[Federal Register Volume 83, Number 39 (Tuesday, February 27, 2018)]
[Notices]
[Pages 8509-8523]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-03727]
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NUCLEAR REGULATORY COMMISSION
[NRC-2018-0031]
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 January 30, 2018, to February 12, 2018. The
last biweekly notice was published on February 13, 2018.
DATES: Comments must be filed by March 29, 2018. A request for a
hearing must be filed by April 30, 2018.
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 website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0031. Address
questions about NRC dockets to Jennifer Borges; telephone: 301-287-
9127; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: May Ma, Office of Administration, Mail
Stop: TWFN-7-A60M, 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: Paula Blechman, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-2242, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2018-0031, facility name, unit
number(s), 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 website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0031.
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 DAMS) 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-2018-0031, 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 will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, 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
[[Page 8510]]
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
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and petition for leave to intervene
(petition) with respect to the action. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309. The NRC's regulations are accessible
electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
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 petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in 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 petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) 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.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the 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). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
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 establish 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 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 the 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 no later
than 60 days from the date of publication of this notice. 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 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. Alternatively, a State, local
governmental body, Federally-recognized Indian Tribe, or agency thereof
may participate as a non-party under 10 CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
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
his or her 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 and petition for leave to intervene (petition), 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 that request to
participate 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. Detailed
guidance on making electronic submissions may be found in the Guidance
for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit
paper copies of their filings unless they seek an exemption in
[[Page 8511]]
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 (1) request a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign submissions and access the E-Filing
system for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition or
other adjudicatory document (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 website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a
digital ID certificate and a docket has been created, the participant
can then submit adjudicatory documents. Submissions must be in Portable
Document Format (PDF). Additional guidance on PDF submissions is
available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the
time the document is 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
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public website 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 6 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, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff. Participants filing adjudicatory documents 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
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the
Commission or the presiding officer. If you do not have an NRC-issued
digital ID certificate as described above, click cancel when the link
requests certificates and you will be automatically directed to the
NRC's electronic hearing dockets where you will be able to access any
publicly available documents in a particular hearing docket.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or personal phone
numbers in their filings, unless an NRC regulation or other law
requires submission of such information. For example, in some
instances, individuals provide home addresses in order to demonstrate
proximity to a facility or site. 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.
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: October 9, 2017. A publicly-available
version is in ADAMS under Accession No. ML17283A248.
Description of amendment request: The amendment would revise
Limiting Condition for Operation (LCO) 3.10.1, to expand its scope to
include provisions for temperature excursions greater than 200 degrees
Fahrenheit ([deg]F) as a consequence of inservice leak and hydrostatic
testing, and as a consequence of scram time testing initiated in
conjunction with an inservice leak or hydrostatic test, while
considering operational conditions to be in Mode 4. This change is
consistent with NRC approved Technical Specification Task Force (TSTF)
Improved Standard Technical Specification Change Traveler, TSTF-484,
``Use of TS 3.10.1 for Scram Time Testing Activities,'' Revision 0.
The NRC staff issued a Notice of Availability for TSTF-484 in the
Federal Register on October 27, 2006 (71 FR 63050). The staff also
issued a Federal Register notice on August 21, 2006 (71 FR 48561), that
provided a model safety evaluation and a model no significant hazards
consideration (NSHC) determination that licensees could reference in
their plant-specific application. In its application dated October 9,
2017, the licensee affirmed the applicability of the model NSHC
determination for Fermi 2.
Basis for proposed no NSHC determination: As required by 10 CFR
50.91(a), the licensee affirmed the applicability of the model NSHC,
which is presented below:
Criterion 1: The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 8512]]
Technical Specifications currently allow for operation at
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. Extending the activities that can apply this allowance will not
adversely impact the probability or consequences of an accident
previously evaluated. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Criterion 2: The proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Technical Specifications currently allow for operation at
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. No new operational conditions beyond those currently allowed by
LCO 3.10.1 are introduced. The changes do not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements or eliminate any existing requirements. The
changes do not alter assumptions made in the safety analysis. The
proposed changes are consistent with the safety analysis assumptions
and current plant operating practice. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3: The proposed change does not involve a significant
reduction in a margin of safety.
Technical Specifications currently allow for operation at
greater than 200[emsp14][deg]F while imposing MODE 4 requirements in
addition to the secondary containment requirements required to be
met. Extending the activities that can apply this allowance will not
adversely impact any margin of safety. Allowing completion of
inspections and testing and supporting completion of scram time
testing initiated in conjunction with an inservice leak or
hydrostatic test prior to power operation results in enhanced safe
operations by eliminating unnecessary maneuvers to control reactor
temperature and pressure. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the above 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: Jon P. Christinidis, DTE Energy, Expert
Attorney--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-
1279.
NRC Branch Chief: David J. Wrona.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2 (CNS), York County, South Carolina
Date of amendment request: May 2, 2017, as supplemented by letters
dated July 20 and November 21, 2017. Publicly-available versions are in
ADAMS under Accession Nos. ML17122A116, ML17201Q132, and ML17325A588,
respectively.
Description of amendment request: The amendments would modify CNS
Technical Specifications (TSs) to extend the Completion Time (CT) of TS
3.8.1, ``AC [Alternating Current] Sources--Operating,'' Required Action
B.6 (existing Required Action B.4, numbered as B.6) for an inoperable
emergency diesel generator (DG) from 72 hours to 14 days. A conforming
change is also proposed to extend the maximum CT of TS 3.8.1 Required
Actions A.3 and B.4. To support this request, the licensee will add a
supplemental power source (i.e., two supplemental diesel generators
(SDGs) per station) with the capability to power any emergency bus. The
SDGs will have the capacity to bring the affected unit to cold
shutdown. Additionally, the amendments would modify TS 3.8.1 to add new
two limiting conditions for operation (LCOs), TS LCO 3.8.1.c and TS LCO
3.8.1.d, to ensure that at least one train of shared components has an
operable emergency power supply.
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 involves extending the TS CT for an
inoperable DG at CNS [. . .]. The proposed change also involves
adding a new Required Action to TSs to ensure that at least one
train of shared components at CNS [. . .] has an operable emergency
power supply whenever one DG is inoperable. The DGs at both stations
are safety related components which provide a backup electrical
power supply to the onsite emergency power distribution system. The
proposed change does not affect the design of the DGs, the
operational characteristics or function of the DGs, the interfaces
between the DGs and other plant systems or the reliability of the
DGs. The DGs are not accident initiators; the DGs are designed to
mitigate the consequences of previously evaluated accidents
including a loss of offsite power. Extending the CT for a single DG
would not affect the previously evaluated accidents since the
remaining DGs supporting the redundant engineered safety feature
systems would continue to be available to perform the accident
mitigation functions. Thus, allowing a DG to be inoperable for an
additional 11 days for performance of maintenance or testing does
not increase the probability of a previously evaluated accident.
