[Federal Register Volume 81, Number 70 (Tuesday, April 12, 2016)]
[Notices]
[Pages 21593-21607]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-08323]
=======================================================================
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2016-0073]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
-----------------------------------------------------------------------
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
[[Page 21594]]
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 March 15, 2016, to March 28, 2016. The last
biweekly notice was published on March 29, 2016.
DATES: Comments must be filed by May 12, 2016. A request for a hearing
must be filed by June 13, 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-2016-0073. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected].
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Lynn Ronewicz, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-1927, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2016-0073 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-2016-0073.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in the
SUPPLEMENTARY INFORMATION section of this document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2016-0073, facility name, unit
number(s), application date, and subject in your comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov, as well as entering the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment submissions into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated; (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
[[Page 21595]]
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 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) through (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 June
13, 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
June 13, 2016.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139;
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [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.
[[Page 21596]]
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC's public
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to
[email protected], or by a toll-free call at 1-866-672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, in some instances, a request to intervene will require
including information on local residence in order to demonstrate a
proximity assertion of interest in the proceeding. With respect to
copyrighted works, except for limited excerpts that serve the purpose
of the adjudicatory filings and would constitute a Fair Use
application, participants are requested not to include copyrighted
materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii).
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County, North
Carolina; Docket No. 50-261, H. B. Robinson Steam Electric Plant (RNP)
Unit No. 2, Darlington County, South Carolina; and Docket No. 50-400,
Shearon Harris Nuclear Power Plant (HNP), Unit 1, Wake and Chatham
Counties, North Carolina
Date of amendment request: February 1, 2016. A publicly-available
version is in ADAMS under Accession No. ML16040A077.
Description of amendment request: The amendments would change the
licensee's name from Duke Energy Progress, Inc. to Duke Energy
Progress, LLC.
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 changes do not involve a significant increase in
the probability of any accident previously evaluated because no
accident initiators or assumptions are affected. The proposed
conversion and name change is administrative in nature and has no
direct effect on any plant system, plant
[[Page 21597]]
personnel qualifications, or the operation and maintenance of BSEP,
RNP, and HNP.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated because the
proposed name change is administrative in nature and does not
involve new failure mechanisms, malfunctions, or accident
initiators. The proposed changes have no direct effect on any plant
system, plant personnel qualifications, or operation and maintenance
of BSEP, RNP, and HNP.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes will not involve a significant reduction in
the margin of safety because the proposed changes do not involve
changes to the initial conditions contributing to accident severity
or consequences, or reduce response or mitigation capabilities. The
proposed name change is administrative in nature and has no direct
effect on any plant system, plant personnel qualifications, or
operation and maintenance of BSEP, RNP, and HNP.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Deputy General Counsel,
Duke Energy Corporation, 550 South Tryon St., M/C DEC45A, Charlotte, NC
28202.
NRC Branch Chief: Benjamin G. Beasley.
Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana
Date of amendment request: October 29, 2015. A publicly-available
version is in ADAMS under Accession No. ML15307A293.
Description of amendment request: The amendment proposes to modify
Technical Specification (TS) 5.5.13, ``Primary Containment Leakage Rate
Testing Program,'' by incorporating Nuclear Energy Institute (NEI)
topical report 94-01, Revision 3-A, as the implementation document for
the RBS performance-based containment leakage rate testing program.
Based on the guidance in NEI 94-01, Revision 3-A, the proposed change
would allow the RBS Type A Test (Integrated Leak Rate Test) frequency
to be extended from 10 to 15 years, and the Type C Tests (Local Leak
Rate Tests) frequency to be extended from 60 to 75 months.
Additionally, the amendment proposes to modify Surveillance Requirement
(SR) 3.6.5.1.3 to extend the frequency of the Drywell Bypass Test from
10 to 15 years and to revise its allowed extension per SR 3.0.2 from 12
to 9 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review 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 incorporates NEI topical report 94-01,
Revision 3-A, into TS 5.5.13 as the basis for the RBS containment
leakage rate testing program, which would allow for extensions to
the frequencies of the Type A and Type C Tests. The proposed
amendment also requests an extension to the Drywell Bypass Test
frequency. The proposed changes do not involve any physical changes
to the plant or any changes in the normal operation or control of
the plant. In its license amendment request, the licensee identified
the loss-of-coolant accident (LOCA) inside containment and the fuel
handling accident (FHA) as the previously evaluated accidents in the
Updated Safety Analysis Report that could potentially be impacted by
the change. Changing the frequency of containment leakage rate
testing has no impact upon the likelihood of a LOCA or of an FHA.
Therefore, the probability of occurrence of an accident previously
evaluated is not significantly increased by the proposed amendment.
The guidelines in NEI 94-01, Revision 3-A, provide a framework
for a licensee's containment leakage rate testing program, the
purpose of which is to ensure that the primary containment limits
the uncontrolled release of radioactivity to the environment during
a design-basis accident. As part of its amendment request, the
licensee evaluated the potential consequences of extending the test
intervals and determined that the change in risk was estimated to be
acceptably small and within the guidelines, as published in
Regulatory Guide 1.174. The proposed amendment does not change the
overall containment leakage rate limit specified by the TSs.
Therefore, it is concluded that the proposed amendment does not
significantly increase the consequences of an accident previously
evaluated.
