[Federal Register Volume 81, Number 168 (Tuesday, August 30, 2016)]
[Notices]
[Pages 59659-59669]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-20391]
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NUCLEAR REGULATORY COMMISSION
[NRC-2016-0180]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954,
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued, and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued, from August 2, 2016, to August 15, 2016. The
last biweekly notice was published on August 16, 2016.
DATES: Comments must be filed by September 29, 2016. A request for a
hearing must be filed by October 31, 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-2016-0180. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-5411, email: [email protected].
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2016-0180, facility name, unit
number(s),
[[Page 59660]]
plant docket number, application date, and subject 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-2016-0180.
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 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-2016-0180, facility name, unit
number(s), plant docket number, application date, and subject in your
comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment submissions into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period if 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. If the Commission
takes 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. If the Commission makes a final no significant hazards
consideration determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license or
combined license. Requests for a hearing and a petition for leave to
intervene shall be filed in accordance with the Commission's ``Agency
Rules of Practice and Procedure'' in 10 CFR part 2. Interested
person(s) should consult a current copy of 10 CFR 2.309, which is
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The
NRC's regulations are accessible electronically from the NRC Library on
the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is
filed within 60 days, the Commission or a presiding officer designated
by the Commission or by the Chief Administrative Judge of the Atomic
Safety and Licensing Board Panel, will rule on the request and/or
petition; and the Secretary or the Chief Administrative Judge of the
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends
[[Page 59661]]
to rely to establish those facts or expert opinion to support its
position on the issue. 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 the NRC's regulations, policies, and procedures.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i)-(iii).
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of any amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission by
October 31, 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 10 CFR 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. Details regarding the opportunity to
make a limited appearance will be provided by the presiding officer if
such sessions are scheduled.
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, as amended at 77 FR 46562, August 3, 2012). 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.
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 to the NRC,'' which is available on the agency's
public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.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 Electronic Filing 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.
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/electronic-sub-ref-mat.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
[[Page 59662]]
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 Electronic
Filing 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 Electronic Filing Help Desk is available between 9 a.m.
and 7 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
stating why there is good cause for not filing electronically and
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 hearing request and petition to intervene
will require including information on local residence in order to
demonstrate a proximity assertion of interest in the proceeding. With
respect to copyrighted works, except for limited excerpts that serve
the purpose of the adjudicatory filings and would constitute a Fair Use
application, participants are requested not to include copyrighted
materials in their submission.
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Dominion Nuclear Connecticut, Inc. (DNC), Docket No. 50-336, Millstone
Power Station, Unit No. 2 (MPS2), New London County, Connecticut
Date of amendment request: May 25, 2016. A publicly-available
version is in ADAMS under Accession No. ML16153A026.
Description of amendment request: The amendment would add the AREVA
topical report, EMF-2103(P)(A), ``Realistic Large Break [loss of
coolant accident] LOCA [RLBLOCA] Methodology for Pressurized Water
Reactors,'' Revision 3, to MPS2 Technical Specification (TS) 6.9.1.8.b,
``Core Operating Limits Report,'' which lists the analytical methods
used to determine the core operating limits. The methodology in EMF-
2013(P)(A) for RLBLOCA has been used for the MPS2 LBLOCA analysis of
the AREVA Standard CE-14 HTP fuel product with M5 cladding, which DNC
plans to introduce beginning with the fresh fuel for MPS2 Cycle 25 in
spring 2017.
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 revisions
provided in [brackets]:
1. Does the proposed [amendment] involve a significant increase
in the probability or consequences of an accident previously
evaluated?
Response: No.
