
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: Kathryn M. Sutton, Morgan, Lewis & Bockius
LLC, 1111 Pennsylvania Avenue NW, Washington, DC 20004-2514.
NRC Acting Branch Chief: John McKirgan.
South Carolina Electric and Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: March 14, 2016. A publicly-available
version is in ADAMS under Accession No. ML16075A264.
Description of amendment request: The proposed change would amend
the Combined License (COL) No. NPF-93 and NPF-94 for the VCSNS. The
requested amendment proposes to depart from approved AP1000 Design
Control Document (DCD) Tier 2 information (text, tables, and figures)
and involved Tier 2* information (as incorporated into the Updated
Final Safety Analysis Report as plant specific DCD information), and
also involves a change to the plant-specific Technical Specifications.
Specifically, the amendment request proposes changes to the plant-
specific AP1000 fuel system design, nuclear design, thermal hydraulic
design, and accident analyses as described in the licensing basis
documents. These proposed changes are consistent with those generically
approved in WCAP-17524-P-A, Revision 1, ``AP1000 Core Reference
Report.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
[[Page 28901]]
Response: No.
The proposed changes will revise the licensing basis documents
related to the fuel system design, nuclear design, thermal hydraulic
design, and accident analyses.
The UFSAR [Updated Final Safety Analysis Report] Chapter 15
accident analyses describe the analyses of various design basis
transients and accidents to demonstrate compliance of the AP1000
design with the acceptance criteria for these events. The acceptance
criteria for the various events are based on meeting the relevant
regulations, general design criteria, the Standard Review Plan, and
are a function of the anticipated frequency of occurrence of the
event and potential radiological consequences to the public. As
such, each design-basis event is categorized accordingly based on
these considerations. As discussed in Section 5.3 of WCAP-17524-P-A
Revision 1, the revised accident analyses maintain their plant
conditions, and thus their frequency designation and consequence
level as previously evaluated. As confirmed in the Safety Evaluation
Report (SER), the revised analyses meet the applicable guidelines in
the Standard Review Plan.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes will revise the licensing basis documents
related to the fuel system design, nuclear design, thermal hydraulic
design, and accident analyses.
The proposed changes would not introduce a new failure mode,
fault, or sequence of events that could result in a radioactive
material release. The proposed changes do not alter the design,
configuration, or method of operation of the plant beyond standard
functional capabilities of the equipment.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes will revise the licensing basis documents
related to the fuel system design, nuclear design, thermal hydraulic
design, and accident analyses.
Safety margins are applied at many levels to the design and
licensing basis functions and to the controlling values of
parameters to account for various uncertainties and to avoid
exceeding regulatory or licensing limits. UFSAR Subsection 4.1.1
presents the Principle Design Requirements imposed on the fuel and
control rod mechanism design to ensure that the performance and
safety criteria described in UFSAR Chapter 4 and Chapter 15 are met.
The revised fuel system design, nuclear design, thermal hydraulic
design, and accident analyses maintain the same Principle Design
Requirements, and further, satisfy the applicable regulations,
general design criteria, and Standard Review Plan. The effects of
the changes do not result in a significant reduction in margin for
any safety function, and were evaluated in the Safety Evaluation
Report for WCAP-17524-P-A Revision 1 and found to be acceptable.
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: Kathryn M. Sutton, Morgan, Lewis & Bockius
LLC, 1111 Pennsylvania Avenue NW, Washington, DC 20004-2514.
NRC Acting Branch Chief: John McKirgan.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: February 23, 2016. A publicly-available
version is in ADAMS under Accession No. ML16054A585.
Description of amendment request: The amendment would revise the
WBN Dual Unit Fire Protection Report and would revise the associated
License Condition regarding the WBN fire protection program.
Specifically, the amendment requests approval of a deviation from the
physical separation requirements of 10 CFR part 50, appendix R, section
III.G.2.d.
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.