Based on the above discussion, 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 do not involve any physical changes to the
plant or any changes in the normal operation or control of the
plant. The proposed changes do not create any new accident
precursors or initiators, and do not change any existing accident
precursors or initiators, as described in the RBS safety analyses.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, for the development of the RBS performance-
based leakage rate testing program, to allow for frequency
extensions for the Type A and Type C Tests. The proposed amendment
also requests an extension to the Drywell Bypass Test frequency. The
proposed changes do not alter the manner in which safety limits,
limiting safety system setpoints, or limiting conditions for
operation are determined. The specific requirements and conditions
of the containment leakage rate testing program, as defined in the
TSs, ensure that the primary containment will continue to provide a
leaktight barrier to the uncontrolled release of radioactivity to
the environment during a design-basis accident. The proposed
amendment does not change the overall containment leakage rate limit
specified by the TSs. Additionally, the proposed amendment does not
include any changes to the Containment Inservice Inspection Plan at
RBS, which serves to provide a high degree of assurance that the
containment will not degrade in a manner that is not detectable by
the Type A Test.
Based on the above discussion, 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
its review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, LA 70113.
NRC Branch Chief: Meena K. Khanna.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station (PNPS), Plymouth County, Massachusetts
Date of amendment request: January 14, 2016. A publicly available
version is in ADAMS under Accession No. ML16021A459.
Description of amendment request: The amendment would revise the
PNPS Emergency Plan to decrease the Emergency Response Organization
(ERO) staff training requirements
[[Page 21598]]
identified for the ``on-site'' Chemistry Technician.
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, along with NRC edits in square
brackets:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed training requirements change has no effect on
normal plant operation or on any accident initiator. The change
affects the response to radiological emergencies addressed in the
SEP [site emergency plan]. The ability of the emergency response
organization to respond adequately to radiological emergencies has
been evaluated. Changes in the training provided to the on-shift
organization, such as the reassignment of key on-shift emergency
personnel to perform related RP [radiation protection] functions,
provide assurance of an effective emergency response without
competing or conflicting duties. An analysis was also performed on
the effect of the proposed change on the timeliness of performing
major tasks for the major functional areas of the SEP. The analysis
concluded that the reduction in training requirements for the ``on-
shift'' Chemistry Technician to support the initial RP support tasks
does not affect the ability to perform the required RP Technician or
Chemistry Technician tasks.
Therefore, the change in ERO staff training does not increase
the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change affects the training requirements for the
``on-shift'' Chemistry Technician and for supplementing onsite
personnel in response to a radiological emergency. It has been
evaluated and determined not to significantly affect the ability to
perform required or related functions. It has no effect on the plant
design or on the normal operation of the plant and does not affect
how the plant is physically operated under emergency conditions. The
reduction in ERO training requirements for the ``on shift''
Chemistry Technician in the SEP does not affect the plant operating
procedures which are performed by plant staff during all plant
conditions.
No new or different accidents are postulated to occur and there
are no changes in any of the accidents previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not affect plant design or method of
operation. 10 CFR 50.47(b) and 10 CFR 50 Appendix E establish
emergency planning standards and requirements that require adequate
staffing, satisfactory performance of key functional areas and
critical tasks, and timely augmentation of the response capability.
Since the SEP was originally developed, there have been improvements
in the technology used to support the SEP functions and in the
capabilities of onsite personnel. A functional analysis was
performed on the effect of the proposed change on the timeliness of
performing major tasks for the functional areas of the SEP. The
analysis concluded that a reduction in training requirements for the
``on-shift'' Chemistry Technician would not significantly affect the
ability to perform the required SEP tasks. Thus, the proposed change
has been determined not to adversely affect the ability to meet the
emergency planning standards as described in 10 CFR 50.47(b) and
requirements in 10 CFR 50 Appendix E.
The proposed ERO staff training change does not involve a
reduction in any margin of safety. The proposed change is consistent
with the original and current ERO staffing levels implemented at
PNPS.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
Date of amendment request: February 4, 2016. A publicly available
version is in ADAMS under Accession No. ML16035A227.
Description of amendment request: The amendments would add
Surveillance Requirement (SR) 3.5.2.10 to the list of applicable SRs
shown in SR 3.5.3.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with NRC staff corrections
shown in [brackets]:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The proposed LAR [license amendment request] is purely an
administrative change; therefore, the probability of any accident
previously evaluated is not significantly increased. The systems and
components required by the TS [technical specifications] for which
SR 3.5.2.10 is applicable, continue to be operable and capable of
performing any mitigation function assumed in the accident analysis.
As a result, the consequences of any accident previously evaluated
are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an[y] 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 LAR is purely an administrative change. The
proposed change to add SR 3.5.2.10 to the list of applicable
surveillances in SR 3.5.3.1 does not create a new or different kind
of accident [than] previously evaluated.
The change does not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. In addition,
the change does not impose any new or different requirements. The
change does not alter assumptions made in the safety analysis. The
proposed change is consistent with the safety analysis assumptions
and current plant operating practice.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed LAR is purely an administrative change to add SR
3.5.2.10 to the list of applicable surveillances in SR 3.5.3.1.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the Final Safety Analysis Report and Bases to TS).
Similarly, there is no impact to safety analysis acceptance criteria
as described in the plant licensing 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 Rd., Warrenville, IL
60555.
NRC Branch Chief: Travis L. Tate.
[[Page 21599]]
Exelon Generation Company, LLC (EGC), Docket No. 50-461, Clinton Power
Station (CPS), Unit No. 1, DeWitt County, Illinois
Date of amendment request: January 29, 2016. A publicly-available
version is in ADAMS under Accession No. ML16029A418.
Description of amendment request: The amendment would revise the
post-loss-of-coolant-accident (post-LOCA) drawdown time for secondary
containment from 12 to 19 minutes as described in the CPS Updated
Safety Analysis Report and technical specification bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change results in additional heat added to
Secondary Containment and the resultant increase in the time to
achieve and maintain the required negative pressure in Secondary
Containment following a LOCA. Neither the additional heat load from
DCS [dry-cask storage] activities, nor the resultant increase in the
time to achieve and maintain the required negative pressure in
Secondary Containment affect any initiator or precursor of any
accident previously evaluated. Therefore, the proposed change does
not involve a significant increase in the probability of an accident
previously evaluated.