Deterministic and probabilistic risk assessment techniques
evaluated the effect of the proposed TS change to extend the CT for
an inoperable DG on the availability of an electrical power supply
to the plant emergency safeguards feature systems. These assessments
concluded that the proposed CNS [. . .] TS change does not involve a
significant increase in the risk of power supply unavailability.
There is a small incremental risk associated with continued
operation for an additional 11 days with one DG inoperable; however,
the calculated impact provides risk metrics consistent with the
acceptance guidelines contained in Regulatory Guides 1.177 and
1.174. The remaining operable DGs and paths are adequate to supply
electrical power to the onsite emergency power distribution system.
A DG is required to operate only if both offsite power sources fail
and there is an event which requires operation of the plant
engineered safety features such as a design basis accident. The
probability of a design basis accident occurring during this period
is low.
The consequences of previously evaluated accidents will remain
the same during the proposed 14 day CT as during the current CNS [.
. .] 72 hour CT. The ability of the remaining TS required DGs to
mitigate the consequences of an accident will not be affected since
no additional failures are postulated while equipment is inoperable
within the TS CT.
Regarding the proposed change to add Required Action to ensure
that at least one train of shared components has an operable
emergency power supply, there is no change to how or under what
conditions offsite circuits or DGs are operated nor are there any
changes to acceptable operating parameters. Power source operability
requirements for shared components are being moved from the TS Bases
to TS with the proposed change. The proposed change will ensure that
at least one train of shared components has an operable emergency
power supply whenever a DG is inoperable.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change involves extending the TS CT for an
inoperable DG at CNS [. . .]. The proposed change also involves
adding a new Required Action to TSs to ensure that at least one
train of shared components at CNS [. . .] has an operable emergency
power supply whenever one DG is inoperable.
The proposed change does not involve a change in the CNS [. . .]
plant design, plant
[[Page 8513]]
configuration, system operation or procedures involved with the DGs.
The proposed change allows a DG to be inoperable for additional
time. Equipment will be operated in the same configuration and
manner that is currently allowed and designed for. The functional
demands on credited equipment is unchanged. There are no new failure
modes or mechanisms created due to plant operation for an extended
period to perform DG maintenance or testing. Extended operation with
an inoperable DG does not involve any modification to the
operational limits or physical design of plant systems. There are no
new accident precursors generated due to the extended CT.
Regarding the proposed change to add Required Action to ensure
that at least one train of shared components has an operable
emergency power supply, there is no change to how or under what
conditions offsite circuits or DGs are operated nor are there any
changes to acceptable operating parameters. Power source operability
requirements for shared components are being moved from the TS Bases
to TS with the proposed change. The proposed change will ensure that
at least one train of shared components has an operable emergency
power supply whenever a DG is inoperable. This change does not alter
the nature of events postulated in the Updated Final Safety Analysis
Report nor does it introduce any unique precursor mechanisms.
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 the margin of safety?
Response: No.
The proposed change involves extending the TS CT for an
inoperable DG at CNS [. . .]. The proposed change also involves
adding a new Required Action to TSs to ensure that at least one
train of shared components at CNS [. . .] has an operable emergency
power supply whenever one DG is inoperable.
Currently, if an inoperable DG is not restored to operable
status within 72 hours at CNS [. . .], TS 3.8.1, requires the units
to be in Mode 3 (i.e., Hot Standby) within a CT of 6 hours, and to
be in Mode 5 (i.e., Cold Shutdown) within a CT of 36 hours. The
proposed TS changes will allow steady state plant operation at 100
percent power for an additional 11 days for performance of DG
planned reliability improvements and preventive and corrective
maintenance.
Deterministic and probabilistic risk assessment techniques
evaluated the effect of the proposed TS change to extend the CT for
an inoperable DG on the availability of an electrical power supply
to the plant emergency safeguards feature systems. These assessments
concluded that the proposed CNS [. . .] TS change does not involve a
significant increase in the risk of power supply unavailability.
The DGs continue to meet their design requirements; there is no
reduction in capability or change in design configuration. The DG
response to loss of offsite power, loss of coolant accident, station
blackout or fire scenarios is not changed by this proposed
amendment; there is no change to the DG operating parameters. In the
extended CT, as in the existing CT, the remaining operable DGs and
paths are adequate to supply electrical power to the onsite
emergency power distribution system. The proposed change to extend
the CT for an inoperable DG does not alter a design basis safety
limit; therefore, it does not significantly reduce the margin of
safety. The DGs will continue to operate per the existing design and
regulatory requirements.
The proposed TS changes (i.e., the inoperable DG CT extension
request and proposed change to add Required Action to ensure that at
least one train of shared components has an operable emergency power
supply) do not alter the plant design nor do they change the
assumptions contained in the safety analyses. The standby AC power
system is designed with sufficient redundancy such that a DG may be
removed from service for maintenance or testing. The remaining DGs
are capable of carrying sufficient electrical loads to satisfy the
Updated Final Safety Analysis Report requirements for accident
mitigation or unit safe shutdown. The proposed change does not
impact the redundancy or availability requirements of offsite power
circuits or change the ability of the plant to cope with a station
blackout. Therefore, based on the considerations given above, the
proposed changes do not involve a significant reduction in the
margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2 (MNS), Mecklenburg County, North
Carolina
Date of amendment request: May 2, 2017, as supplemented by letters
dated July 20 and November 21, 2017. Publicly-available versions are in
ADAMS under Accession Nos. ML17122A116, ML17201Q132, and ML17325A588,
respectively.
Description of amendment request: The amendments would modify MNS
Technical Specifications (TSs) to extend the Completion Time (CT) of TS
3.8.1, ``AC [Alternating Current] Sources--Operating,'' Required Action
B.6 (existing Required Action B.4, numbered as B.6) for an inoperable
emergency diesel generator (DG) from 72 hours to 14 days. A conforming
change is also proposed to extend the maximum CT of TS 3.8.1 Required
Actions A.3 and B.4. To support this request, the licensee will add a
supplemental power source (i.e., two supplemental diesel generators
(SDGs) per station) with the capability to power any emergency bus. The
SDGs will have the capacity to bring the affected unit to cold
shutdown. Additionally, the amendments would modify TS 3.8.1 to add new
two limiting conditions for operation (LCOs), TS LCO 3.8.1.c and TS LCO
3.8.1.d, to ensure that at least one train of shared components has an
operable emergency power supply.