The proposed change to TS 6.9.1.8.b permits the use of the AREVA
RLBLOCA methodology to analyze the MPS2 LBLOCA to ensure that the
plant continues to meet the Emergency Core Cooling System (ECCS)
performance acceptance criteria in 10 CFR 50.46. The RLBLOCA
analysis demonstrates MPS2 continues to satisfy the 10 CFR 50.46
ECCS performance acceptance criteria using an NRC-approved
evaluation model. The proposed change to the list of NRC-approved
methodologies listed in TS 6.9.1.8.b has no impact on how the plant
is operated or configured. Addition of this methodology to the list
of methodologies in TS 6.9.1.8.b does not impact either the
probability or consequences of an accident currently evaluated in
Chapter 14 of the [Updated Final Safety Analysis Report] UFSAR.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed [amendment] create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
The proposed change to TS 6.9.1.8.b adds topical report EMF-
2103(P)(A) to the list of approved methodologies for determining
core operating limits at MPS2. The proposed amendment has no adverse
effect on plant operation or accident mitigation equipment. The
amendment does not create any new credible failure mechanisms,
malfunctions, or accident initiators not considered in the current
design basis accidents (DBAs). The response of the plant and
operators following a DBA will not be changed. The proposed
amendment does not create the possibility of a new failure mode
associated with any equipment or human performance failures. Thus,
the possibility of a new or different type of accident is not
created.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from those previously
evaluated within the FSAR.
3. Does the proposed [amendment] involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to TS 6.9.1.8.b adds topical report EMF-
2103(P)(A) to the list of approved methodologies for determining
core operating limits at MPS2. Approved methodologies will be used
to ensure that the plant continues to meet applicable design
criteria and safety analysis acceptance criteria. The proposed
amendment has no [e]ffect on the ability of the plant to mitigate
DBAs and ensure consequences of the existing DBA remains bounding.
The margin of safety to mitigate consequences of DBAs is not
reduced. Structures, systems and components used to mitigate DBAs
are not affected. No changes are being made to safety limits or
safety system settings required by TS. Therefore, the proposed
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three
[[Page 59663]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA
23219.
NRC Branch Chief: Travis L. Tate.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant (PNP), Van Buren County, Michigan
Date of amendment request: July 11, 2016. A publicly-available
version is in ADAMS under Accession No. ML16193A005.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to eliminate TS Section
5.5.7, ``Inservice Testing [IST] Program.'' A new defined term,
``INSERVICE TESTING PROGRAM,'' is added to the TS Definitions section.
This amendment request is consistent with Technical Specifications Task
Force (TSTF)-545, Revision 3, ``TS Inservice Program Removal & Clarify
SR [Surveillance Requirement] Usage Rule Application to Section 5.5
Testing.''
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 revise TS Chapter 5, Administrative
Controls, Section 5.5, Programs and Manuals, by eliminating the TS
5.5.7, Inservice Testing Program, specification. Most requirements
in the IST Program would be removed, as they are duplicative of
requirements in the ASME [American Society of Mechanical Engineers]
OM Code [ASME Code for Operation and Maintenance of Nuclear Power
Plants], as clarified by Code Case OMN-20, Inservice Test Frequency.
The remaining requirements in the Section 5.5 IST Program would be
eliminated because the NRC has determined their inclusion in the TS
is contrary to regulations. A new defined term, INSERVICE TESTING
PROGRAM, would be added to the TS, which references the requirements
of 10 CFR 50.55a(f),
Performance of IST is not an initiator to any accident
previously evaluated. As a result, the probability of occurrence of
an accident is not significantly affected by the proposed change.
IST frequencies under Code Case OMN-20 are equivalent to the current
testing period allowed by the TS with the exception that testing
frequencies greater than 2 years may be extended by up to 6 months
to facilitate test scheduling and consideration of plant operating
conditions that may not be suitable for performance of the required
testing. The testing frequency extension will not affect the ability
of the components to mitigate any accident previously evaluated as
the components are required to be operable during the testing period
extension. Performance of inservice tests utilizing the allowances
in OMN-20 will not significantly affect the reliability of the
tested components. As a result, the availability of the affected
components, as well as their ability to mitigate the consequences of
accidents previously evaluated, is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any [accident] previously evaluated?
Response: No.
The proposed change does not alter the design or configuration
of the plant. The proposed change does not involve a physical
alteration of the plant; no new or different kind of equipment will
be installed. The proposed change does not alter the types of IST
performed. In most cases, the frequency of IST would be unchanged.
However, the frequency of testing would not result in a new or
different kind of accident from any previously evaluated since the
testing methods are not altered.