The proposed change results in an increase in the post-LOCA
radiological dose to a Control Room occupant. However, the resultant
post-LOCA Control Room dose remains within the regulatory limits of
10 CFR 50.67 and GDC [General Design Criterion] 19. Therefore, the
proposed change does not involve a significant increase in the
consequences of an accident previously evaluated.
In summary, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the design function or
operation of Secondary Containment or the Standby Gas Treatment
system [SGTS], or the ability of each to perform its design
function. EGC has evaluated the post-LOCA pressure response of
Secondary Containment assuming the higher heat load, utilizing the
design basis short-term pressure response analysis. The results of
this analysis validated that SGTS will achieve and maintain the
required negative pressure in Secondary Containment within the
specified timeframe. The proposed change does not alter the safety
limits, or safety analysis associated with the operation of the
plant. Accordingly, the change does not introduce any new accident
initiators. Rather, this proposed change is the result of an
evaluation of the Control Room doses following the most limiting
LOCA that can occur at CPS. The proposed change does not introduce
any new modes of plant operation. As a result, no new failure modes
are introduced.
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 revised post-LOCA dose consequences to a Control Room
occupant were calculated in accordance with the requirements of 10
CFR 50.67, Regulatory Guide 1.183, and SRP [Standard Review Plan]
15.0.1 and are consistent with the post-LOCA dose calculations
approved by the NRC in Amendment No. 167 to the CPS Facility
Operating License NPF-62.
The margin of safety is considered to be that provided by
meeting the applicable regulatory limits. The additional heat load
that is added to Secondary Containment during DCS activities,
leading to an increase in Secondary Containment drawdown time
results in an increase in Control Room dose following the LOCA
design basis accident. However, since the Control Room dose
following the design basis accident remains within the regulatory
limits, there is not a significant reduction in a margin of safety.
Therefore, operation of CPS in accordance with the proposed
change will 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.
Acting NRC Branch Chief: Justin C. Poole.
FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
346, Davis-Besse Nuclear Power Station (DBNPS), Unit No. 1, Ottawa
County, Ohio
Date of amendment request: December 16, 2015, as supplemented by
letters dated February 2 and March 7, 2016. Publicly-available versions
are in ADAMS under Accession Nos. ML15350A314, ML16033A085, and
ML16067A195.
Description of amendment request: The amendment would allow the
licensee to transition the current fire protection program at DBNPS to
a performance-based, risk-informed fire protection program consistent
with 10 CFR, Section 50.48(c), ``National Fire Protection Association
Standard NFPA 805.'' The 2001 Edition of NFPA 805, ``Performance-Based
Standard for Fire Protection for Light Water Reactor Electric
Generating Plants,'' is incorporated by reference into 10 CFR 50.48(c),
with exceptions, modifications, and supplementation. The amendment
would also allow the licensee to make changes to the DBNPS fire
protection program without prior NRC approval, provided that specified
conditions are met. The proposed amendment would change the facility
operating license, technical specifications, and design basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Operation of DBNPS in accordance with the proposed amendment
does not increase the probability or consequences of accidents
previously evaluated. The Updated Final Safety Analysis Report
(UFSAR) documents the analyses of design basis accidents (DBAs) at
DBNPS. The proposed amendment does not affect accident initiators,
nor does it alter design assumptions, conditions, or configurations
of the facility that would increase the probability of accidents
previously evaluated. Further, the changes to be made for fire
hazard protection and mitigation do not adversely affect the ability
of SSCs [structures, systems, and components] to perform their
design functions for accident mitigation, nor do they affect the
postulated initiators or assumed failure modes for accidents
described and evaluated in the UFSAR. SSCs required to shut down the
reactor safely and to maintain it in a safe and stable condition
will remain capable of performing their design functions.
The purpose of the proposed amendment is to permit DBNPS to
adopt a new fire protection licensing basis, which complies with the
requirements of 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance
in [Regulatory Guide] RG 1.205, Revision 1. The NRC considers that
NFPA 805 provides an acceptable methodology and performance criteria
for licensees to identify fire protection requirements that are an
acceptable alternative to the 10 CFR 50, Appendix R required fire
protection features (69 FR 33536, June 16, 2004). Engineering
analyses, which may include engineering evaluations, probabilistic
safety assessments, and fire modeling calculations, have been
performed to demonstrate that the
[[Page 21600]]
performance-based requirements of NFPA 805 have been satisfied.
NFPA 805, taken as a whole, provides an acceptable alternative
for satisfying General Design Criterion 3 (GDC 3) of Appendix A to
10 CFR 50, meets the underlying intent of the NRC's existing fire
protection regulations and guidance, and provides for DID [defense-
in-depth]. The goals, performance objectives, and performance
criteria specified in Chapter 1 of the standard ensure that, if
there are any increases in CDF [core damage frequency] or risk, the
increase will be small and consistent with the intent of the
Commission's Safety Goal Policy.
Based on this, the implementation of the proposed amendment does
not increase the probability of any accident previously evaluated.
Equipment required to mitigate an accident remains capable of
performing the assumed function(s). The proposed amendment will not
affect the source term, containment isolation, or radiological
release assumptions used in evaluating the radiological consequences
of any accident previously evaluated. The applicable radiological
dose criteria will continue to be met. Therefore, the consequences
of any accident previously evaluated are not significantly increased
with the implementation of the proposed amendment.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Operation of DBNPS in accordance with the proposed amendment
does not create the possibility of a new or different kind of
accident from any accident previously evaluated. The proposed change
does not alter the requirements or functions for systems required
during accident conditions. Implementation of the new fire
protection licensing basis that complies with the requirements of 10
CFR 50.48(a) and 10 CFR 50.58(c) and the guidance in RG 1.205,
Revision 1, will not result in new or different accidents.