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 involves extending the TS CT for an
inoperable DG at [. . .] MNS. The proposed change also involves
adding a new Required Action to TSs to ensure that at least one
train of shared components at [. . .] MNS has an operable emergency
power supply whenever one DG is inoperable. The DGs at both stations
are safety related components which provide a backup electrical
power supply to the onsite emergency power distribution system. The
proposed change does not affect the design of the DGs, the
operational characteristics or function of the DGs, the interfaces
between the DGs and other plant systems or the reliability of the
DGs. The DGs are not accident initiators; the DGs are designed to
mitigate the consequences of previously evaluated accidents
including a loss of offsite power. Extending the CT for a single DG
would not affect the previously evaluated accidents since the
remaining DGs supporting the redundant engineered safety feature
systems would continue to be available to perform the accident
mitigation functions. Thus, allowing a DG to be inoperable for an
additional 11 days for performance of maintenance or testing does
not increase the probability of a previously evaluated accident.
Deterministic and probabilistic risk assessment techniques
evaluated the effect of the proposed TS change to extend the
[completion time] CT for an inoperable DG on the availability of an
electrical power supply to the plant emergency safeguards feature
systems. These assessments concluded that the proposed [. . .] MNS
TS change does not involve a significant increase in the risk of
power supply unavailability.
[[Page 8514]]
There is a small incremental risk associated with continued
operation for an additional 11 days with one DG inoperable; however,
the calculated impact provides risk metrics consistent with the
acceptance guidelines contained in Regulatory Guides 1.177 and
1.174.
The remaining operable DGs and paths are adequate to supply
electrical power to the onsite emergency power distribution system.
A DG is required to operate only if both offsite power sources fail
and there is an event which requires operation of the plant
engineered safety features such as a design basis accident. The
probability of a design basis accident occurring during this period
is low.
The consequences of previously evaluated accidents will remain
the same during the proposed 14 day CT as during the current [. . .]
MNS 72 hour CT. The ability of the remaining TS required DGs to
mitigate the consequences of an accident will not be affected since
no additional failures are postulated while equipment is inoperable
within the TS CT.
Regarding the proposed change to add Required Action to ensure
that at least one train of shared components has an operable
emergency power supply, there is no change to how or under what
conditions offsite circuits or DGs are operated nor are there any
changes to acceptable operating parameters. Power source operability
requirements for shared components are being moved from the TS Bases
to TS with the proposed change. The proposed change will ensure that
at least one train of shared components has an operable emergency
power supply whenever a DG is inoperable.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change involves extending the TS CT for an
inoperable DG at [. . .] MNS. The proposed change also involves
adding a new Required Action to TSs to ensure that at least one
train of shared components at [. . .] MNS has an operable emergency
power supply whenever one DG is inoperable.
The proposed change does not involve a change in the [. . .] MNS
plant design, plant configuration, system operation or procedures
involved with the DGs. The proposed change allows a DG to be
inoperable for additional time. Equipment will be operated in the
same configuration and manner that is currently allowed and designed
for. The functional demands on credited equipment is unchanged.
There are no new failure modes or mechanisms created due to plant
operation for an extended period to perform DG maintenance or
testing. Extended operation with an inoperable DG does not involve
any modification to the operational limits or physical design of
plant systems. There are no new accident precursors generated due to
the extended CT.
Regarding the proposed change to add Required Action to ensure
that at least one train of shared components has an operable
emergency power supply, there is no change to how or under what
conditions offsite circuits or DGs are operated nor are there any
changes to acceptable operating parameters. Power source operability
requirements for shared components are being moved from the TS Bases
to TS with the proposed change. The proposed change will ensure that
at least one train of shared components has an operable emergency
power supply whenever a DG is inoperable. This change does not alter
the nature of events postulated in the Updated Final Safety Analysis
Report nor does it introduce any unique precursor mechanisms.
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 the margin of safety?
Response: No.
The proposed change involves extending the TS CT for an
inoperable DG at [. . .] MNS. The proposed change also involves
adding a new Required Action to TSs to ensure that at least one
train of shared components at [. . .] MNS has an operable emergency
power supply whenever one DG is inoperable.
Currently, if an inoperable DG is not restored to operable
status within 72 hours at [. . .] MNS, TS 3.8.1, requires the units
to be in Mode 3 (i.e., Hot Standby) within a CT of 6 hours, and to
be in Mode 5 (i.e., Cold Shutdown) within a CT of 36 hours. The
proposed TS changes will allow steady state plant operation at 100
percent power for an additional 11 days for performance of DG
planned reliability improvements and preventive and corrective
maintenance.
Deterministic and probabilistic risk assessment techniques
evaluated the effect of the proposed TS change to extend the CT for
an inoperable DG on the availability of an electrical power supply
to the plant emergency safeguards feature systems. These assessments
concluded that the proposed [. . .] MNS TS change does not involve a
significant increase in the risk of power supply unavailability.
The DGs continue to meet their design requirements; there is no
reduction in capability or change in design configuration. The DG
response to loss of offsite power, loss of coolant accident, station
blackout or fire scenarios is not changed by this proposed
amendment; there is no change to the DG operating parameters. In the
extended CT, as in the existing CT, the remaining operable DGs and
paths are adequate to supply electrical power to the onsite
emergency power distribution system. The proposed change to extend
the CT for an inoperable DG does not alter a design basis safety
limit; therefore, it does not significantly reduce the margin of
safety. The DGs will continue to operate per the existing design and
regulatory requirements.
The proposed TS changes (i.e., the inoperable DG CT extension
request and proposed change to add Required Action to ensure that at
least one train of shared components has an operable emergency power
supply) do not alter the plant design nor do they change the
assumptions contained in the safety analyses. The standby AC power
system is designed with sufficient redundancy such that a DG may be
removed from service for maintenance or testing. The remaining DGs
are capable of carrying sufficient electrical loads to satisfy the
Updated Final Safety Analysis Report requirements for accident
mitigation or unit safe shutdown. The proposed change does not
impact the redundancy or availability requirements of offsite power
circuits or change the ability of the plant to cope with a station
blackout. Therefore, based on the considerations given above, the
proposed changes do not involve a significant reduction in the
margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2 (ANO-2), Pope County, Arkansas
Date of amendment request: November 20, 2017. A publicly-available
version is in ADAMS under Accession No. ML17326A387.
Description of amendment request: The amendment would revise the
Technical Specifications (TSs) to replace the current pressure-
temperature limits for heatup, cooldown, and the inservice leak
hydrostatic tests for the reactor coolant system presented in TS 3.4.9
that expire at 32 Effective Full Power Years (EFPY) with limitations
that extend out to 54 EFPY.
