[Federal Register Volume 83, Number 49 (Tuesday, March 13, 2018)]
[Notices]
[Pages 10911-10927]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-04827]
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NUCLEAR REGULATORY COMMISSION
[NRC-2018-0045]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954,
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued, and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued, from February 13 to February 26, 2018. The last
biweekly notice was published on February 27, 2018.
DATES: Comments must be filed by April 12, 2018. A request for a
hearing must be filed by May 14, 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-2018-0045. 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.
[[Page 10912]]
FOR FURTHER INFORMATION CONTACT: Janet Burkhardt, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-1384, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2018-0045, 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-2018-0045.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in this
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2018-0045, 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
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
[[Page 10913]]
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
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
[[Page 10914]]
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.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: September 14, 2017. A publicly-available
version is in ADAMS under Accession No. ML17261B255.
Description of amendment request: The amendments would modify
Catawba Nuclear Station, Units 1 and 2, Technical Specification (TS)
3.7.8, ``Nuclear Service Water System (NSWS).'' Specifically, the
proposed change would add a new Condition D for one NSWS pond return
header being inoperable due to the NSWS being aligned for single pond
return header operation with a Completion Time (CT) of 30 days. This
would involve isolating one train of the NSWS pond return piping at the
Auxiliary Building wall and maintaining the discharge crossover lines
open between trains in the Auxiliary Building and Emergency Diesel
Generator Buildings. This provides a common safety-related discharge
path through the single remaining in-service pond return line. This
alignment, single pond return header operation, allows a pond return
header to be removed from service while a flow path is maintained
through both trains of NSWS supplied equipment to the Standby Nuclear
Service Water Pond (SNSWP).
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 single pond return header operation configuration
for NSWS operation and the associated proposed TS and TS Bases
changes have been evaluated to assess their impact on plant
operation and to ensure that the design basis safety functions of
safety related systems are not adversely impacted. During single
pond return header operation, the operating NSWS header will be able
to discharge all required NSWS flow from safety related components.
PRA [Probabilistic Risk Assessment] has demonstrated that due to the
limited proposed time in the single pond return header
configuration, the resultant plant risk remains acceptable.
The purpose of this amendment request is to ultimately
facilitate inspections and modifications of the NSWS Pond Return
buried piping between the Auxiliary Building and the Discharge to
the SNSWP. Therefore, NRC approval of this request will ultimately
help to enhance the long-term structural integrity of the NSWS and
will help to ensure the system's reliability for many years.
In general, the NSWS serves as an accident mitigation system and
cannot by itself initiate an accident or transient situation. The
only exception is that the NSWS piping can serve as a source of
floodwater to safety related equipment in the Auxiliary Building or
in the diesel generator buildings in the event of a leak or a break
in the system piping. The probability of such an event is not
significantly increased as a result of this proposed request. Safety
related NSWS piping is tested and inspected in accordance with all
applicable in-service testing and in-service inspection
requirements. Given the negligible influence of flooding events on
the NSWS for the submittal configuration (i.e., no dominant
contribution from floods), it is judged that the analyses assessing
the influence of these events provide an acceptable evaluation of
the contribution of the flood risk for the requested CT of 30 days.
The proposed 30 day TS Required Action CT has been evaluated for
risk significance and the results of this evaluation have been found
acceptable. The probabilities of occurrence of accidents presented
in the UFSAR [Updated Final Safety Analysis Report] will not
increase as a result of implementation of this change. Because the
PRA analysis supporting the proposed change yielded acceptable
results, the NSWS will maintain its required availability in
response to accident situations. Since NSWS availability is
maintained, the response of the plant to accident situations will
remain acceptable and the consequences of accidents presented in the
UFSAR will not increase.
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.
Implementation of this amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed request does not affect the basic operation
of the NSWS or any of the systems that it supports. These include
the Emergency Core Cooling System, the Containment Spray System, the
Containment Valve Injection Water System, the Auxiliary Feedwater
System, the Component Cooling Water System, the Control Room Area
Ventilation System, the Control Room Area Chilled Water System,
[[Page 10915]]
the Auxiliary Building Filtered Ventilation Exhaust System, or the
Diesel Generators. During proposed single pond return header
operation, the NSWS will remain capable of fulfilling all of its
design basis requirements.
No new accident causal mechanisms are created as a result of NRC
approval of this amendment request. No changes are being made to the
plant, which will introduce any new type of accident outside those
assumed in the UFSAR.
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.