The proposed amendment does not adversely affect accident
initiators or alter design assumptions, conditions, or
configurations of the facility. The proposed amendment does not
adversely affect the ability of SSCs to perform their design
function. SSCs required to maintain the plant in a safe and stable
condition remain capable of performing their design functions.
The proposed amendment does not introduce new or different
accident initiators, nor does it alter design assumptions,
conditions, or configurations of the facility. The proposed
amendment does not adversely affect the ability of SSCs to perform
their design function. SSCs required to safely shutdown the reactor
and maintain it in a safe and stable condition remain capable of
performing their design functions.
The purpose of the proposed amendment is to permit DBNPS to
adopt a new fire protection licensing basis that complies with the
requirements of 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance
in Regulatory Guide 1.205, Revision 1. The NRC considers that NFPA
805 provides an acceptable methodology and appropriate performance
criteria for licensees to identify fire protection systems and
features that are an acceptable alternative to the 10 CFR 50,
Appendix R required fire protection features (69 FR [Federal
Register] 33536, June 16, 2004).
The requirements of NFPA 805 address only fire protection and
the impacts of fire on the plant that have previously been
evaluated. Based on this, implementation of the proposed amendment
would not create the possibility of a new or different kind of
accident from any kind of accident previously evaluated. No new
accident scenarios, transient precursors, failure mechanisms, or
limiting single failures will be introduced as a result of this
amendment. There will be no adverse effect or challenges imposed on
any safety-related system as a result of this amendment. Therefore,
the possibility of a new or different kind of accident from any kind
of accident previously evaluated is not created with the
implementation of this amendment.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Operation of DBNPS in accordance with the proposed amendment
does not involve a significant reduction in the margin of safety.
The proposed amendment does not alter the manner in that safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not affected by this change. The proposed amendment does not
adversely affect existing plant safety margins or the reliability of
equipment assumed to mitigate accidents in the UFSAR. The proposed
amendment does not adversely affect the ability of SSCs to perform
their design function. SSCs required to safely shut down the reactor
and to maintain it in a safe and stable condition remain capable of
performing their design functions.
The purpose of the proposed amendment is to permit FENOC to
adopt a new fire protection licensing basis which complies with the
requirements in 10 CFR 50.48(a) and 10 CFR 50.48(c) and the guidance
in RG 1.205, Revision 1. The NRC considers that NFPA 805 provides an
acceptable methodology and performance criteria for licensees to
identify fire protection systems and features that are an acceptable
alternative to the 10 CFR 50 Appendix R required fire protection
features (69 FR 33536, June 16, 2004). Engineering analyses, which
may include engineering evaluations, probabilistic safety
assessments, and fire modeling calculations, have been performed to
demonstrate that the performance-based requirements of NFPA 805 do
not result in a significant reduction in the margin of safety.
The proposed changes are evaluated to ensure that risk and
safety margins are kept within acceptable limits. Therefore, the
transition to NFPA 805 does not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
Acting NRC Branch Chief: Justin C. Poole.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 11, 2016. A publicly-available
version is in ADAMS under Accession No. ML16076A433.
Description of amendment request: The amendment would adopt
Technical Specification (TS) Task Force (TSTF) Change Traveler TSTF-
535, Revision 0, ``Revise Shutdown Margin [SDM] Definition to Address
Advanced Fuel Designs.'' The SDM (i.e., the amount of reactivity by
which the reactor is subcritical), is calculated under the conservative
conditions that the reactor is Xenon free, the most reactive control
rod is outside the reactor core, and the moderator temperature produces
the maximum reactivity. For standard fuel designs, maximum reactivity
occurs at a moderator temperature of 68 degrees Fahrenheit ([deg]F),
which is reflected in the temperature specified in the TSs. New,
advanced boiling water reactor fuel designs can have a higher
reactivity at moderator shutdown temperatures above 68[emsp14][deg]F.
Therefore, the proposed amendment, consistent with TSTF-535, Revision
0, seeks to modify the TSs to require the SDM to be calculated at
whatever temperature produces the maximum reactivity (i.e.,
temperatures at or above 68[emsp14][deg]F). The availability of this TS
improvement was announced in the Federal Register (FR) published on
February 26, 2013 (78 FR 13100), as part of the Consolidated Line Item
Improvement Process, and has been requested with no variations or
deviations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, 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 the definition of SDM. SDM is not an
initiator to any accident previously evaluated. Accordingly, the
proposed change to the definition of SDM has no effect on the
probability of any
[[Page 21601]]
accident previously evaluated. SDM is an assumption in the analysis
of some previously evaluated accidents and inadequate SDM could lead
to an increase in consequences for those accidents. However, the
proposed change revises the SDM definition to ensure that the
correct SDM is determined for all fuel types at all times during the
fuel cycle. As a result, the proposed change does not adversely
affect the consequences of any accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the definition of SDM. The change
does not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operations. The change does not alter
assumptions made in the safety analysis regarding SDM.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the definition of SDM. The proposed
change does not alter the manner in which safety limits, limiting
safety system settings or limiting conditions for operation are
determined. The proposed change ensures that the SDM assumed in
determining safety limits, limiting safety system settings or
limiting conditions for operation is correct for all Boiling Water
Reactor fuel types at all times during the fuel cycle.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, P.O. Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Meena K. Khanna.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1,
Fairfield County, South Carolina
Date of amendment request: December 16, 2015, as supplemented by
letter dated March 7, 2016. Publicly-available versions are in ADAMS
under Accession Nos. ML15356A048 and ML16069A021, respectively.