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 will revise the pressure-temperature (P-T)
limits for heatup, cooldown, and inservice leak hydrostatic test
limitations for the Reactor Coolant System
[[Page 8515]]
(RCS) to a maximum of 54 Effective Full Power Years (EFPY) in
accordance with 10 CFR 50, Appendix G. This is the end of the period
of extended operation for the renewed ANO-2 operating License. The
P-T limits were developed in accordance with the requirements of 10
CFR 50, Appendix G, utilizing the analytical methods and flaw
acceptance criteria of Topical Report WCAP-14040, Revision 4, and
American Society of Mechanical Engineers (ASME) Code, Section XI,
Appendix G. These methods and criteria are the previously NRC
approved standards for the preparation of P-T limits. Updating the
P-T limits for additional EFPYs maintains the level of assurance
that reactor coolant pressure boundary integrity will be maintained,
as specified in 10 CFR 50, Appendix G.
The proposed changes do not adversely affect accident initiators
or precursors, and do not alter the design assumptions, conditions,
or configuration of the plant or the manner in which the plant is
operated and maintained. The ability of structures, systems, and
components to perform their intended safety functions is not altered
or prevented by the proposed changes, and the assumptions used in
determining the radiological consequences of previously evaluated
accidents are not affected.
Therefore, this 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 changes implement methodologies that have been
approved by the NRC (provided that any conditions/limitations are
satisfied). The P-T limits will ensure the protection consistent
with assuring the integrity of the reactor coolant pressure boundary
as was previously evaluated. Reactor coolant pressure boundary
integrity will continue to be maintained in accordance with 10 CFR
50, Appendix G, and the assumed accident performance of plant
structures, systems and components will not be affected. These
changes do not involve any physical alteration of the plant (i.e.,
no new or different type of equipment will be installed), and
installed equipment is not being operated in a new or different
manner. Thus, no new failure modes are introduced.
Therefore, this change does not create the possibility of a new
or different kind of accident from an accident 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 function of the reactor
coolant pressure boundary or its response during plant transients.
By calculating the P-T limits using NRC-approved methodology,
adequate margins of safety relating to reactor coolant pressure
boundary integrity are maintained. The proposed changes do not alter
the manner in which safety limits, limiting safety system settings,
or limiting conditions for operation are determined. These changes
will ensure that protective actions are initiated and the
operability requirements for equipment assumed to operate for
accident mitigation are not affected.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel,
Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East,
Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2 (ANO-2), Pope County, Arkansas
Date of amendment request: December 14, 2017. A publicly-available
version is in ADAMS under Accession No. ML17348A150.
Description of amendment request: The amendment would revise ANO-2
Technical Specification (TS) 3.3.3.6, ``Post-Accident
Instrumentation,'' to ensure that both Category 1 and Type A Regulatory
Guide (RG) 1.97, Revision 3, ``Instrumentation for Light-Water-Cooled
Nuclear Power Plants to Assess Plant and Environs Conditions During and
Following an Accident,'' instrumentation is included in the
specification (unless already addressed within another specification)
and gains greater consistency with NUREG-1432, Revision 4, ``Standard
Technical Specifications for Combustion Engineering Plants.''
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 PAM [Post-Accident Monitoring] instrumentation is not an
initiator of any design basis accident or event and, therefore, the
proposed change does not increase the probability of any accident
previously evaluated. The proposed change ensures required
instrumentation is included in and controlled by the station TSs and
does not change the response of the plant to any accidents.
The proposed change does not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained. The removal and addition of
specific instrumentation within ANO-2 TS 3.3.3.6 is consistent with
the ANO-2 SAR [Safety Analysis Report], Table 7.5-3 RG 1.97
variables classified as Type A or Category 1 variables.
Modifications to the TS Actions associated with inoperable
instrumentation are consistent with the current ANO-2 licensing
basis or act to improve consistency with NUREG 1432. The proposed
change does not adversely affect the ability of structures, systems,
and components (SSCs) to perform the associated intended safety
function to mitigate the consequences of an initiating event within
the assumed acceptance limits. The proposed change does not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of any
accident previously evaluated. Further, the proposed change does not
increase the types and amounts of radioactive effluent that may be
released offsite, nor significantly increase individual or
cumulative occupational/public radiation exposures.
Instrumentation that does not meet the RG 1.97 inclusion
criteria as established in NUREG-1432 are removed from the TS;
however, the instrumentation remains applicable to other RG 1.97
criteria and is maintained accordingly. Instrumentation added to the
ANO-2 PAM TS does not change the manner in which the instrumentation
is currently maintained since these instruments are currently
designated as Type A and/or Category 1 variables in the ANO-2 SAR.
However, including these instruments within the TSs will now require
different mitigating actions during periods of inoperability, which
may include a plant shutdown, establishment of alternate monitoring
methods, and/or submittal of a special report to the NRC.
Therefore, this 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 result in a change in the manner in
which the plant is operated during post-accident conditions and does
not change the established mitigating actions associated with any
necessary response to a DBA [design-basis accident]. The proposed
change continues to ensure important instrumentation remains
available to station operators such that currently established
mitigating actions are not impacted. 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 or post-accident plant operation. The change does not alter
assumptions made in the safety analysis.
Therefore, this change does not create the possibility of a new
or different kind of accident from an accident previously evaluated.
[[Page 8516]]
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The safety analysis acceptance criteria
and assumptions are not impacted by the proposed change. The
proposed change will not result in plant operation in a
configuration outside the design basis. The proposed change ensures
appropriate PAM instrumentation is controlled by the station TSs and
that specified remedial action will be taken when required
instrumentation is inoperable. The proposed change continues to
support the operator ability to monitor and control vital systems
during post-accident conditions.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel,
Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East,
Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., System Energy Resources, Inc., Cooperative
Energy, A Mississippi Electric Cooperative, and Entergy Mississippi,
Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS),
Claiborne County, Mississippi
Date of amendment request: November 3, 2017, as supplemented by
letters dated December 6, 2017, and January 22, 2018. Publicly-
available versions are in ADAMS under Accession Nos. ML17307A440,
ML17340B025, and ML18022A598, respectively.
Description of amendment request: The amendment would revise the
GGNS Updated Final Safety Analysis Report (UFSAR) to incorporate the
Tornado Missile Risk Evaluator (TMRE) methodology contained in Nuclear
Energy Institute (NEI) 17-02, Revision 1, ``Tornado Missile Risk (TMRE)
Industry Guidance Document,'' September 2017 (ADAMS Accession No.