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 would eliminate some requirements from the
TS in lieu of requirements in the ASME Code, as modified by use of
Code Case OMN-20. Compliance with the ASME Code is required by 10
CFR 50.55a. The proposed change also would allow inservice tests
with frequencies greater than 2 years to be extended by 6 months to
facilitate test scheduling and consideration of plant operating
conditions that may not be suitable for performance of the required
testing. The testing frequency extension will not affect the ability
of the components to respond to an accident as the components are
required to be operable during the testing period extension. The
proposed change would eliminate the existing TS SR 3.0.3 allowance
to defer performance of missed inservice tests up to the duration of
the specified testing frequency, and instead would require an
assessment of the missed test on equipment operability. This
assessment will consider the effect on a margin of safety (equipment
operability). Should the component be inoperable, the Technical
Specifications provide actions to ensure that the margin of safety
is protected. The proposed change also would eliminate a statement
that nothing in the ASME Code should be construed to supersede the
requirements of any TS. The NRC has determined that statement to be
incorrect. However, elimination of the statement will have no effect
on plant operation or safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeanne Cho, Senior Counsel, Entergy
Services, Inc., 440 Hamilton Ave., White Plains, NY 10601.
NRC Branch Chief: David J. Wrona.
LaCrosseSolutions, Inc., and Dairyland Power Cooperative, Docket Nos.:
50-409 and 72-046, La Crosse Boiling Water Reactor (LACBWR), La Crosse
County, Wisconsin
Date of amendment request: June 27, 2016. A publicly-available
version is in ADAMS under Accession No. ML16200A083.
Description of amendment request: The proposed amendment would
amend the Possession Only License for the LACBWR to reflect the
approval of the LACBWR License Termination Plan (LTP) when that review
and approval process is completed by the NRC staff. The LTP will become
a supplement to LACBWR's other decommissioning documents and will be
implemented by the licensee to complete decommissioning activities at
the LACBWR site. Once decommissioning is complete, a separate request
will be made to the NRC by the licensee to terminate the LACBWR
license.
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?
The only remaining accident following completion of fuel
transfer to the [Independent Spent Fuel Storage Installation]
(ISFSI) is a radioactive release accident where spontaneous release
of the (non-ISFSI-related) radioactive source term remaining at the
LACBWR site in a form and quantity is immediately released through
an airborne or liquid release path.
A radioactive release analysis was performed to establish the
bounding event at the site considering the current stage of LACBWR
decommissioning. 1.175 [Curies]
[[Page 59664]]
(Ci) of radioactive material is conservatively estimated in the
analysis to be present on plant surfaces, and as such represents the
assumed total non-ISFSI radioactive source term remaining at the
LACBWR site. The LACBWR analysis of postulated release events
separately considers the portion of this remaining radioactive
contamination that is immediately releasable as airborne
contamination and that is immediately releasable as contaminated
liquid.
A conservative fraction of 30 percent of the total remaining
source term is assumed in the analysis to be immediately available
for airborne release. The analysis results demonstrate that the
consequences of releasing 30 percent of the non-ISFSI radioactive
source term remaining at the LACBWR site to the atmosphere are well
within the applicable 10 CFR 100.11 and [U.S. Environmental
Protection Agency] (EPA) [Protective Action Guides] (PAG) limits.
The portion of the total remaining source term conservatively
assumed in the analysis to be available for liquid release at any
one time is 80 percent of the radioactively contaminated liquid
stored in the site retention tank. In the unlikely event that 80
percent of the retention tank volume at a total radionuclide
concentration of 3.9E-03 [mu]Ci/cc were to be released from the
retention tank at a flow rate of 20 [gallons per minute] (gpm), the
normal effluent concentration limits of 10 CFR 20, Appendix B, Table
2, would not be exceeded. Thus, the liquid release analysis
demonstrates that there is no reasonable likelihood that a
postulated radioactive liquid release event could result in
exceeding the normal effluent concentration limits of 10 CFR 20,
Appendix B.
With consideration for the current stage of LACBWR
decommissioning and with spent nuclear fuel now stored in the ISFSI,
the bounding radioactive release analysis, for both airborne and
liquid releases, confirms that the minimal radioactive material
resulting from LACBWR operation and remaining on the LACBWR site is
insufficient for any potential event to result in exceeding dose
limits or otherwise involving a significant adverse effect on public
health and safety.
The proposed change does not affect the boundaries used to
evaluate compliance with liquid or gaseous effluent limits, and has
no impact on plant operations. The proposed changes do not have an
adverse impact on the remaining decommissioning activities or any
decommissioning related postulated accident consequences.