Description of amendment request: The licensee proposes to revise
TS 3/4.3.1, ``Reactor Trip System Instrumentation,'' and TS 3/4.3.2,
``Engineered Safety Feature Actuation System Instrumentation,'' to
implement the Allowed Outage Time, Bypass Test Time, and Surveillance
Frequency changes approved by the NRC in WCAP-15376-P-A, Rev. 1,
``Risk-Informed Assessment of the Reactor Trip System (RTS) and
Engineered Safety Features Actuation System (ESFAS) Surveillance Test
Intervals and Reactor Trip Breaker Test and Completion Times.''
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 overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed. The same reactor trip system (RTS)
and engineered safety feature actuation system (ESFAS)
instrumentation will continue to be used. The protection systems
will continue to function in a manner consistent with the plant
design basis. These changes to the Technical Specifications do not
result in a condition where the design, material, and construction
standards that were applicable prior to the change are altered.
The proposed changes will not modify any system interfaces. The
proposed changes will not affect the probability of any event
initiators. There will be no degradation in the performance of or an
increase in the number of challenges imposed on safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance. The proposed changes will not alter any
assumptions or change any mitigation actions in the radiological
consequence evaluations in the Final Safety Analysis Report (FSAR).
The determination that the results of the proposed changes are
acceptable was established in the NRC Safety Evaluation prepared for
WCAP-1 5376-P-A (issued by letter dated December 20, 2002
[ML023540534]). Implementation of the proposed changes will result
in an insignificant risk impact. Applicability of these conclusions
has been verified through plant-specific reviews and implementation
of the generic analysis results in accordance with the NRC Safety
Evaluation conditions.
The proposed changes to the Completion Times, bypass test times,
and Surveillance Frequencies reduce the potential for inadvertent
reactor trips and spurious engineered safety feature (ESF)
actuations, and therefore do not increase the probability of any
accident previously evaluated. The proposed changes do not change
the response of the plant to any accidents and have an insignificant
impact on the reliability of the RTS and ESFAS signals. The RTS and
ESFAS instrumentation will remain highly reliable and the proposed
changes will not result in a significant increase in the risk of
plant operation. This is demonstrated by showing that the impact on
plant safety as measured by the increase in core damage frequency
(CDF) is less than 1.0E-06 per year and the increase in large early
release frequency (LERF) is less than 1.0E-07 per year. In addition,
for the Completion Time changes, the incremental conditional core
damage probabilities (ICCDP) and incremental conditional large early
release probabilities (ICLERP) are less than 5.0E-07 and 5.0E-08,
respectively. These changes meet the acceptance criteria in
Regulatory Guides 1.174 and 1.177. Therefore, since the RTS and
ESFAS instrumentation will continue to perform their functions with
high reliability as originally assumed, and the risk impact as
measured by the [Delta]CDF, [Delta]LERF, ICCDP, and ICLERP risk
metrics is within the acceptance criteria of existing regulatory
guidance, there will not be a significant increase in the
consequences of any accidents.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems, and components (SSCs)
from performing their intended function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
The proposed changes are consistent with safety analysis assumptions
and resultant consequences.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. The proposed changes will not affect the normal method of
plant operation. No performance requirements will be affected or
eliminated.
The proposed changes will not result in physical alteration to
any plant system nor will there be any change in the method by which
any safety-related plant system performs its safety function. The
proposed changes do not include any changes to the instrumentation
setpoints or changes to the accident analysis assumptions.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting
[[Page 21602]]
single failures are introduced as a result of these changes. There
will be no adverse effect or challenges imposed on any safety-
related system as a result of these changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Analysis
Limit (SAL). There will be no effect on the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined nor will there be any effect on those plant
systems necessary to assure the accomplishment of protection
functions.
The redundancy of RTS and ESFAS is maintained, and diversity
with regard to the signals that provide reactor trip and ESF
actuation is also maintained. All signals credited as primary or
secondary, and all operator actions credited in the accident
analyses will remain the same. The proposed changes will not result
in plant operation in a configuration outside the design basis. The
calculated impact on risk is insignificant and meets the acceptance
criteria contained in Regulatory Guides 1.174 and 1.177. Although
there was no attempt to quantify any positive human factors benefit
due to increased Completion Times and bypass test times, it is
expected that there would be a net benefit due to a reduced
potential for spurious reactor trips and actuations associated with
testing.
Implementation of the proposed changes is expected to result in
an overall improvement in safety, as follows:
(a) Reduced testing should result in fewer inadvertent reactor
trips, less frequent actuation of ESFAS components, less frequent
distraction of operations personnel without significantly affecting
RTS and ESFAS reliability.
(b) The Completion Time extensions for the reactor trip breakers
should provide additional time to complete test and maintenance
activities while at power, potentially reducing the number of forced
outages related to compliance with reactor trip breaker Completion
Times, and provide consistency with the Completion Times for the
logic trains.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. Hagood Hamilton, Jr., South Carolina
Electric & Gas Company, P.O. Box 764, Columbia, SC 29218.
NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke
County, Georgia
Date of amendment request: February 15, 2016. A publicly-available
version is in ADAMS under Accession No. ML16046A009.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-91 and NPF-92 for the VEGP Units 3 and 4. The
requested amendment proposes changes to the Updated Final Safety
Analysis Report (UFSAR) in the form of departures from the incorporated
plant-specific Design Control Document Tier 2 information and involves
related changes to the associated plant-specific Tier 2* information.