ML17268A036). This methodology can only be applied to discovered
conditions where tornado missile protection is not currently provided,
and cannot be used to avoid providing tornado missile protection in the
plant modification process.
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:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The proposed amendment is to incorporate the TMRE methodology
into the GGNS UFSAR. The TMRE methodology is an alternative
methodology for determining whether protection from tornado-
generated missiles is required. The methodology can only be applied
to discovered conditions where tornado missile protection was not
provided, and cannot be used to avoid providing tornado missile
protection in the plant modification process.
The proposed amendment does not involve an increase in the
probability of an accident previously evaluated. The relevant
accident previously evaluated is a Design Basis Tornado impacting
the GGNS site. The probability of a Design Basis Tornado is driven
by external factors and is not affected by the proposed amendment.
There are no changes required to any of the previously evaluated
accidents in the UFSAR.
The proposed amendment does not involve a significant increase
in the consequences or a Design Basis Tornado. [The methodology as
proposed does not alter any input assumptions or results of the
accident analyses. Instead, it reflects a methodology to more
realistically evaluate the probability of unacceptable consequences
of a Design Basis Tornado. As such, there is no significant increase
in the consequence of an accident previously evaluated. A similar
consideration would apply in the event additional non-conforming
conditions are discovered in the future.]
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: No.
The proposed amendment is to incorporate the TMRE methodology
into the GGNS UFSAR. The TMRE methodology is an alternative
methodology for determining whether protection from tornado-
generated missiles is required. The methodology can only be applied
to discovered conditions where tornado missile protection was not
provided, and cannot be used to avoid providing tornado missile
protection in the plant modification process.
The proposed amendment will involve no physical changes to the
existing plant, so no new malfunctions could create the possibility
of a new or different kind of accident. The proposed amendment makes
no changes to conditions external to the plant that could create the
possibility of a new or different kind of accident. The proposed
change will not create the possibility of a new or different kind of
accident due to new accident precursors, failure mechanisms,
malfunctions, or accident initiators not considered in the design
and licensing bases. The existing Updated Final Safety Analysis
Report accident analysis will continue to meet requirements for the
scope and type of accidents that require analysis.
Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident than those
previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
The proposed amendment is to incorporate the TMRE methodology
into the GGNS UFSAR. The TMRE methodology is an alternative
methodology for determining whether protection from tornado-
generated missiles is required. The methodology can only be applied
to discovered conditions where tornado missile protection was not
provided, and cannot be used to avoid providing tornado missile
protection in the plant modification process.
The change does not exceed or alter any controlling numerical
value for a parameter established in the UFSAR or elsewhere in the
GGNS licensing basis related to design basis or safety limits. The
change does not impact any UFSAR Chapter 6 or 15 Safety Analyses,
and those analyses remain valid. The change does not reduce
diversity or redundancy as required by regulation or credited in the
UFSAR. The change does not reduce defense-in-depth as described in
the UFSAR.
Therefore, the changes associated with this license amendment
request do not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's modified 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: William B. Glew, Associate General Counsel,
Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New York
10601.
NRC Branch Chief: Douglas A. Broaddus.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of amendment request: December 21, 2017. A publicly-available
version is in ADAMS under Accession No. ML17355A184.
Description of amendment request: The amendments would revise the
Technical Specifications (TSs) pertaining to the Engineered Safety
Features Actuation System instrumentation to resolve non-conservative
actions associated with the containment ventilation isolation and the
control room ventilation isolation functions. In addition, the
amendments would revise the control room
[[Page 8517]]
ventilation isolation function to no longer credit containment
radiation monitoring instrumentation, eliminate redundant radiation
monitoring instrumentation requirements, eliminate select core
alterations applicability requirements, relocate radiation monitoring
and reactor coolant system leakage detection requirements within the
TSs to align with their respective functions, and relocate the spent
fuel pool area monitoring requirements to licensee-controlled
documents.
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 instrumentation associated with the proposed changes to the
technical specifications (TS) is not an initiator of any accidents
previously evaluated, so the probability of accidents previously
evaluated is unaffected by the proposed changes. There is no change
to any equipment response or accident scenario, with the exception
of the Control Room isolation on Containment high-radiation
instrumentation function which impose no additional challenges to
fission product barrier integrity. The exception is supported by
revised radiological analyses which demonstrate that the Control
Room air intake radioactivity monitoring instrumentation provides
timely automatic isolation of the Control Room ventilation system
and thereby limits Control Room operator doses to within regulatory
limits for any design basis accident. The proposed changes also
eliminate limitations imposed on Containment and Control Room
ventilation instrumentation during CORE ALTERATIONS since the
applicable postulated accidents do not result in fuel cladding
integrity damage. Hence, the capability of any TS-required SSC
[structure, system, or component] to perform its specified safety
function is not impacted by the proposed changes and the outcomes of
accidents previously evaluated are unaffected. Therefore, the
proposed changes do not result in a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
The changes do not challenge the integrity or performance of any
safety-related systems. No plant equipment is installed or removed,
and the changes do not alter the design, configuration, or method of
operation of any plant SSC with the exception of the Control Room
isolation on Containment high-radiation instrumentation function
which is supported by revised accident analyses which demonstrate
that the radiological consequences remain within applicable
regulatory limits. The elimination of core alterations applicability
requirements do not impact the outcome of any applicable postulated
accident since none result in fuel cladding damage. No physical
changes are made to the plant, so no new causal mechanisms are
introduced. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The ability of any operable SSC to perform its designated safety
function is unaffected by the proposed changes. The proposed change
do not revise any safety limits or limiting safety system settings.
The proposed changes revises safety analyses assumptions and the
method of operating the plant with regard to the Control Room
isolation on Containment high-radiation instrumentation function.
The changes are supported by revised accident analyses which
demonstrate that no adverse impact will result to either the plant
operating margins or the reliability of equipment credited in the
safety analyses. The existing margin in dose assessment currently
afforded Control Room operators during any design basis accident is
maintained. No other safety margins are impacted by the proposed
changes. Therefore, the proposed changes do 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: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd. MS LAW/JB,
Juno Beach, FL 33408-0420.
NRC Branch Chief: Undine Shoop.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: November 10, 2017. A publicly-available
version is in ADAMS under Accession No. ML17318A240.
Description of amendment request: The proposed amendment revises
Technical Specification (TS) 3.6.4.1, ``Secondary Containment,''
Surveillance Requirement (SR) 3.6.4.1.2. The SR is modified to
acknowledge that secondary containment access openings may be open for
entry and exit.