The proposed changes related to the approval of the LTP do not
affect operating procedures or administrative controls that have the
function of preventing or mitigating the remaining decommissioning
design basis accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The accident analysis for the facility related to
decommissioning activities is described in the [Decommissioning
Plan/Post-Shutdown Decommissioning Activities Report] (D-Plan/
PSDAR). The requested license amendment is consistent with the plant
activities described in the D-Plan/PSDAR. Thus, the proposed changes
do not affect the remaining plant systems, structures, or components
in a way not previously evaluated.
There are sections of the LTP that refer to the decommissioning
activities still remaining. These activities are performed in
accordance with approved site processes and undergo a 10 CFR 50.59
review as required prior to initiation. The proposed amendment
merely makes mention of these processes and does not bring about
physical changes to the facility.
Therefore, the facility conditions for which the remaining
postulated accident has been evaluated is still valid and no new
accident scenarios, failure mechanisms, or single failures are
introduced by this amendment. The system operating procedures are
not affected.
Therefore, the proposed changes will 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?
The LTP is a plan for demonstrating compliance with the
radiological criteria for license termination as provided in 10 CFR
20.1402. The margin of safety defined in the statements of
consideration for the final rule on the Radiological Criteria for
License Termination is described as the margin between the 100
[millirem per year] (mrem/yr) public dose limit established in 10
CFR 20.1301 for licensed operation and the 25 mrem/yr dose limit to
the average member of the critical group at a site considered
acceptable for unrestricted use (one of the criteria of 10 CFR
20.1402). This margin of safety accounts for the potential effect of
multiple sources of radiation exposure to the critical group. Since
the License Termination Plan is designed to comply with the
radiological criteria for license termination for unrestricted use,
the LTP supports this margin of safety.
In addition, the LTP provides the methodologies and criteria
that will be used to perform remediation activities of residual
radioactivity to demonstrate compliance with the [As Low As
Reasonably Achievable] (ALARA) criterion of 10 CFR 20.1402.
Additionally, the LTP is designed with recognition that (a) the
methods in MARSSIM (Multi-Agency Radiation Survey and Site
Investigation Manual) and (b) the building surface contamination
levels are not directly applicable to use with complex nonstructural
components. Therefore, the LTP states that nonstructural components
remaining in buildings (e.g., pumps, heat exchangers, etc.) will be
evaluated against the criteria of Regulatory Guide 1.86,
``Termination of Operating Licenses for Nuclear Reactors,'' to
determine if the components can be released for unrestricted use.
The LTP also states that materials, surveyed and evaluated as a-part
of normal decommissioning activities and prior to implementation of
the final radiation surveys, will be surveyed for release using
current site procedures to demonstrate compliance with the ``no
detectable'' criteria. Such materials that do not pass these
criteria will be controlled as contaminated.
Also, as previously discussed, the bounding radioactive release
accident analysis for decommissioning is based on a conservative
estimate of the radioactive material remaining onsite. Since the
bounding accident results in a release of more airborne and liquid
radioactivity than can be released from planned LTP decommissioning
events, the margin of safety associated with the consequences of
decommissioning accidents is not reduced by this activity.
Thus, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Russ Workman, General Counsel, Energy
Solutions, 299 South Main Street, Suite 1700, Salt Lake City, Utah
84111.
NRC Branch Chief: Bruce Watson.
South Carolina Electric and Gas Company and South Carolina Public
Service Authority, 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: June 16, 2016, as revised August 8,
2016. Publicly-available versions are in ADAMS under Accession Nos.
ML16168A257 and ML16221A649, respectively.
Description of amendment request: The requested amendment proposes
to depart from approved AP1000 Design Control Document Tier 2* and
associated Tier 2 information in the Updated Final Safety Analysis
Report (UFSAR). Specifically, the requested amendment proposes to
depart from UFSAR text and figures that describe the connections
between floor modules and structural wall modules in the containment
internal structures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
[[Page 59665]]
Response: No.
The design functions of the nuclear island structures are to
provide support, protection, and separation for the seismic Category
I mechanical and electrical equipment located in the nuclear island.
The nuclear island structures are structurally designed to meet
seismic Category I requirements as defined in Regulatory Guide 1.29.