Implementation of this amendment will not involve a significant
reduction in any margin of safety. Margin of safety is related to
the confidence in the ability of the fission product barriers to
perform their design functions during and following an accident
situation. These barriers include the fuel cladding, the reactor
coolant system, and the containment system. The performance of these
fission product barriers will not be impacted by implementation of
this proposed TS amendment. During single pond return header
operation, the NSWS and its supported systems will remain capable of
performing their required functions. No safety margins will be
impacted.
The PRA analysis conducted for this proposed amendment
demonstrated that the impact on overall plant risk remains
acceptable during single pond return header operation. Therefore,
there is not a significant reduction in the margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, Duke Energy concludes that the proposed
amendment does not involve a significant hazards consideration under
the standards set forth in 10 CFR 50.92, and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: 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 Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina
Date of amendment request: October 3, 2017. A publicly-available
version is in ADAMS under Accession No. ML17277A855.
Description of amendment request: The amendments would revise
Surveillance Requirement (SR) 3.8.4.5 contained in Technical
Specification (TS) 3.8.4, ``DC [Direct Current] Sources--Operating.''
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 to the battery charger amperage requirements
of SR 3.8.4.5 contained in TS 3.8.4 does not impact the physical
function of plant structures, systems, or components (SSC) or the
manner in which SSCs perform their design function. The proposed
change does not authorize the addition of any new plant equipment or
systems, nor does it alter the assumptions of any accident analyses.
The DC electrical power system, including the battery chargers, is
not an initiator of any accident sequence analyzed in the Updated
Final Safety Analysis Report. Rather, the DC electrical power system
supports operation of equipment used to mitigate accidents.
Specifically, the purpose of the battery chargers is to continuously
maintain their respective battery in a charged standby condition
while providing power to the system loads. The proposed change does
not adversely affect accident initiators or precursors, nor does it
alter the design assumptions, conditions, and configuration or the
manner in which the plant is operated and maintained.
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 to the battery charger amperage requirements
of SR 3.8.4.5 contained in TS 3.8.4 does not require any
modification to the plant or change equipment operation. The
proposed change will not introduce failure modes that could result
in a new accident, and the change does not alter assumptions made in
the safety analysis. Performance of battery testing is not a
precursor to any accident previously evaluated. The proposed change
will not alter the design configuration, or method of operation of
plant equipment beyond its normal functional capabilities. The
proposed change does not create any new credible failure mechanisms,
malfunctions, or accident initiators.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from those that have been
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to the battery charger amperage requirements
of SR 3.8.4.5 contained in TS 3.8.4 does not alter or exceed a
design basis or safety limit. There is no change being made to
safety analysis assumptions or the safety limits that would
adversely affect plant safety as a result of the proposed change.
Margins of safety are unaffected by the proposed change and the
applicable requirements of 10 CFR 50.36(c)(2)(ii) and 10 CFR 50,
Appendix A will continue to be met.
Therefore, the proposed change does not involve any reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel,
550 South Tryon Street, M/C DEC45A, Charlotte, NC 28202.
NRC Branch Chief: Undine Shoop.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: December 12, 2017. A publicly-available
version is in ADAMS under Accession No. ML17346B280.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) 3.6.4.1, ``Secondary
Containment,'' Surveillance Requirement (SR) 3.6.4.1.1. The proposed
changes are based on Technical Specifications Task Force (TSTF)
Traveler TSTF-551, Revision 3, ``Revise Secondary Containment
Surveillance Requirements'' (ADAMS Accession No. ML16277A226).
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 SRs are 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
[[Page 10916]]
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. 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 analyzed?
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 SRs are 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. 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 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 A. Horin, Esq., Winston & Strawn,
1700 K Street NW, Washington, DC 20006-3817.
NRC Branch Chief: Robert J. Pascarelli.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: December 18, 2017. A publicly-available
version is in ADAMS under Accession No. ML17352B255.
Description of amendment request: The amendment would revise the
Environmental Protection Plan to incorporate the terms and conditions
of the Incidental Take Statement included in the Biological Opinion
issued to Energy Northwest by the National Marine Fisheries Service.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The changes are administrative in nature and would in no way
affect the initial conditions, assumptions, or conclusions of
Columbia's accident analyses. In addition, the proposed changes
would not affect the operation or performance of any equipment
assumed in the accident analyses.
Therefore there is no 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 analyzed?
Response: No.
The changes are administrative in nature and would in no way
impact or alter the configuration or operation of the facility and
would create no new modes of operation.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The changes are administrative in nature and would in no way
affect plant or equipment operation or the accident analysis.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street NW, Washington, DC 20006-3817.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Nuclear Indian Point 2, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2
(IP2), Westchester County, New York
Date of amendment request: December 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17354A007.