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 addresses conditions during which the
secondary containment SR is not met. The secondary containment is
not an initiator of any accident previously evaluated. As a result,
the probability of any accident previously evaluated is not
increased. The consequences of an accident previously evaluated
while utilizing the proposed changes are no different than the
consequences of an accident while utilizing the existing four-hour
Completion Time for an inoperable secondary containment. As a
result, the consequences of an accident previously evaluated are not
significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed change does not alter the protection system design,
create new failure modes, or change any modes of operation. The
proposed change does not involve a physical alteration of the plant;
and no new or different kind of equipment will be installed.
Consequently, there are no new initiators that could result in a new
or different kind of accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change addresses conditions during which the
secondary containment SR is not met. The allowance for both an inner
and outer secondary containment door to be open simultaneously for
entry and exit does not affect the safety function of the secondary
containment as the doors are promptly closed after entry or exit,
thereby restoring the secondary containment boundary.
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: William Blair, P. O. Box 14000, Juno Beach,
FL 33408-0420.
NRC Branch Chief: David J. Wrona.
[[Page 8518]]
Northern States Power Company--Minnesota (NSPM), Docket No. 50-263,
Monticello Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: December 19, 2017. A publicly-available
version is in ADAMS under Accession No. ML17353A189.
Description of amendment request: The proposed amendment would
adopt Technical Specifications Task Force (TSTF) traveler TSTF-425,
``Relocate Surveillance Frequencies to Licensee Control--RITSTF [Risk-
Informed Technical Specifications Task Force Initiative 5b,'' Revision
3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change relocates the specified frequencies for
periodic surveillance requirements to licensee control under a new
SFCP [Surveillance Frequency Control Program]. Surveillance
frequencies are not an initiator to any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The systems and components
required by the technical specifications for which the surveillance
frequencies are relocated are still required to be operable, meet
the acceptance criteria for the surveillance requirements, and be
capable of performing any mitigation function assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly increased.
Therefore, the proposed 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.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for systems, structures, and components (SSCs), specified in
applicable codes and standards (or alternatives approved for use by
the NRC) will continue to be met as described in the plant licensing
basis (including the final safety analysis report and bases to TS),
since these are not affected by changes to the surveillance
frequencies. Similarly, there is no impact to safety analysis
acceptance criteria as described in the plant licensing basis. To
evaluate a change in the relocated surveillance frequency, [NSPM]
will perform a probabilistic risk evaluation using the guidance
contained in NRC approved NEI 04-10, Rev. 1 in accordance with the
TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable
acceptance guidelines and methods for evaluating the risk increase
of proposed changes to surveillance frequencies consistent with
Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: David J. Wrona.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: November 30, 2017. A publicly-available
version is in ADAMS under Accession No. ML17334B211.
Description of amendment request: The proposed changes include
changes to the Updated Final Safety Analysis Report (UFSAR) in the form
of departures from the incorporated plant-specific Design Control
Document (DCD) Tier 2* and Tier 1 information and related changes to
the VEGP Units 3 and 4 Combined License (COL) Appendix C information.
Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from the
elements of the design as certified in 10 CFR part 52, Appendix D,
design certification rule is also requested for the plant-specific Tier
1 material departures. This submittal requests approval of the license
amendment, necessary to implement these changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), licensee has provided
its analysis of the issue on no significant hazards consideration
determination, 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 consistency and editorial changes to COL Appendix C
(and associated plant-specific Tier 1) and Tier 2 and Tier 2*
information in the UFSAR do not involve a technical change, (e.g.
there is no design parameter or requirement, calculation, analysis,
function or qualification change). No structure, requirement,
calculation, analysis, function or qualification change). No
structure, system, or component (SSC) design or function would be
affected. No design or safety analysis would be affected. The
proposed changes do not affect any accident initiating event or
component failure, thus the probabilities of the accidents
previously evaluated are not affected. No function used to mitigate
a radioactive material release and no radioactive material release
source term is involved, thus the radiological releases in the
accident analyses are not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed consistency and editorial changes to COL Appendix C
(and associated plant specific Tier 1) and Tier 2 and Tier 2*
information in the UFSAR do not change the design or functionality
of safety-related SSCs. The proposed change does not affect plant
electrical systems, and does not affect the design function,
support, design, or operation of mechanical and fluid systems. The
proposed change does not result in a new failure mechanism or
introduce any new accident precursors. No design function described
in the UFSAR is affected by the proposed changes.
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 consistency and editorial changes to COL Appendix C
(and associated plant specific Tier 1) and Tier 2 and Tier 2*
information in the UFSAR do not involve any change to the design as
described in the COL. There would be no change to an existing design
basis, design function, regulatory criterion, or analysis. No safety
analysis or design basis acceptance limit/criterion is involved.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
[[Page 8519]]
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 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: February 1, 2018. A publicly-available
version is in ADAMS under Accession No. ML18032A359.
Description of amendment request: The requested amendment proposes
changes to relax the minimum gap requirement above grade between the
nuclear island and the annex building/turbine building and removing the
minimum gap requirement for the radwaste building from the Inspections,
Tests, Analyses and Acceptance Criteria. 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 with NRC staff edits in square
brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes are to relax the minimum gap requirement
above grade between the nuclear island and the annex building/
turbine building from a 4 inch gap to a 3 inch gap. The proposed
changes modify and clarify the gap requirements between the nuclear
island and the annex building/turbine building and radwaste
building, respectively. The proposed change deletes the gap
requirement for the radwaste building from the Inspections, Tests,
Analyses and Acceptance Criteria (ITAAC) in (COL) [Combined License]
Appendix C. 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 structure, system or component (SSC) such
that a new accident initiator or initiating sequence of events is
created.
The changes do not impact the support, design, or operation of
mechanical and fluid systems. The changes do not impact the support,
design, or operation of any safety-related structures. 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 do the proposed changes 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 changes are to relax the minimum gap requirement
above grade between the nuclear island and the annex building/
turbine building from a 4 inch gap to a 3 inch gap. The proposed
changes modify and clarify the gap requirements between the nuclear
island and the annex building/turbine building and radwaste
building, respectively. The proposed changes delete the gap
requirement for the radwaste building from the ITAAC in COL Appendix
C. 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 design function
of the nuclear island and adjoining buildings' 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 non-safety-related equipment. 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 result
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 margin and provide
adequate protection through continued application of the existing
requirements in the UFSAR [Updated Final Safety Analysis Report].
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.
Because no safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by these changes, no significant
margin of safety is reduced.
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: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County,
Georgia
Date of amendment request: January 31, 2018. A publicly-available
version is in ADAMS under Accession No. ML18031B142.