The change of the design details for the floor modules and the
connections between floor modules and the structural wall modules,
and the change to more clearly state the design requirement that
these connections meet criteria and requirements of American
Concrete Institute (ACI) 349 and American Institute of Steel
Construction (AISC) N690, do not have an adverse impact on the
response of the nuclear island structures to safe shutdown
earthquake ground motions or loads due to anticipated transients or
postulated accident conditions. The change of the design details for
the connections between floor modules and the structural wall
modules, and the clarification of design requirements for these
connections, do not impact the support, design, or operation of
mechanical and fluid systems. There is no change to plant systems or
the response of systems to postulated accident conditions. There is
no change to the predicted radioactive releases due to normal
operation or postulated accident conditions. The plant response to
previously evaluated accidents or external events is not adversely
affected, nor does the change described create any new accident
precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change is to revise design details for the floor
modules and the connections between floor modules and the structural
wall modules, and more clearly state the design requirement that
these connections meet criteria and requirements of ACI 349 and AISC
N690. The clarification and changes to the design details for the
floor modules and the connections between floor modules and the
structural wall modules do not change the design requirements of the
nuclear island structures. The clarification and changes of the
design details for the floor modules and the connections between
floor modules and the structural wall modules do not result in a new
failure mechanism for the nuclear island structures or new accident
precursors. As a result, the design function of the nuclear island
structures is not adversely affected by the proposed change.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed changes, thus, no margin of
safety is reduced. The acceptance limits for the design of seismic
Category I structures are included in the codes and standards used
for the design, analysis, and construction of the structures. The
two primary codes for the seismic Category I structures are American
Institute of Steel Construction (AISC) N690 and American Concrete
Institute (ACI) 349. The changes to the design of the connection of
the floor module to the structural wall modules in the containment
internal structures satisfy applicable provisions of AISC N690 and
ACI 349 and supplemental requirements included in the UFSAR.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety previously evaluated.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius
LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
Acting NRC Branch Chief: Jennifer Dixon-Herrity.
South Carolina Electric and Gas Company and South Carolina Public
Service Authority, 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: July 19, 2016. A publicly-available
version is in ADAMS under Accession No. ML16202A035.
Description of amendment request: The amendment request proposes
changes to the Technical Specifications and Updated Final Safety
Analysis Report (UFSAR) Tier 2 information to update the Protection and
Safety Monitoring System (PMS) to align with the requirements in
Institute of Electrical and Electronics Engineers (IEEE) 603-1991,
``IEEE Standard Criteria for Safety Systems for Nuclear Power
Generating Stations.'' IEEE 603-1991, Clause 6.6, ``Operating
Bypasses,'' imposes requirements on the operating bypasses (i.e.,
``blocks'' and ``resets'') used for the AP1000 PMS. The PMS functional
logic for blocking the source range neutron flux doubling signal shown
in UFSAR Figure 7.2-1 (Sheet 3) requires revision to fully comply with
this requirement.
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 modifies the PMS logic used to terminate an
inadvertent boron dilution accident which results in a source range
flux doubling signal. An inadvertent boron dilution is caused by the
failure of the demineralized water transfer and storage system or
chemical and volume control system, either by controller, operator
or mechanical failure. The proposed changes to PMS and Technical
Specification requirements do not adversely affect any of these
accident initiators or introduce any component failures that could
lead to a boron dilution event; thus the probabilities of accidents
previously evaluated are not affected. The proposed changes do not
adversely interface with or adversely affect any system containing
radioactivity or affect any radiological material release source
term; thus the radiological releases in an accident are not
affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The accident analysis evaluates events involving a decrease in
reactor coolant system boron concentration due to a malfunction of
the chemical and volume control system in Modes 1 through 6. The
Technical Specifications currently provide administrative controls
to prevent a boron dilution event in Mode 6. The proposed change
would provide additional PMS interlocks and administrative controls
for prevention of a boron dilution event applicable in Modes 2, 3,
4, and 5. The proposed changes to the PMS design do not adversely
affect the design or operation of safety related equipment or
equipment whose failure could initiate an accident from 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, the proposed amendment does not create the
possibility of a new or different kind of accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change would add additional restrictions on the
source range flux doubling signal operational bypass to align it
with the requirements in IEEE 603 and provide assurance that the
protection logic is enabled whenever the plant is in a condition
where protection might be required. These changes to the PMS design
do not adversely impact nor affect the design, construction, or
operation of any plant [structure, system, and components (SSCs)],
including any equipment whose failure could initiate an accident or
a failure of a fission product barrier. No analysis is adversely
[[Page 59666]]
affected by the proposed changes. Furthermore, no system function,
design function, or equipment qualification will be adversely
affected by the changes.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety previously evaluated.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius
LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
Acting NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
Date of amendment request: July 25, 2016. A publicly-available
version is in ADAMS under Accession No. ML16207A340.