Description of amendment request: The amendment would revise
Technical Specification (TS) Limiting Condition for Operation 3.7.13,
``Spent Fuel Pit Storage,'' and TS 4.3, ``Fuel Storage.'' Specifically,
the proposed changes would (1) resolve a nonconservative TS associated
with TS Limiting Condition for Operation 3.7.13, (2) negate the need
for the associated compensatory measures, and (3) remove credit for the
installed Boraflex panels as a neutron absorber in the criticality
analysis of record. The proposed changes in the criticality analysis of
record would instead credit empty cells, rod cluster control assemblies
(RCCAs), and neutron leakage along the outer two storage rows of the
spent fuel pit (SFP).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment was evaluated for impact on the following
previously evaluated events and accidents:
--Multiple Misloads
--Misplaced Assembly
--Dropped Assembly
--Misloaded Assembly
--Over Temperature
--Seismic
--Boron Dilution
--Fuel Handling Accident
--Loss of Spent Fuel Pool Cooling
Multiple Misloads, Misplaced Assembly, Dropped Assembly, Misloaded
Accidents
Operation in accordance with the proposed Technical
Specifications will not significantly increase the probability of
multiple misloads, misplaced assembly, dropped assembly and
misloaded assembly accidents because:
a. There are no changes to the equipment for fuel handling or
how fuel assemblies are handled, including how fuel assemblies are
inserted into and removed from SFP storage locations. There are no
changes to how RCCAs will be handled, including how
[[Page 10917]]
RCCAs are inserted into, or removed from, a fuel assembly.
b. The processes and procedures that are currently in place are
sufficiently robust. The proposed Technical Specifications utilize
the same basic fuel assembly classification and storage location
concepts as those currently in place. However, they do represent a
minimal increase in complexity:
--The current TS for fuel storage are complex because the Boraflex
neutron absorber built into the SFP racks has degraded. To address
this degradation the SFP is divided into four irregularly shaped
Regions (Region 1-1, Region 1-2, Region 2-1, and Region 2-2). In
addition to the four regions there are six special locations known
as peripheral locations in Region 2-2 which are treated as suitable
for storage of fuel otherwise designated for Region 1-1 or 1-2.
These regions are graphically depicted in the current TS Figure
3.7.13-5.
Each one of these regions has its own rules for fuel placement which
are identified in the TS.
--The current Technical Specifications determine a minimum required
burnup for each fuel assembly based on initial enrichment, burnup,
and cooling time with individual fuel assembly storage location
within the SFP restricted based on this minimum required burnup. The
minimum required burnup is determined for each of the four regions
(1-1, 1-2, 2-1, and 2-2) that utilize a total of ten curves. The
proposed assembly categorization is slightly more complex due to the
following:
[cir] The minimum required burnup is dependent on the averaged
assembly peaking factor in addition to the initial enrichment,
burnup, and cooling time.
[cir] the minimum required burnup is used to determine the
reactivity category of each fuel assembly.
[cir] the minimum required burnup is adjusted, as necessary, to
account for hafnium inserts, a reconstituted fuel assembly with
missing stainless steel replacement rods, and a maximum burnup
average boron concentration in excess of 950 ppm [parts per
million].
--The current Technical Specifications restrict acceptable SFP
storage locations to Regions 1-1, 1-2, 2-1 and 2-2 based on minimum
required burnup. The proposed Technical Specifications are minimally
more complex due to the following:
[cir] Acceptable storage locations are defined by fuel assembly
category and a base configuration is specified. There are five
reactivity categories. Certain cell locations in Region 2 require
that Category 5 fuel assemblies contain a full length RCCA.
[cir] the base configurations in Region 1 and Region 2 may be
changed in accordance with certain well-defined criteria. An example
of a change to a base configuration is that a checkerboard area may
be formed in Region 2 where all four sides of the checkerboard are
rows of empty cells.
The minimal increase in complexity of current and future fuel
categorization and SFP storage restrictions is offset by the
significant number of fuel assemblies that have been pre-categorized
in TS Tables 3.7.13-2 and Table 3.7.13-3. The minimal increase is
also offset by the use of two curves to determine the minimum
required burnup (instead of the 10 currently used).
Operation in accordance with the proposed TS will not
significantly increase the consequences of multiple misloads,
misplaced assembly, dropped assembly and misloaded assembly
criticality accidents because the proposed CSA [criticality safety
analysis] demonstrates that the acceptance criteria continue to be
met for each of these accidents.
Over Temperature Accident
Operation in accordance with the proposed TS will not
significantly increase the probability of an over temperature
accident because the proposed change does not alter the manner in
which the IP2 spent fuel cooling loop is designed, operated, or
maintained.