Specifically, the proposed departures consist of changes to UFSAR text
and tables, and information incorporated by reference into the UFSAR
related to updates to WCAP-16096, ``Software Program Manual for Common
QTM Systems,'' and WCAP-16097, ``Common Qualified Platform
Topical 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?
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 UFSAR. 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 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 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 accident initiator or initiating sequence of events. Thus,
the probabilities of the accidents previously evaluated in the UFSAR
are not affected.
Therefore, the proposed amendment does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed 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 accident 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.
[[Page 21603]]
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
Acting NRC Branch Chief: John McKirgan.
Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328,
Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: March 11, 2016. A publicly-available
version is in ADAMS under Accession No. ML16071A333.
Description of amendment request: The amendments would revise the
Technical Specifications to add a new condition to extend the allowed
completion time to restore one Essential Raw Cooling Water train to
OPERABLE status from 72 hours to 7 days for planned maintenance, when
the opposite unit is defueled or in Mode 6, following defueling under
certain restrictions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
1. Does the proposed amendment involve a significant increase in
the probability or consequence of an accident previously evaluated?
Response: No.
The proposed change adds new Condition A to Technical
Specification (TS) 3.7.8, Essential Raw Cooling Water (ERCW) System
for Sequoyah Nuclear Plant (SQN) Units 1 and 2. The proposed change
will extend the allowed completion time to restore ERCW System train
to OPERABLE status from 72 hours to 7 days for planned maintenance
when the opposite unit is defueled or in mode 6 following defueled
with refueling water cavity level >= [greater than or equal to] 23
ft. above top of reactor vessel flange and UHS [ultimate heat sink]
Temperature is <= [less than or equal to 79 degrees F. This change
does not result in any physical changes to plant safety-related
structures, systems, or components (SSCs). The UHS and associated
ERCW system function is to remove plant system heat loads during
normal and accident conditions. As such, the UHS and ERCW system are
not design basis accident initiators, but instead perform accident
mitigation functions by serving as the heat sink for safety-related
equipment to ensure the conditions and assumptions credited in the
accident analyses are preserved. During operation under the proposed
change with one ERCW train inoperable, the other ERCW train will
continue to perform the design function of the ERCW system.
Therefore, the proposed change does not involve a significant
increase in the probability of an accident previously evaluated.
Accordingly, as demonstrated by TVA design heat transfer and
flow modeling calculations, operation with one ERCW System
inoperable for 7 days for planned maintenance when the opposite unit
is defueled or in mode 6 following defueled with refueling water
cavity level >= 23 ft. above top of reactor vessel flange, the fuel
cladding, Reactor Coolant System (RCS) pressure boundary, and
containment integrity limits are not challenged during worst-case
post-accident conditions. Accordingly, the conclusions of the
accident analyses will remain as previously evaluated such that
there will be no significant increase in the post-accident dose
consequences.
Therefore, the proposed change does not involve a significant
increase in the probability or consequence 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 any physical changes to
plant safety related SSCs or alter the modes of plant operation in a
manner that is outside the bounds of the current UHS and ERCW system
design heat transfer and flow modeling analyses. The proposed change
to add new Condition A to TS 3.7.8, ERCW System, which would extend
the allowed completion time to restore ERCW System train to OPERABLE
status from 72 hours to 7 days for planned maintenance when the
opposite unit is defueled or in mode 6 following defueled with
refueling water cavity level >= 23 ft. above top of reactor vessel
flange. Thus, although the specified ERCW system alignments result
in reduced heat transfer flow capability, the plant's overall
ability to reject heat to the UHS during normal operation, normal
shutdown, and hypothetical worst-case accident conditions will not
be significantly affected by this proposed change. Because the
safety and design requirements continue to be met and the integrity
of the RCS pressure boundary is not challenged, no new credible
failure mechanisms, malfunctions, or accident initiators are
created, and there will be no effect on the accident mitigating
systems in a manner that would significantly degrade the plant's
response to an accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to add new Condition A to TS 3.7.8, ERCW
System, which would extend the allowed completion time to restore
ERCW System train to OPERABLE status from 72 hours to 7 days for
planned maintenance when the opposite unit is defueled or in mode 6
following defueled with refueling water cavity level >= 23 ft. above
top of reactor vessel flange. As demonstrated by TVA design basis
heat transfer and flow modeling calculations, the design limits for
fuel cladding, RCS pressure boundary, and containment integrity are
not exceeded under both normal and post-accident conditions. As
required, these calculations include evaluation of the worst-case
combination of meteorology and operational parameters, and establish
adequate margins to account for measurement and instrument
uncertainties. While operating margins have been reduced by the
proposed change in order to support necessary maintenance
activities, the current limiting design basis accidents remain
applicable and the analyses conclusions remain bounding such that
the accident safety margins are maintained. Accordingly, the
proposed change will not significantly degrade the margin of safety
of any SSCs that rely on the UHS and ERCW system for heat removal to
perform their safety related functions.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Benjamin G. Beasley.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: January 27, 2016. A publicly-available
version is in ADAMS under Accession No. ML16033A470.
Description of amendment request: The amendment would revise the
Technical Specifications to allow the use of Optimized ZIRLO\TM\ as an
approved fuel rod cladding.
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 would allow the use of Optimized ZIRLO\TM\
clad nuclear fuel in the reactor. The NRC approved topical report
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, 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. Use of Optimized ZIRLO\TM\ fuel cladding will
[[Page 21604]]
continue to meet the 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 ZIRLOTM 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 accident 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 because it has been 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 in 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.
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: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
NRC Branch Chief: Robert J. Pascarelli.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
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: June 30, 2015, as supplemented by
letters dated August 11, 2015; September 24, 2015; October 8, 2015;
December 7, 2015; February 10, 2016; and February 25, 2016.