Description of amendment request: The amendment request proposes
changes to a plant-specific Tier 1 (and combined license Appendix C)
table and the Updated Final Safety Analysis Report (UFSAR) tables to
clarify the flow area for the Automatic Depressurization System (ADS)
fourth stage squib valves and to reduce the minimum effective flow area
for the second and third stage ADS control valves. Pursuant to the
provisions of 10 CFR 52.63(b)(1), an exemption from elements of the
design as certified in the 10 CFR part 52, Appendix D, design
certification rule is also requested for the plant-specific Design
Control Document Tier 1 material departures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not adversely affect the operation of
any systems or equipment that initiate an analyzed accident or alter
any structures, systems, and components (SSC) accident initiator or
initiating sequence of events. The proposed changes do not adversely
affect the physical design and operation of the second and third
stage ADS control valves and fourth stage ADS squib valves,
including as-installed inspections, testing, and maintenance
requirements, as described in the UFSAR. Therefore, the operation of
the second and third stage ADS control valves and fourth stage ADS
squib valves is not adversely affected.
The proposed changes do not adversely affect the ability of the
second and third stage ADS control valves and fourth stage ADS squib
valves to perform their design functions. The designs of the second
and third stage ADS control valves and fourth stage ADS squib valves
continue to meet the same regulatory acceptance criteria, codes, and
standards as required by the UFSAR. In addition, the proposed
changes maintain the capabilities of the second and third stage ADS
control valves and fourth stage ADS squib valves to mitigate the
consequences of an accident and to meet the applicable regulatory
acceptance criteria. The proposed changes do not adversely affect
the prevention and mitigation of other abnormal events, e.g.,
anticipated operational occurrences, earthquakes, floods and turbine
missiles, or their safety or design analyses. Therefore, the
consequences of the accidents evaluated in the UFSAR are not
affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not affect the operation of any systems
or equipment that may initiate a new or different kind of accident,
or alter any SSC such that a new accident initiator or initiating
sequence of events is created. The proposed changes do not adversely
affect the physical design and operation of the second and third
stage ADS control valves and fourth stage ADS squib valves,
including as-installed inspections, testing, and maintenance
requirements, as described in the UFSAR. Therefore, the operation of
the second and third stage ADS control valves and fourth stage ADS
squib valves is not adversely affected. These proposed changes do
not adversely affect any other SSC design functions or methods of
operation in a manner that results in a new failure mode,
malfunction or sequence of events that affect safety-related or
nonsafety-related equipment. Therefore, this activity does not allow
for a new fission product release path, result in a new fission
product barrier failure mode, or create a new sequence of events
that results in significant fuel cladding failures.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes maintain existing safety margins. The
proposed changes maintain the capabilities of the second and third
stage ADS control valves and fourth stage ADS squib valves to
perform their design functions. The proposed changes maintain
existing safety margin through continued application of the existing
requirements of the UFSAR, while updating the acceptance criteria
for verifying the design features necessary to confirm the second
and third stage ADS control valves and fourth stage ADS squib valves
perform the design functions required to meet the existing safety
margins in the safety analyses. Therefore, the proposed changes
satisfy the same design functions in accordance with the same codes
and standards as stated in the UFSAR. These changes do not adversely
affect any design code, function, design analysis, safety analysis
input or result, or design/safety margin.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed changes, and no margin of
safety is reduced. Therefore, the requested amendment does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Acting Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket No. 50-425, Vogtle
Electric Generating Plant, Unit 2, Burke County, Georgia
Date of amendment request: August 12, 2016. A publicly-available
version is in ADAMS under Accession No. ML16225A619.