Operation in accordance with the proposed TS will not
significantly increase the consequences of an over temperature
accident because the proposed CSA demonstrates that the acceptance
criteria continue to be met for this accident.
Seismic Event
Operation in accordance with the proposed TS will not
significantly increase the probability of a seismic event because
there are no elements of the proposed changes that influence the
occurrence of any natural event.
Operation in accordance with the proposed TS will not
significantly increase the consequences of a seismic event because
the proposed changes do not significantly alter the physical
arrangement of the spent fuel racks and do not increase the
allowable number of fuel assemblies to be stored in the pool. The
proposed TS changes require two cell blockers to be in place. These
cell blockers have been evaluated and they have a negligible effect
on the seismic response of the SFP racks. In addition, the proposed
TS changes allow for the placement of miscellaneous non-actinide
materials, for example, empty or full trash baskets in fuel
positions of any category, in Water Holes and in 50% Water Holes.
The placement of miscellaneous materials in the identified locations
has been evaluated and has a negligible effect on the seismic
response of the SFP racks.
Boron Dilution Accident
Operation in accordance with the proposed TS will not
significantly increase the probability of a boron dilution event
because the proposed change does not alter the manner in which the
IP2 spent fuel cooling system or any other plant system is designed,
operated, or maintained, or otherwise increase the likelihood of
adding significant quantities of unborated water into the spent fuel
pit.
Operation in accordance with the proposed TS will not
significantly increase the consequences of a boron dilution event
because the TS minimum soluble boron concentration remains unchanged
at 2000 ppm and the boron concentration required to ensure
keff less than or equal to 0.95 has been evaluated at 700
ppm. The proposed CSA demonstrates that the acceptance criteria
continue to be met for this accident.
Fuel Handling Accident
Operation in accordance with the proposed TS will not
significantly increase the probability of a[n] FHA [fuel handling
accident] because the individual fuel assemblies will be moved using
the same equipment, procedures, and other administrative controls
(i.e. fuel move sheets) that are currently used.
Operation in accordance with the proposed TS will not
significantly increase the consequences of a[n] FHA because the
radiological source term of a single fuel assembly will remain the
same.
Loss of Spent Fuel Pool Cooling
Operation in accordance with the proposed TS will not
significantly increase the probability of a loss of spent fuel pit
cooling because the proposed change does not alter the manner in
which the IP2 spent fuel cooling loop is designed, operated, or
maintained.
Operation in accordance with the proposed TS will not
significantly increase the consequences of a loss of spent fuel pit
cooling because the proposed change credits empty cells whereas the
thermal design basis for the spent fuel pit cooling loop provides
for all fuel pit rack locations to be filled at the end of a full
core discharge. The proposed TS changes require two cell blockers to
be in place. These cell blockers have been evaluated and they have a
negligible effect on the thermal response to a loss of spent fuel
pool cooling. In addition, the proposed TS changes allow for the
placement of miscellaneous non-actinide materials, for example,
empty or full trash baskets in fuel positions of any category, in
Water Holes and in 50% Water Holes. The placement of miscellaneous
materials in the identified locations has been evaluated and has a
negligible effect on the thermal response to a loss of spent fuel
pool cooling.
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.
Operation in accordance with the proposed TS do not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new modes of operation are introduced by
the proposed changes. The proposed changes will not create any
failure mode not bounded by previously evaluated accidents.
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.
Operation in accordance with the proposed TS does not involve a
significant reduction in a margin of safety.
[[Page 10918]]
The margin of safety required by 10 CFR 50.68(b)(4) remains
unchanged. The evaluations in the CSA confirm that operation in
accordance with the proposed amendment continues to meet the
required subcriticality margins for both normal operations and
accident conditions. In addition, the SFP seismic and thermal
margins are essentially unchanged.
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: Bill Glew, Associate General Counsel,
Entergy Services, Inc., 639 Loyola Avenue, 22nd Floor, New Orleans, LA
70113.
NRC Branch Chief: James G. Danna.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3 (IP3), Westchester County, New York
Date of amendment request: December 8, 2017. A publicly-available
version is in ADAMS under Accession No. ML17349A131.