Brief description of amendments: The amendments revised selected
Technical Specification Completion Times to support repair activity
associated with the Nuclear Service Water System, Train `A'.
Date of issuance: March 16, 2016.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 282 and 261. A publicly-available version is in
ADAMS under Accession No. ML15306A141; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Facility Operating License Nos. NPF-9 and NPF-18: Amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: August 20, 2015 (80 FR
50663). The supplemental letters dated August 11, 2015; September 24,
2015; October 8, 2015; December 7, 2015; February 10, 2016; and
February 25, 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 March 16, 2016.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station (VY),
Vernon, Vermont
Date of amendment request: June 24, 2015.
Brief description of amendment request: The amendment changed the
VY Cyber Security Plan Implementation Schedule Milestone 8 full
implementation date of June 30, 2016, to December 15, 2017. The
amendment also revised the existing Renewed Facility Operating License
Security Plan license condition.
Date of issuance: March 14, 2016.
Effective date: As of the date of issuance, and shall be
implemented by June 30, 2015.
Amendment No.: 265. A publicly-available version is in ADAMS under
Accession No. ML16014A169; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-28: The amendment
revised the Facility Operating License.
Date of initial notice in Federal Register: September 8, 2015 (80
FR 53900).
The Commission's related evaluation of this amendment is contained
in the Safety Evaluation dated March 14, 2016.
No significant hazards consideration comments received: No.
[[Page 21605]]
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland
Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine
Mile Point Nuclear Station, Units 1 and 2, Oswego County, New York
Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear
Power Plant, Wayne County, New York
Date of amendment request: July 29, 2015, as supplemented by letter
dated November 4, 2015.
Brief description of amendments: The amendments revised the
emergency plan definition of annual training frequency to ``once per
calendar year not to exceed 18 months between training sessions.''
Date of issuance: March 17, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 316/294; 221/155; and 121. A publicly-available
version is in ADAMS under Accession No. ML15352A164; documents related
to these amendments are listed in the safety evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-53, DPR-69, DPR-63,
NPF-69, and DPR-18: The amendments revised the emergency plans.
Date of initial notice in Federal Register: December 8, 2015 (80 FR
76320).
The Commission's related evaluation of the amendments is contained
in a safety evaluation dated March 17, 2016.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-352, Limerick Generating
Station (LGS), Unit 1, Montgomery County, Pennsylvania
Date of amendment request: November 19, 2015.
Brief description of amendment: The amendment revised the technical
specifications (TSs) related to the safety limit minimum critical power
ratios. The changes resulted from a cycle-specific analysis performed
to support the operation of LGS, Unit 1, in the upcoming Cycle 17.
Date of issuance: March 15, 2016.
Effective date: As of the date of issuance and shall be implemented
prior to startup from the spring 2016 refueling outage.
Amendment No.: 221. A publicly-available version is in ADAMS under
Accession No. ML16041A021; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-39: Amendment revised
the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: January 5, 2016 (81 FR
275).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 15, 2016.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277
and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and
Lancaster Counties, Pennsylvania
Date of amendment request: September 4, 2014, as supplemented by
letters dated January 29, February 6, April 28, July 6, September 4,
October 1, and October 26, 2015, and January 15, 2016.
Brief description of amendments: The amendments changed the
Technical Specifications (TSs) and Renewed Facility Operating Licenses
(RFOLs) to allow plant operation from the currently licensed Maximum
Extended Load Line Limit Analysis (MELLLA) domain to plant operation in
the expanded MELLLA Plus (MELLLA+) domain under the previously approved
extended power uprate conditions of 3,951 megawatts thermal rated core
thermal power. The expanded MELLLA+ operating domain increases
operating flexibility by allowing control of reactivity at maximum
power by changing flow rather than by control rod insertion and
withdrawal.
Date of issuance: March 21, 2016.
Effective date: As of the date of issuance and shall be implemented
within 1 year of issuance.
Amendments Nos.: 305 and 309. A publicly-available version is in
ADAMS under Accession No. s; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
RFOL Nos. DPR-44 and DPR-56: The amendments revised the RFOLs and
TSs.
Date of initial notice in Federal Register: December 2, 2014 (79 FR
71454). The supplemental letters dated January 29, February 6, April
28, July 6, September 4, October 1, and October 26, 2015, and January
15, 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 March 21, 2016.
No significant hazards consideration comments received: Yes.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: September 11, 2014, as supplemented by
letters dated October 15, 2014, and December 18, 2014.
Description of amendment: The amendments revised the Updated Final
Safety Analysis Report by clarifying how human diversity was applied
during the design process for the Component Interface Module and
Diverse Actuation System.
Date of issuance: July 17, 2015.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 28. A publicly-available version is in ADAMS under
Accession No. ML15176A703; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Combined License Nos. NPF-93 and NPF-94: Amendments
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: December 9, 2014 (79 FR
73111). The supplemental letters dated October 15, 2014, and December
18, 2014, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 17, 2015.
No significant hazards consideration comments received: No.
South Carolina Electric and Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: February 10, 2015.
Brief description of amendment: The amendments revised the VCSNS
Units 2 and 3 Updated Final Safety Analysis Report (UFSAR) by revising
the references to human factors-related plans. The UFSAR-referenced
plans are the Human Factors Engineering Design Verification plan, Task
Support Verification plan, and the Integrated
[[Page 21606]]
System Validation plan. The UFSAR references to those plans required an
update to the latest version of those plans due to changes within the
plans. The amendments involved changes to the approved VCSNS Units 2
and 3 UFSAR Tier 2* information, as defined in 10 CFR part 52, appendix
D, section II.F.