Description of amendment request: The licensee proposes to modify
the Vogtle Electric Generating Plant, Unit 2, Technical Specifications
(TSs) Limiting Condition for Operation 3.7.9, ``Ultimate Heat Sink
(UHS),'' such that with the 2B Nuclear Service Cooling Water (NSCW)
transfer pump inoperable for refurbishment, the Completion Time of
Condition 3.7.9.D.2.2 would be 46 days as opposed to 31 days. This TS
change would be a one-time change and in effect only for the 2B NSCW
transfer pump for the remainder of Cycle 19.
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.
[[Page 59667]]
The proposed change does not alter any plant equipment or
operating practices in such a manner that the probability of an
accident is increased. The proposed changes will not alter
assumptions relative to the mitigation of an accident or transient
event. Furthermore, the UHS will remain capable of adequately
responding to a design basis event during the period of the extended
CT [Completion Time]. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different accident from any accident previously evaluated?
Response: No.
The proposed change does not introduce any new or unanalyzed
modes of operation. The refurbishment of the pump does not involve
any unanalyzed modifications to the design or operational limits of
the NSCW system. The redundant pump and compensatory measures
allowed by the Technical Specifications will remain unaffected.
Therefore, no new failure modes or accident precursors are created
due to the pump refurbishment during the extended Completion Time.
For the reasons noted above, the proposed change will not create the
possibility of a new or different accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is related to the ability of the fission
product barriers to perform their design functions during and
following an accident. These barriers include the fuel cladding, the
reactor coolant system, and the containment. The performance of
these fission product barriers will not be affected by the proposed
change; therefore, the margin to the onsite and offsite radiological
dose limits are not significantly reduced.
During the extended Completion Time for the 2B NSCW transfer
pump, the NSCW system and the UHS will remain capable of mitigating
the consequences of a design basis event such as a LOCA [loss-of-
coolant accident]. Technical Specifications Action 3.7.9.D.2.1 will
be taken to provide an alternate method of basin transfer.
For the reasons noted above, there is no 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: Jennifer M. Buettner, Associate General Counsel,
Southern Nuclear Operating Company, Inc., 40 Inverness Center Parkway,
Birmingham, AL 35242.
NRC Branch Chief: Michael T. Markley.
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.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3 (MPS3), New London County, Connecticut
Date of amendment request: August 31, 2015.
Brief description of amendment: The amendment revised the MPS3
Design Features--Fuel Storage Technical Specification 5.6.3,
``Capacity,'' to specify the spent fuel pool storage capacity limit in
terms of the total number of fuel assemblies.
Date of issuance: August 4, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 270. A publicly-available version is in ADAMS under
Accession No. ML16206A001; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-49: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 24, 2015 (80
FR 73235).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 4, 2016.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne
County, Mississippi
Date of application for amendment: September 15, 2015.
Brief description of amendment: The amendment revised the GGNS
Technical Specifications (TSs) to eliminate the ``Inservice Testing
Program,'' specification in Section 5.5, ``Programs and Manuals,''
which is superseded by Code Case OMN-20. A new defined term,
``INSERVICE TESTING PROGRAM,'' would be added to TS Section 1.1,
``Definitions.'' This request is consistent with TS Task Force (TSTF)-
545, Revision 1, ``TS Inservice Testing Program Removal & Clarify SR
[Surveillance Requirement] Usage Rule Application to Section 5.5
Testing.''
Date of issuance: August 4, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No: 211. A publicly-available version is in ADAMS under
Accession No. ML16140A133; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-29: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 1, 2016 (81 FR
10679).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 4, 2016.
No significant hazards consideration comments received: No.
[[Page 59668]]
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station (Braidwood), Units 1 and 2, Will County, Illinois and
Docket Nos. STN 50-454 and STN 50-455, Byron Station (Byron), Unit Nos.
1 and 2, Ogle County, Illinois
Date of application for amendments: February 23, 2016.
Brief description of amendment: The amendments revise technical
specifications (TSs) 4.2.1, ``Fuel Assemblies,'' and 5.6.5, ``Core
Operating Limits Report (COLR),'' to allow the use of Optimized
ZIRLO\TM\ fuel cladding material in Braidwood, Units 1 and 2, and
Byron, Unit Nos. 1 and 2 and to add WCAP-12610-P-A, ``VANTAGE+ Fuel
Assembly Reference Core Report,'' and Addendum 1-A to Topical Report
WCAP-12610-P-A and CENPD-404-P-A, ``Optimized ZIRLO'' to the list of
documents previously reviewed and approved by the NRC.