Description of amendment request: The amendment would allow for a
one-time extension to the 15-year frequency of the IP3 containment
leakage rate test (i.e., Integrated Leakage Rate Test (ILRT) or Type A
test). Specifically, Technical Specification 5.5.15, ``Containment
Leakage Rate Testing Program,'' would be revised to allow the existing
ILRT frequency to be extended one time from 15 to 16 years. The next
required ILRT test would be performed no later than the plant restart
after the spring 2021 (3R21) refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment involves changes to the IP3 containment
leakage rate testing program. The proposed amendment does not
involve a physical change to the plant or a change in the manner in
which the plant is operated or controlled. The primary containment
function is to provide an essentially leak tight barrier against the
uncontrolled release of radioactivity to the environment for
postulated accidents. As such, the containment itself and the
testing requirements to periodically demonstrate the integrity of
the containment exist to ensure the plant's ability to mitigate the
consequences of an accident do not involve any accident precursors
or initiators. Therefore, the probability of occurrence of an
accident previously evaluated is not significantly increased by the
proposed amendment.
The proposed amendment adopts the NRC accepted guidelines of NEI
[Nuclear Energy Institute] 94-01, Revision 3-A, for development of
the IP3 performance-based testing program for the Type A testing.
Implementation of these guidelines continues to provide adequate
assurance that during design basis accidents, the primary
containment and its components would limit leakage rates to less
than the values assumed in the plant safety analyses. The potential
consequences of extending the ILRT interval one-time to 16 years
have been evaluated by analyzing the resulting changes in risk. The
increase in risk in terms of person-rem per year within 50 miles
resulting from design basis accidents was estimated to be acceptably
small and determined to be within the guidelines published in the
NRC Final Safety Evaluation for NEI Topical Report (TR) 94-01,
Revision 3-A. Additionally, the proposed change maintains defense-
in-depth by preserving a reasonable balance among prevention of core
damage, prevention of containment failure, and consequence
mitigation. Entergy has determined that the increase in conditional
containment failure probability due to the proposed change would be
very small. Therefore, it is concluded that the proposed amendment
does not significantly increase the 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.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, for the establishment of a one-time only 16-
year interval for the performance of the containment ILRT. The
containment and the testing requirements to periodically demonstrate
the integrity of the containment exist to ensure the plant's ability
to mitigate the consequences of an accident do not involve any
accident precursors or initiators. The proposed change does not
involve a physical change to the plant (i.e., no new or different
type of equipment will be installed) or a change to the manner in
which the plant is operated or controlled. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any [accident] previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, for the establishment of a one-time only 16-
year interval for the performance of the containment ILRT. This
amendment does not alter the manner in which safety limits, limiting
safety system setpoints, or limiting conditions for operation are
determined. The specific requirements and conditions of the
containment leakage rate testing program, as defined in the TS,
ensure that the degree of primary containment structural integrity
and leak-tightness that is considered in the plant's safety analysis
is maintained. The overall containment leakage rate limit specified
by the TS is maintained, and the Type A, Type B, and Type C
containment leakage tests would be performed at the frequencies
established in accordance with the NRC accepted guidelines of NEI
94-01, Revision 3-A. Containment inspections performed in accordance
with other plant programs serve to provide a high degree of
assurance that the containment would not degrade in a manner that is
not detectable by an ILRT. A risk assessment using the current IP3
PSA [probabilistic safety analysis] model concluded that extending
the ILRT test interval one-time from 15 years to 16 years results in
a very small change to the risk profile. 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: Bill Glew, Associate General Counsel,
Entergy Services, Inc., 639 Loyola Avenue, 22nd Floor, New Orleans, LA
70113.
NRC Branch Chief: James G. Danna.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station (CPS), Unit No. 1, DeWitt County, Illinois
Date of amendment request: January 9, 2017. A publicly-available
version is in ADAMS under Accession No. ML18009B037.
Description of amendment request: The proposed change would
incorporate a revised alternative source term dose calculation
resulting from the removal of a reduction factor credit for dual remote
Control Room outside air intakes that had been previously misapplied.
This would modify the loss-of-coolant accident (LOCA) dose calculation
and the subsequent calculation results as described in the CPS Updated
Safety Analysis Report and would revise the affected CPS Technical
Specifications.
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
[[Page 10919]]
consequences of an accident previously evaluated?
Response: No.
The proposed change results in higher Control Room X/Qs
[atmospheric dispersion values] which are equivalent to reduced
atmospheric dispersion. The increased Control Room X/Qs, in turn,
result in higher post-accident Control Room doses. Neither the
higher X/Qs, nor the resultant increase in the Control Room doses
affect any initiator or precursor of any accident previously
evaluated. Therefore, the proposed change does not involve a
significant increase in the probability of an accident previously
evaluated.
The proposed change results in an increase in the post-LOCA
radiological dose to a Control Room occupant. However, the resultant
post-LOCA Control Room dose remains within the regulatory limits of
10 CFR 50.67 [, ``Accident source term''] and 10 CFR 50, Appendix A,
``General Design Criteria for Nuclear Power Plants'' Criterion 19,
``Control Room.'' Therefore, the proposed change does not involve a
significant increase in the consequences of an accident previously
evaluated.