Description of amendment request: The requested amendment proposes
to include changes to Combined License (COL) Appendix A, Technical
Specifications related to fuel management. Specifically, the requested
amendment proposes improvements to the technical specifications for the
Rod Position Indication, the Control Rod Drive Mechanism, Power Range
Neutron Flux Channels and the Mechanical Shim Augmentation.
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 are to clarify proper operation and
methodology associated with the DRPI [Digital Rod Position
Indication], Control Rod Gripper Coils, instrumentation associated
with Quadrant Power Tilt Ratio, or Control or Gray Rods. These
changes do not affect the operation of this equipment and have no
adverse impact on their design functions.
The changes do not involve an interface with any structure,
system, or component (SSC) accident initiator or initiating sequence
of events, and thus, the probabilities of the accidents evaluated in
the plant-specific Updated Final Safety Analysis Report (UFSAR) are
not affected. The proposed changes do not adversely affect any
mitigation sequence or the predicted radiological releases due to
postulated accident conditions, thus, the consequences of the
accidents evaluated in the UFSAR are not affected.
[[Page 8520]]
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 verify and maintain the capabilities of the
DRPI, Control Rod Gripper Coils, instrumentation associated with
Quadrant Power Tilt Ratio, and Control and Gray Rods to perform
their design functions. 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 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 result in significant fuel cladding failures. These
changes are to clarify proper operation and methodology associated
with this equipment and have no adverse impact on their design
functions.
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 do not affect existing safety margins. The
proposed changes verify and maintain the capabilities of the DRPI,
Control Rod Gripper Coils, instrumentation associated with Quadrant
Power Tilt Ratio, and Control and Gray Rods to perform their design
functions. 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 affect any design code, function,
design analysis, safety analysis input or result, or design/safety
margin.
The proposed changes would not affect any safety-related design
code, function, design analysis, safety analysis input or result, or
existing design/safety margin. Because no safety analysis or design
basis acceptance limit/criterion is challenged or exceeded by the
requested changes, no margin of safety is significantly reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
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: December 14, 2017. A publicly-available
version is in ADAMS under Accession No. ML17348B097.
Description of amendment request: The amendments would revise
Technical Specification (TS) 3.6.4.1, ``Secondary Containment,''
Surveillance Requirement (SR) 3.6.4.1.1. The SR would be revised to
address conditions during which the secondary containment pressure may
not meet the SR pressure requirements. The proposed changes are based
on Technical Specifications Task Force (TSTF) Traveler TSTF-551,
Revision 3, ``Revise Secondary Containment Surveillance Requirements.''
Also, the editorial note in SR 3.6.4.1.3 is removed because it is
redundant to the SR itself and does not alter the 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, 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 change addresses conditions during which the
secondary containment SR is not met. The secondary containment is
not an initiator of any accident previously evaluated. As a result,
the probability of any accident previously evaluated is not
increased. The consequences of an accident previously evaluated
while utilizing the proposed changes are no different than the
consequences of an accident while utilizing the existing four hour
Completion Time for an inoperable secondary containment. In
addition, the proposed Note for SR 3.6.4.1.1 provides an alternative
means to ensure the secondary containment safety function is met.
Additionally, the Note removed from SR 3.6.4.1.3 is editorial
because it is redundant to the SR itself and does not alter the
requirement. As a result, the consequences of an accident previously
evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the protection system design,
create new failure modes, or change any modes of operation. The
proposed change does not involve a physical alteration of the plant;
and no new or different kind of equipment will be installed.
Consequently, there are no new initiators that could result in a new
or different kind of accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change addresses conditions during which the
secondary containment SR is not met. Conditions in which the
secondary containment vacuum is less than the required vacuum are
acceptable provided the conditions do not affect the ability of the
SGT [Standby Gas Treatment] System to establish the required
secondary containment vacuum under post-accident conditions within
the time assumed in the accident analysis. This condition is
incorporated in the proposed change by requiring an analysis of
actual environmental and secondary containment pressure conditions
to confirm the capability of the SGT System is maintained within the
assumptions of the accident analysis. Therefore, the safety function
of the secondary containment is not affected.
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: Damon D. Obie, Associate General Counsel,
Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA
18101.
NRC Branch Chief: James G. Danna.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: September 29, 2017. A publicly-available
version is in ADAMS under Accession No. ML17272A940.
Description of amendment request: The amendments would make changes
to the SQN Emergency Plan to extend staff augmentation times for
Emergency Response Organization (ERO) functions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 8521]]
1. Does the proposed change involve a significant increase in
the probability or consequence of an accident previously evaluated?
Response: No.
The proposed removal of maintenance personnel from shift and
extension in staff augmentation times has no effect on normal plant
operation or on any accident initiator or precursor and does not
affect the function of plant structures, systems, or components
(SCCs). The proposed changes do not alter or prevent the ability of
the ERO to perform their intended functions to mitigate the
consequences of an accident or event. The ability of the ERO to
respond adequately to radiological emergencies has been demonstrated
as acceptable through a staffing analysis as required by 10 CFR 50
Appendix E.IV.A.9.
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.
The proposed changes do not affect the accident analyses. The
changes do not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed), a change in
the method of plant operation, or new operator actions. The proposed
changes do not introduce failure modes that could result in a new
accident, and the changes do not alter assumptions made in the
safety analysis. This proposed change removes maintenance personnel
from shift and extends the staff augmentation response times in the
SQN Emergency Plan, which are demonstrated as acceptable through a
staffing analysis as required by 10 CFR 50 Appendix E.IV.A.9. The
proposed changes do not alter or prevent the ability of the ERO to
perform their intended functions to mitigate the consequences of an
accident or event.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in
margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed change is
associated with the SQN Emergency Plan staffing and does not affect
operation of the plant or its response to transients or accidents.
The change does not affect the Technical Specifications. The
proposed changes do not involve a change in the method of plant
operation, and no accident analyses are affected by the proposed
changes. Safety analysis acceptance criteria are not affected by
this proposed change. A staffing analysis and a functional analysis
were performed for the proposed changes on the timeliness of
performing major tasks for the functional areas of the SQN Emergency
Plan. The analysis concluded that removal of maintenance personnel
from shift and an extension in staff augmentation times would not
significantly affect the ability to perform the required Emergency
Plan tasks.
Therefore, the proposed changes are determined to not adversely
affect the ability to meet 10 CFR 50.54(q)(2), the requirements of
10 CFR 50 Appendix E, and the emergency planning standards as
described in 10 CFR 50.47(b).
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: Undine Shoop.
III. Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses and Combined Licenses,
Proposed No Significant Hazards Consideration Determination, and
Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Florida Power & Light Company, Docket Nos. 50-250, Turkey Point Nuclear
Generating Unit No. 3, Miami-Dade County, Florida
Date of amendment request: December 18, 2017. A publicly-available
version is in ADAMS under ML17353A492.