A fire hazards analysis was performed for the areas under the
scope of this amendment. This fire hazards analysis demonstrates
that one train of safe shutdown equipment will remain functional in
the event of an Appendix R fire, even though a radiant energy shield
will not be provided for two raceway containing safe shutdown
circuits.
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.
A fire hazards analysis was performed for the areas under the
scope of this amendment. This fire hazards analysis demonstrates
that one train of safe shutdown equipment will remain functional in
the event of an Appendix R fire, even though a radiant energy shield
will not be provided for two raceway containing safe shutdown
circuits. Based on this, the proposed amendment will not alter the
requirements or function for systems required during accident
conditions.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
A fire hazards analysis was performed for the areas under the
scope of this amendment. This fire hazards analysis demonstrates
that one train of safe shutdown equipment will remain functional in
the event of an Appendix R fire, even though a radiant energy shield
will not be provided for two raceway containing safe shutdown
circuits.
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: Sherry A. Quirk, Executive Vice President
and General Counsel, Tennessee Valley Authority, 400 West Summit Hill
Drive, Knoxville, TN 37902.
NRC Branch Chief: Benjamin G. Beasley.
III. Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses and Combined Licenses,
Proposed No Significant Hazards Consideration Determination, and
Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
[[Page 28902]]
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: March 4, 2016. A publicly-available
version is in ADAMS under Accession No. ML16064A488.
Brief description of amendment request: The amendment would revise
the Cyber Security Plan implementation schedule for Milestone 8 and
would revise the associated license condition in the Facility Operating
License.
Date of publication of individual notice in Federal Register: April
19, 2016 (81 FR 23011).
Expiration date of individual notice: May 19, 2016 (public
comments); June 20, 2016 (hearing requests).
IV. 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.
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: September 24, 2015.
Brief description of amendment: The amendment revises Surveillance
Requirements (SRs) 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 changes
address the concerns discussed in NRC Generic Letter (GL) 2008-01,
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat
Removal, and Containment Spray Systems,'' as described in NRC-approved
Technical Specifications Task Force (TSTF)-523, Revision 2, ``Generic
Letter 2008-01, Managing Gas Accumulation.''
Date of issuance: April 20, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No.: 204. A publicly-available version is in ADAMS under
Accession. No. ML16069A006; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-43: This amendment revises the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 5, 2016 (81 FR
260).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 20, 2016.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: April 30, 2015, as supplemented by
letter dated February 19, 2016.
Brief description of amendments: The amendments approved adoption
of an emergency action level scheme based on Nuclear Energy Institute
(NEI) 99-01, Revision 6, ``Development of Emergency Action Levels for
Non-Passive Reactors,'' for the Catawba Nuclear Station, Units 1 and 2.
Date of issuance: April 18, 2016.
Effective date: As of the date of issuance and shall be implemented
by March 10, 2017.
Amendment Nos.: 279 for Unit 1 and 275 for Unit 2. A publicly-
available version is in ADAMS under Accession No. ML16082A038;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-35 and NPF-52: The
amendments revised the Renewed Facility Operating License.
Date of initial notice in Federal Register: June 23, 2015 (80 FR
35980). The supplemental letter dated February 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 April 18, 2016.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-369, 50-370, 50-413, and 50-
414, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North
Carolina and Catawba Nuclear Station, Units 1 and 2, York County, SC
Date of amendment request: June 23, 2015.
Brief description of amendments: The amendments remove superseded
TS requirements.
Date of issuance: April 8, 2016.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 283, 262, 278, and 274. A publicly-available
version is in ADAMS under Accession No. ML16060A229; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-9, NPF-17, NPF-35, and NPF-52:
Amendments revised the Facility Operating Licenses and Technical
Specifications.
Date of initial notice in Federal Register: August 4, 2015 (80 FR
46347).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 8, 2016.
No significant hazards consideration comments received: No.
Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, Wake and Chatham Counties, North Carolina
Date of amendment request: April 30, 2015, as supplemented by
letters dated November 19, 2015, and January 28, 2016.