In summary, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the design function of
operation of the Control Room heating, ventilation, and air-
conditioning (HVAC) system, or the ability of this system to perform
its design function. The only change is the removal of the Control
Room dose reduction factor credit taken for providing a dual remote
Control Room air intake. The proposed change does not alter the
safety limits, or safety analysis associated with the operation of
the plant. Accordingly, the change does not introduce any new
accident initiators. Rather, this proposed change is the result of
an evaluation of the Control Room doses following the most limiting
LOCA that can occur at CPS. The proposed change does not introduce
any new modes of plant operation. As a result, no new failure modes
are introduced.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The revised post-LOCA dose consequences to a Control Room
occupant were calculated in accordance with the requirements of 10
CFR 50.67, [Regulatory Guide (RG)] 1.183, [``Alternative
Radiological Source Terms for Evaluating Design Basis Accidents at
Nuclear Power Reactors''] and NRC SRP [Standard Review Plan] Section
15.0.1, ``Radiological Consequences Analyses Using Alternative
Source Terms.''
The margin of safety is considered to be that provided by
meeting the applicable regulatory limits. The increased Control Room
X/Qs result in an increase in Control Room dose following the design
basis LOCA; however, since the Control Room dose following the
design basis accident remains within the regulatory limits, there is
not a significant reduction in a margin of safety.
Therefore, operation of CPS in accordance with the proposed
change will not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review it appears the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: David J. Wrona.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: January 24, 2018. A publicly-available
version is in ADAMS under Accession No. ML18024A275.
Description of amendment request: The proposed amendments would
revise Technical Specifications (TSs) 3.7.2, ``Diesel Generator Cooling
Water (DGCW) System''; 3.8.1, ``AC [Alternating Current] Sources-
Operating''; and the associated TS Bases to allow an extended period to
install isolation valves to support replacing degraded Core Standby
Cooling System (CSCS) piping.
The proposed changes modify TS 3.7.2 to include a 7-day Completion
Time (CT) when one or more required DGCW subsystem(s) are inoperable.
The proposed changes to TS 3.8.1 include a 7-day CT when a Division 2
Diesel Generator (DG) and the required opposite unit Division 2 DG are
inoperable. The proposed changes will only be used during four
refueling outages, two for Unit 1 prior to July 1, 2024, and two for
Unit 2 prior to July 1, 2023. The current planned schedule, subject to
change, is L2R17 (2019), L1R18 (2020), L2R18 (2021), and L1R19 (2022).
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 previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The proposed change
does not have a detrimental impact on the integrity of any plant
structure, system, or component that initiates an analyzed event. No
active or passive failure mechanisms that could lead to an accident
are affected. Non-Code line stops required to provide isolation will
maintain the availability of the online unit's CSCS. The non-Code
line stops being used to isolate the system during the specified
refueling outages are being designed to the same or greater pressure
rating and seismic requirements as the CSCS piping.
Redundancy is provided by designing the CSCS as multiple
independent subsystems. Divisional separation between subsystems
assures that no single failure can affect more than one division's
subsystem. Therefore, assuming a single failure in any division's
subsystem including the subsystem shared between units, two other
divisional subsystems in each unit will remain unaffected. This
ensures adequate redundancy to supply the minimum required cooling
water for safe shutdown of the operating unit or mitigate the
consequences of an accident.
The proposed limited use of increased CT's of the operating
unit's CSCS maintains the design basis assumptions. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
The proposed change involves the temporary installation of new
equipment (mechanical line stops) that will be designed and
installed to the same or greater pressure rating and seismic design
as the CSCS piping. The currently installed equipment will not be
operated in a new or different manner. No new or different system
interactions are created and no new processes are introduced. The
proposed changes will not introduce any new failure mechanisms,
malfunctions, or accident initiators not already considered in the
design and licensing bases.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change does not alter any existing setpoints at
which protective actions are initiated and no new setpoints or
protective actions are introduced. The design and operation of the
CSCS remains unchanged. The proposed change provides a limited
period to restore inoperable DGCW subsystems and Division 2 DGs
instead of interrupting plant operations, possibly requiring an
orderly plant shutdown of the operating unit. The potential to avoid
a plant transient in conjunction with maintaining availability of
the DGCW subsystems and Division 2 DGs offsets any risk associated
with the limited CT. The proposed change
[[Page 10920]]
does not impact a design basis, limiting safety system setting, or
safety limit specified in TSs.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: David J. Wrona.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center (DAEC), Linn County, Iowa
Date of amendment request: December 15, 2017. A publicly-available
version is in ADAMS under Accession No. ML17363A067.