Brief description of amendment request: Revise the Technical
Specifications to allow a one-time extension of the allowable outage
time for the Unit 3 Containment Spray System from 72 hours to 14 days.
Date of publication of individual notice in Federal Register:
January 30, 2018 (83 FR 4285).
Expiration date of individual notice: March 1, 2018 (Public
comments); April 2, 2018 (Hearing requests).
IV. Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
[[Page 8522]]
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, Wake County, North Carolina
Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: July 18, 2017, as supplemented by letter
dated October 12, 2017.
Brief description of amendments: The amendments revised the
technical specifications (TSs) based on Technical Specifications Task
Force (TSTF) Traveler TSTF-529, ``Clarify Use and Application Rules.''
The changes revise and clarify the TS usage rules for completion times,
limiting conditions for operation, and surveillance requirements.
Date of issuance: February 1, 2018.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 298 and 294, for the Catawba Nuclear Station Units
1 and 2; 307 and 286, for the McGuire Nuclear Station, Units 1 and 2;
407, 409, and 408, for the Oconee Nuclear Station, Units 1, 2, and 3;
162, for the Shearon Harris Nuclear Power Plant, Unit 1; and 256, for
the H. B. Robinson Steam Electric Plant, Unit No. 2. A publicly-
available version is in ADAMS under Accession No. ML17340A720;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, NPF-
17, DPR-38, DPR-47, DPR-55, NPF-63, and DPR-23: Amendments revised the
Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: August 29, 2017 (82 FR
41067). The supplemental letter dated October 12, 2017, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated February 1, 2018.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2 (McGuire), Mecklenburg County, North
Carolina
Date of amendment requests: December 19, 2016, as supplemented by
letters dated May 25, 2017, and December 12, 2017.
Brief description of amendments: The amendments modified Technical
Specification 5.5.2, ``Containment Leakage Rate Testing Program,'' by
replacing the reference to Regulatory Guide 1.163, ``Performance-Based
Containment Leak-Test Program,'' with a reference to Nuclear Energy
Institute (NEI) Topical Report NEI 94-01, Revision 3-A, ``Industry
Guideline for Implementing Performance-Based Option of 10 CFR part 50,
Appendix J,'' dated July 2012 and the conditions and limitations
specified in NEI 94-01, Revisions 2-A, ``Industry Guideline for
Implementing Performance-Based Option of 10 CFR part 50, Appendix J,''
dated October 2008, as the implementation documents used by McGuire to
implement the performance-based leakage testing program in accordance
with Option B of 10 CFR part 50, Appendix J. The proposed change would
also delete the listing of one-time exceptions previously granted to
Integrated Leak Rate Test frequency.
Date of issuance: January 31, 2018.
Effective date: As of its date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos.: 306 (Unit 1) and 285 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML18009A842; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: May 9, 2017 (82 FR
21557). The supplemental letters dated May 25 and December 12, 2017,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 31, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station (LSCS), Units 1 and 2, LaSalle County, Illinois
Date of amendment request: August 29, 2017, as supplemented by
letter dated January 25, 2018.
Brief description of amendment: The amendments revised the LSCS
technical specification (TS) 2.1.1, ``Reactor Core SLs [Safety
Limits].'' Specifically, this change incorporates revised LSCS, Units 1
and 2, safety limits for minimum critical power ratio for two
circulation loop minimum critical power ratio (MCPR) and single
circulation loop MCPR values for Unit 1 and Unit 2 based on the results
of the cycle-specific analyses performed by Global Nuclear Fuel (GNF)
for LSCS Unit 1, Cycle 17, and LSCS Unit 2, Cycle 17.
Date of issuance: February 6, 2018.
Effective date: As of the date of issuance and shall be implemented
as follows:
Unit 1: Prior to startup from the February 2018 refueling outage
for Unit 1 (i.e., L1R17) for operation starting in Cycle 18.
Unit 2: Prior to startup from the February 2018 refueling outage
for Unit 1 (i.e., L1R17). This will be a mid-Cycle 17 implementation
for Unit 2.
Amendment No.: Unit 1-227; Unit 2-213. A publicly-available version
is in ADAMS under Accession No. ML18008A123; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. NPF-11 and NPF-18: The
amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: December 5, 2017 (82 FR
57482). The supplemental letter dated January 25, 2018, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 6, 2018.
[[Page 8523]]
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 24, 2017, as supplemented by letter
dated August 17, 2017.
Brief description of amendments: The amendments revised
Surveillance Requirement 3.3.1.3 to change the thermal power at which
the surveillance may be performed.
Date of issuance: February 7, 2018.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 194 (Unit 1) and 177 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML18012A068; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License No. NPF-68 and NPF-81:
Amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: July 18, 2017 (82 FR
32883). The supplemental letter dated August 17, 2017, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 7, 2018.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment requests: March 13, 2017, as supplemented by
letter dated August 7, 2017.
Brief description of amendments: The amendments deleted the Note
associated with Technical Specification (TS) Surveillance Requirement
(SR) 3.8.1.17 to allow the performance of the SR in Modes 1 through 4.
Date of issuance: February 2, 2018.
Effective date: As of its date of issuance and shall be implemented
no later than 60 days from the date of issuance.
Amendment Nos.: 340 (Unit 1) and 333 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17296A133; documents related
to these amendments are listed in the Safety Evaluation (SE) enclosed
with the amendments.
Facility Operating License Nos. DPR-77 and DPR-79: The amendments
revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: July 5, 2017 (82 FR
31102). The supplemental letter dated August 7, 2017, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in an SE dated February 2, 2018.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: January 20, 2017, as supplemented by
letter dated September 7, 2017.
Brief description of amendments: The amendments revised the
Technical Specification (TS) 3.5, ``Residual Heat Removal (RHR)
System,'' requirements, as well as the TS 3.13, ``Component Cooling
System,'' RHR support requirements for consistency with the design
basis of the RHR system. In addition, an RHR surveillance requirement
is added in TS Table 4.1-2A, ``Minimum Frequency for Equipment Tests,''
to test the RHR system in accordance with the inservice testing
program, since a TS surveillance does not currently exist for this
system.
Date of issuance: February 9, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 291 and 291. A publicly-available version is in
ADAMS under Accession No. ML17326A225; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License No. DPR-32 and DPR-37:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: March 14, 2017 (82 FR
13672). The supplemental letter dated September 7, 2017, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 9, 2018.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, on February 20, 2018.
For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2018-03727 Filed 2-26-18; 8:45 am]
BILLING CODE 7590-01-P