Brief description of amendment: The amendment adopted the NRC-
endorsed
[[Page 28903]]
Nuclear Energy Institute (NEI) 99-01, Revision 6, ``Methodology for the
Development of Emergency Action Levels for Non-Passive Reactors.''
Date of issuance: April 13, 2016.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment No.: 149. A publicly-available version is in ADAMS under
Accession No. ML16057A838; documents related to this amendment are
listed in the Safety Evaluation (SE) enclosed with the amendment.
Facility Operating License No. NPF-63: The amendment revised the
Emergency Action Level Technical Bases document.
Date of initial notice in Federal Register: July 21, 2015 (80 FR
43128). The supplemental letters dated November 19, 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 amendment is contained
in an SE dated April 13, 2016.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-
286, Indian Point Nuclear Generating Unit Nos. 1, 2, and 3, Westchester
County, New York
Date of amendment request: June 16, 2015.
Brief description of amendments: The amendments revised the Cyber
Security Plan Milestone 8 full implementation date by extending the
full implementation date from June 30, 2016, to December 31, 2017.
Date of issuance: April 12, 2016.
Effective date: As of the date of issuance, and shall be
implemented within 30 days of issuance.
Amendment Nos.: 59 (Unit No. 1), 284 (Unit No. 2), and 260 (Unit
No. 3). A publicly-available version is in ADAMS under Accession No.
ML16064A215; documents related to these amendments are listed in the
Safety Evaluation enclosed with the amendments.
Provisional Operating License No. DPR-5 and Facility Operating
License Nos. DPR-26 and DPR-64: The amendments revised the Provisional
Operating License for Unit No. 1 and the Facility Operating Licenses
for Unit Nos. 2 and 3.
Date of initial notice in Federal Register: August 4, 2015 (80 FR
46348).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 2016.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland
Date of amendment request: November 5, 2015.
Brief description of amendments: The amendments revise the
Surveillance Requirement (SR) frequencies for SRs 3.4.6.4, 3.4.7.4,
3.4.8.3, 3.5.2.10, 3.6.6.9, 3.9.4.2, and 3.9.5.4. The changes to the SR
frequencies relocate the frequencies to the Surveillance Frequency
Control Program.
Date of issuance: April 11, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 317 and 295. A publicly-available version is in
ADAMS under Accession No. ML16060A401; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: January 5, 2016 (81 FR
261).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 11, 2016.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: March 23, 2015, as supplemented by
letters dated January 8, 2016, and March 21, 2016.
Brief description of amendment: The amendment revised the technical
specifications (TS) and relocated the secondary containment bypass
leakage paths table from the TS to the Technical Requirements Manual.
Date of issuance: April 19, 2016.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 156. A publicly-available version is in ADAMS under
Accession No. ML16088A053; documents related to this amendment is
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-69: Amendment revised
the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: September 29, 2015 (80
FR 58517). The supplemental letters dated January 8, 2016, and March
21, 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 April 19, 2016.
No significant hazards consideration comments received: No.
Florida Power & Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2 (PSL-2), St. Lucie County, Florida
Date of amendment request: December 30, 2014, as supplemented by
letters dated March 23, June 2, June 18, July 30, October 2, November
3, 2015; and December 8, 2015.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to allow the use of AREVA fuel and AREVA
M5[supreg] material as an approved fuel rod cladding at PSL-2.
Date of issuance: April 19, 2016.
Effective date: As of the date of issuance and shall be implemented
upon the start of the PSL-2 Cycle 23 spring 2017 refueling outage to
support the AREVA fuel transition project plan.
Amendment No.: 182. A publicly-available version is in ADAMS under
Accession No. ML16063A121; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-16: Amendment revised
the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: June 9, 2015 (80 FR
32620). The supplements dated June 2, June 18, July 30, October 2,
November 3, and December 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 amendment is contained
in a Safety Evaluation dated April 19, 2016.