Description of amendment request: The proposed amendment would
revise the Emergency Plan for the DAEC to adopt the Nuclear Energy
lnstitute's (NEl's) revised Emergency Action Level (EAL) scheme
described in NEI 99-01, Revision 6, ``Development of Emergency Action
Levels for Non-Passive Reactors,'' which has been endorsed by the NRC.
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 does not impact the physical configuration
or function of plant structures, systems, or components (SSCs) or
the manner in which SSCs are operated, maintained, modified, tested,
or inspected. No actual facility equipment or accident analyses are
affected by the proposed changes.
The change revises the NextEra Emergency Action Levels to be
consistent with the NRC endorsed EAL scheme contained in NEI 99-01,
Revision 6, ``Methodology for Development of Emergency Action
Levels,'' but does not alter any of the requirements of the
Operating License or the Technical Specifications.
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 involve a physical alteration of
the plant (no new or different type of equipment will be installed).
The proposed change does not create any new failure modes for
existing equipment or any new limiting single failures.
Additionally, the proposed change does not involve a change in the
methods governing normal plant operation, and all safety functions
will continue to perform as previously assumed in the accident
analyses. Thus, the proposed change does not adversely affect the
design function or operation of any structures, systems, and
components important to safety.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed change.
The proposed change does not challenge the performance or integrity
of any safety-related system. 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 margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed change will have no affect
[sic] on the availability, operability, or performance of safety-
related systems and components. The proposed change will not
adversely affect the operation of plant equipment or the function of
equipment assumed in the accident analysis.
The proposed amendment does not involve changes to any safety
analyses assumptions, safety limits, or limiting safety system
settings. The changes do not adversely impact plant operating
margins or the reliability of equipment credited in the safety
analyses.
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.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: December 19, 2017. A publicly-available
version is in ADAMS under Accession No. ML17353A928.
Description of amendment request: The proposed amendment would
separate the Linear Heat Generation Rate (LHGR) requirements and
actions from the Average Planar Linear Heat Generation Rate (APLHGR)
requirements and actions contained in Technical Specification (TS)
3.2.1. The proposed amendment adds new TS 3.2.3, ``Linear Heat
Generation Rate (LHGR),'' and modifies TS 1.1, ``Definitions,'' TS
3.4.1, ``Recirculation Loops Operating,'' and TS 5.6.5, ``Core
Operating Limits Report (COLR),'' to reflect the LHGR change.
Modifications associated with TS 3.2.1 and the new TS 3.2.3 are also
being added to the actions for TS 3.3.4.1, ``End of Cycle Recirculation
Pump Trip (EOC-RPT) Instrumentation,'' and TS 3.7.7, ``The Main Turbine
Bypass System.''
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 separation of the LHGR requirements and actions from the
APLHGR TS is an administrative change. No actions within the TS are
changed. The addition of the LCO [limiting condition for operation]
for APLHGR and the proposed LCO for LHGR to the LCO for 3.3.4.1, End
of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation and the
LCO for TS 3.7.7, Main Turbine Bypass System reflect within the TS
requirements APLHGR and LHGR actions which are already occurring via
the core monitoring processes in place. None of those changes affect
any plant systems, structures, or components designed for the
prevention or mitigation of previously evaluated accidents. No new
equipment is added nor is installed equipment being changed or
operated in a different manner.
LHGR limits have been defined to provide sufficient margin
between the steady-state operating condition and any fuel damage
condition to accommodate uncertainties and to assure that no fuel
damage results even during the worst anticipated transient condition
at any time.
The proposed change does not modify the limits, change
assumptions for the accident analysis, or change operation of the
station. Therefore, the proposed change does not involve an increase
in the probability or consequences of a previously evaluated
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The separation of the LHGR requirements and actions from the
APLHGR TS is an administrative change. No actions within the TS are
changed. The addition of the LCO for APLHGR and the proposed LCO for
LHGR to the LCO for 3.3.4.1, End of Cycle
[[Page 10921]]
Recirculation Pump Trip (EOC-RPT) Instrumentation and the LCO for TS
3.7.7, Main Turbine Bypass System reflect within the TS requirements
APLHGR and LHGR actions which are already occurring via the core
monitoring processes in place. None of those changes affect any
plant systems, structures, or components designed for the prevention
or mitigation of previously evaluated accidents. No new equipment is
added nor is installed equipment being changed or operated in a
different manner.