Date of issuance: August 1, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos: 190/196. A publicly-available version is in ADAMS
under Accession No. ML16180A251; documents related to these amendments
are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-72, NPF-77, NPF-37, and
NPF-66: The amendments revised the Technical Specifications and
Licenses.
Date of initial notice in Federal Register: May 10, 2016 (81 FR
28897).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 1, 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: March 24, 2016, as supplemented by
letter dated May 11, 2016.
Brief description of amendments: The amendments revised the
frequency for cycling of the recirculation pump discharge valves as
specified in Technical Specification (TS) Surveillance Requirement (SR)
3.5.1.5. Specifically, the amendments changed the frequency for the SR
such that it is performed in accordance with the Inservice Testing
Program.
Date of issuance: August 10, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendments Nos.: 309 (Unit 2) and 313 (Unit 3). A publicly-
available version is in ADAMS under Accession No. ML16165A002;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: June 7, 2016 (81 FR
36619).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 10, 2016.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant (CNP), Units 1 and 2, Berrien County, Michigan
Date of amendment request: January 29, 2016.
Brief description of amendments: The amendments revised the CNP,
Units 1 and 2, technical specification (TS) requirements to address
Generic Letter 2008-01, ``Managing Gas Accumulation in Emergency Core
Cooling, Decay Heat Removal, and Containment Spray Systems,'' as
described in the Technical Specifications Task Force (TSTF) Traveler,
TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas
Accumulation.''
Date of issuance: August 4, 2016.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment Nos.: 331--Unit 1 and 312--Unit 2. A publicly-available
version is in ADAMS under Accession No. ML16195A004; documents related
to this amendment are listed in the Safety Evaluation enclosed with the
amendment.
Renewed Facility Operating License Nos. DPR-58 and DPR-74: The
amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: March 15, 2016 (81 FR
13843).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 4, 2016.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: November 24, 2014, as supplemented by
letter dated September 28, 2015.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) by adopting 21 previously NRC-approved
Technical Specifications Task Force (TSTF) Travelers and one request
not associated with TSTF Travelers. SNC stated that these TSTF
Travelers are generic changes chosen to increase the consistency
between the Joseph M. Farley Nuclear Plant, Units 1 and 2; the Improved
Standard Technical Specifications for Westinghouse plants (NUREG-1431);
and the TSs of the other plants in the SNC fleet.
Date of issuance: August 3, 2016.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 203 (Unit 1) and 199 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML15233A448; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: The
amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: February 3, 2015 (80 FR
5804). The supplemental letter dated September 28, 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 August 3, 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: March 16, 2016.
Brief description of amendments: The amendments revised the
Technical Specifications to allow the use of Optimized
ZIRLOTM as an approved fuel rod cladding.
Date of issuance: August 4, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 182 (Unit 1) and 163 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16179A386; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
[[Page 59669]]
Renewed Facility Operating License Nos. NPF-68 and NPF-81:
Amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: May 24, 2016 (81 FR
32809).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 4, 2016.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendment request: March 16, 2016.
Brief description of amendments: The amendments revised the
Technical Specifications to allow the use of Optimized
ZIRLOTM as an approved fuel rod cladding.
Date of issuance: August 4, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 204 (Unit 1) and 200 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16179A386; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments
revised the Renewed Facility Operating Licenses and Technical
Specifications.
Date of initial notice in Federal Register: May 24, 2016 (81 FR
32808).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 4, 2016.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: January 27, 2016, as supplemented by
letter dated May 19, 2016.
Brief description of amendment: The amendment revised the Technical
Specifications to allow the use of Optimized ZIRLO\TM\ as an approved
fuel rod cladding.
Date of issuance: The amendment is effective upon issuance and
shall be implemented within 90 days of the date of issuance.
Effective date: August 3, 2016.
Amendment No.: 216. A publicly-available version is in ADAMS under
Accession No. ML16179A293; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 12, 2016 (81 FR
21603). The supplemental letter dated May 19, 2016, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 3, 2016.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 18th day of August 2016.
For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2016-20391 Filed 8-29-16; 8:45 am]
BILLING CODE 7590-01-P