No significant hazards consideration comments received: No.
[[Page 28904]]
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 application for amendments: June 26, 2013, as supplemented
by letters dated September 29, October 27, October 29, November 26, and
December 31, 2014; February 25 (two letters), May 7, October 15, and
December 31, 2015; and January 28, 2016.
Brief description of amendments: The amendments permit the PG&E
(the licensee) to adopt a new fire protection licensing basis based on
National Fire Protection Association (NFPA) Standard 805,
``Performance-Based Standard for Fire Protection for Light Water
Reactor Generating Plants (2001 Edition),'' at Diablo Canyon Power
Plant, Units 1 and 2, that complies with the requirements of 10 CFR
50.48(a) and (c) and the guidance in Revision 1 of Regulatory Guide
1.205, ``Risk Informed Performance-Based Fire Protection for Existing
Light-Water Nuclear Power Plants,'' December 2009.
Date of issuance: April 14, 2016.
Effective date: As of its date of issuance and shall be implemented
as described in the transition license conditions.
Amendment Nos.: Unit 1--225; Unit 2--227. A publicly-available
version is in ADAMS under Accession No. ML16035A441; 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 Technical Specifications.
Date of initial notice in Federal Register: December 26, 2013 (78
FR 78408). The supplemental letters dated October 3, 2013; September
29, October 27, October 29, November 26, and December 31, 2014;
February 25 (two letters), May 7, October 15, and December 31, 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 April 14, 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: September 1, 2015.
Brief description of amendment: The amendment authorized changes to
the VEGP Units 3 and 4 plant specific emergency planning inspections,
tests, analyses, and acceptance criteria (ITAAC) in Appendix C of VEGP
Units 3 and 4 Combined Operating Licenses (COLs). The changes authorize
the removal of the copy of Updated Final Safety Analysis Report Table
7.5-1, ``Post-Accident Monitoring System'' from ITAAC in Appendix C of
the VEGP Units 3 and 4 COLs.
Date of issuance: March 30, 2016.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 47. A publicly-available version is in ADAMS under
Accession No. ML16061A220; 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: October 27, 2015 (80 FR
65807).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 30, 2015.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia
Date of amendment request: January 13, 2015, as supplemented by
letters dated June 16 and November 24, 2015.
Brief description of amendments: The amendments adopt Technical
Specification Task Force change number 523, Revision 2, ``Generic
Letter 2008-01, Managing Gas Accumulation,'' for the Hatch Nuclear
Plant, Unit Nos 1 and 2, technical specifications. The change revised
or added 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.
Date of issuance: April 14, 2016.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 278 and 222. A publicly-available version is in
ADAMS under Accession No. ML16090A174; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: March 17, 2015 (80 FR
13911). The supplemental letters dated June 16 and November 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 April 14, 2016.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: September 23, 2015.
Brief description of amendment: The amendment revised the diesel
generator (DG) full load rejection test and endurance and margin test
specified by Technical Specification (TS) 3.8.1, ``AC [Alternating
Current] Sources--Operating,'' Surveillance Requirements (SR) 3.8.1.10
and 3.8.1.14, respectively. The change adds a new Note to SR 3.8.1.10
and SR 3.8.1.14, consistent with Technical Specification Task Force
(TSTF) traveler TSTF-276-A, Revision 2, ``Revise DG full load rejection
test.'' The Note allows the full load rejection test and endurance and
margin test to be performed at the specified power factor with
clarifications addressing situations when the power factor cannot be
achieved.
Date of issuance: April 15, 2016.
Effective date: As of its date of issuance and shall be implemented
within 90 days of issuance.
Amendment No.: 215. A publicly-available version is in ADAMS under
Accession No. ML16081A194; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 24, 2015 (80
FR 73242).
[[Page 28905]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 15, 2016.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 2nd day of May 2016.
For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2016-10949 Filed 5-9-16; 8:45 am]
BILLING CODE 7590-01-P