Date of issuance: September 23, 2015.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 33. A publicly-available version is in ADAMS under
Accession No. ML15189A363; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Combined License Nos. NPF-93 and NPF-94: Amendments
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: March 31, 2015 (80 FR
17094). The supplemental letter dated March 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 amendments is contained
in a Safety Evaluation dated September 23, 2015.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: August 24, 2015.
Brief description of amendment: The amendments authorized changes
to the VCSNS Units 2 and 3 Updated Final Safety Analysis Report Tier 2
and Tier 2* information to revise the seismic Category I and II
structures containing mechanical couplers welded to structural steel
utilizing combined partial joint penetration weld with fillet weld
reinforcement with fillet welds satisfying the minimum size
requirements for C2/C3J couplers to demonstrate the capacity required
by code is established by appropriate analyses and testing.
Date of issuance: November 12, 2015.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 36. A publicly-available version is in ADAMS under
Accession No. ML15301A100; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Combined License Nos. NPF-93 and NPF-94: Amendments
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: September 3, 2015 (80
FR 53336). The supplemental letters dated September 23, 2015, and
October 1, 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 November 12, 2015.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield
County, South Carolina
Date of amendment request: October 22, 2015.
Brief description of amendment: The amendments authorized changes
to the VCSNS Combined Licenses (COLs). Specifically, the changes were
to VCSNS Units 2 and 3 COLs, Appendix A, Technical Specifications,
Section 5.0, ``Administrative Controls,'' by revising the title ``Shift
Supervisor'' to ``Shift Manager.''
Date of issuance: February 29, 2016.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 42. A publicly-available version is in ADAMS under
Accession No. ML16042A476; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Combined License Nos. NPF-93 and NPF-94: Amendment revised
the Facility Combined Licenses.
Date of initial notice in Federal Register: November 24, 2015 (80
FR 73242).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 29, 2016.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of amendment request: May 12, 2015, as supplemented by letters
dated September 21, 2015; November 25, 2015; and January 28, 2016.
Brief description of amendments: The amendments revised and added
Surveillance Requirements to verify that the system locations
susceptible to gas accumulation are sufficiently filled with water and
to provide allowances that permit performance of the verification. The
changes are consistent with TSTF-523, Revision 2, ``Generic Letter
2008-01, Managing Gas Accumulation.''
Date of issuance: March 21, 2016.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 178 (Unit 1) and 159 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16063A475, documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-2 and NPF-8: The amendments
revised the Renewed Facility Operating Licenses and Technical
Specifications.
Date of initial notice in Federal Register: June 23, 2015 (80 FR
35984). The supplemental letters dated September 21, 2015; November 25,
2015; and January 28, 2016, provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 21, 2016.
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: April 2, 2015, as supplemented by
letters dated November 12, 2015, and February 9, 2016.
Brief description of amendments: The amendments revised the
technical specifications (TSs) as necessary to relocate the pressure
and temperature (P-T or P/T) limit curves and associated references to
a pressure and temperature limits report (PTLR). Specifically, the
request modified Section 1.0, ``Definitions''; Limiting Conditions for
Operation and Surveillance Requirement Applicability Section 3.4.9,
``RCS Pressure and Temperature
[[Page 21607]]
(P/T) Limits''; and Section 5.0, ``Administrative Controls,'' of the
TSs for both units to delete reference to the P-T curves and to include
reference to the unit-specific PTLRs. The amendments also implemented
new P-T limits for both units.
Date of issuance: March 23, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 277 and 221. A publicly-available version is in
ADAMS under Accession No. ML16062A099; 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 TSs.
Date of initial notice in Federal Register: July 7, 2015 (80 FR
38760). The supplemental letters dated November 12, 2015, and February
9, 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 March 23, 2016.
No significant hazards consideration comments received: No.
Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna
Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: March 19, 2015, as supplemented by
letters dated October 15, 2015; October 16, 2015; and January 8, 2016.
A publicly-available version is in ADAMS under Accession Nos.
ML15091A657, ML15296A048, ML15296A057, and ML16011A103, respectively.
Brief description of amendments: The amendments revised the
Emergency Plan for the Susquehanna Steam Electric Station (SSES) to
adopt the Nuclear Energy Institute's (NEI's) revised Emergency Action
Level scheme described in NEI 99-01, Revision 6, ``Development of
Emergency Action Levels for Non-Passive Reactors'' (ADAMS Accession No.
ML12326A805), which was endorsed by the NRC as documented in NRC letter
dated March 28, 2013 (ADAMS Accession No. ML12346A463). This request
was submitted by PPL Susquehanna, LLC; however, on June 1, 2015 (ADAMS
Accession No. ML15054A066), the NRC staff issued an amendment changing
the name on the SESS license from PPL Susquehanna, LLC to Susquehanna
Nuclear, LLC. This amendment was issued subsequent to an order issued
on April 10, 2015 (ADAMS Accession No. ML15058A073), to SSES, approving
an indirect license transfer of the SESS license to Talen Energy
Corporation.
Date of issuance: March 28, 2016.
Effective date: As of the date of issuance and shall be implemented
on or before December 31, 2016.
Amendment Nos.: 265 (Unit 1) and 246 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16062A216; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Facility Operating Licenses.
Date of initial notice in Federal Register: July 7, 2015 (80 FR
38762). The supplemental letters dated October 15, 2015; October 16,
2015; and January 8, 2016, provided additional information that
clarified the application and expanded the scope of the application as
originally noticed, and changed the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register. As such, the NRC staff published a subsequent notice
in the Federal Register on February 2, 2016 (81 FR 5500).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 28, 2016.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 1st day of April 2016.
For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2016-08323 Filed 4-11-16; 8:45 am]
BILLING CODE 7590-01-P