The proposed change does not modify the limits, change
assumptions for the accident analysis, or change operation of the
station. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is not affected by the separation of the
LHGR requirements and actions from the APLHGR TS. Similarly, the
margin of safety is not affected by the addition of the LCO for
APLHGR and the proposed LCO for LHGR to the LCO for 3.3.4.1, End of
Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation and the LCO
for TS 3.7.7, Main Turbine Bypass System.
Appropriate measures exist to control the values of these limits
since it is required by TS that only NRC-approved methods be used to
determine the limits. The proposed change continues to require
operation within the core thermal limits as obtained from NRC-
approved reload design methodologies and the actions to be taken if
a limit is exceeded remain unchanged, again, in accordance with
existing TS.
The proposed change does not modify the limits, change
assumptions for the accident analysis, or change operation of the
station. Therefore, the proposed change has no impact to 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 Blair, P.O. Box 14000, Juno Beach,
FL 33408-0420.
NRC Branch Chief: David J. Wrona.
PSEG Nuclear LLC and Exelon Generation Company, LLC, Docket Nos. 50-272
and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: December 18, 2017, as supplemented by
letter dated February 9, 2018. Publicly-available versions are in ADAMS
under Accession Nos. ML17352A502 and ML18040A319, respectively.
Description of amendment request: The amendments would revise
Technical Specification (TS) 3/4.3.1, ``Reactor Trip System
Instrumentation,'' and TS 3/4.3.2, ``Engineered Safety Feature
Actuation System Instrumentation,'' to increase the completion times
and bypass test times at Salem Nuclear Generating Station, Unit Nos. 1
and 2. The proposed changes are consistent with NRC-approved Technical
Specifications Task Force (TSTF) Travelers TSTF-411, Revision 1,
``Surveillance Test Interval Extensions for Components of the Reactor
Protection System (WCAP-15376-P),'' and TSTF-418, Revision 2, ``RPS
[Reactor Protection System] and ESFAS [Engineered Safety Feature
Actuation System] Test Times and Completion Times (WCAP-14333),'' or
are supported by plant-specific analysis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the completion times and bypass test
time reduce the potential for inadvertent reactor trips and spurious
actuations, and therefore do not increase the probability of any
accident previously evaluated. The proposed changes to the
completion times and bypass test time do not change the response of
the plant to any accidents and have an insignificant impact on the
reliability of the reactor trip system and engineered safety feature
actuation system (RTS and ESFAS) signals. The RTS and ESFAS will
remain highly reliable and the proposed changes will not result in a
significant increase in the risk of plant operation. This is
demonstrated by showing that the impact on plant safety as measured
by core damage frequency (CDF) is less than 1.0E-06 per year and the
impact on large early release frequency (LERF) is less than 1.0E-07
per year. In addition, for the completion time change, the
incremental conditional core damage probabilities (ICCDP) and
incremental conditional large early release probabilities (ICLERP)
are less than 5.0E-7 and 5.0E-08, respectively. These changes meet
the acceptance criteria in Regulatory Guides 1.174 and 1.177.
Therefore, since the RTS and ESFAS will continue to perform their
functions with high reliability as originally assumed, and the
increase in risk as measured by CDF, LERF, ICCDP, ICLERP is within
the acceptance criteria of existing regulatory guidance, there will
not be a significant increase in the consequences of any accidents.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems, and components (SSCs)
from performing their intended function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
The proposed changes are consistent with safety analysis assumptions
and resultant consequences.
Therefore, this change does not significantly increase the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not result in a change in the manner in
which the RTS and ESFAS provide plant protection. The RTS and ESFAS
will continue to have the same setpoints after the proposed changes
are implemented. There are no design changes associated with the
license amendment. The changes to completion times and bypass test
time do not change any existing accident scenarios, nor create any
new or different accident scenarios.
The proposed changes do not involve a modification to the
physical configuration of the plant or changes in the methods
governing normal plant operation. The proposed changes will not
impose any new or different requirement or introduce a new accident
initiator, accident precursor, or malfunction mechanism.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by these changes. Redundant RTS and ESFAS trains
are maintained, and diversity with regard to the signals that
provide reactor trip and engineered safety features actuation is
also maintained. All signals credited as primary or secondary, and
all operator actions credited in the accident analyses will remain
the same. The proposed changes will not result in plant operation in
a configuration outside the design basis. The calculated impact on
risk is insignificant and meets the acceptance criteria contained in
Regulatory Guides 1.174 and 1.177.
Therefore, since the proposed changes do not impact the response
of the plant to a design basis accident, 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
[[Page 10922]]
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: James G. Danna.
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: December 20, 2017. A publicly-available
version is in ADAMS under Accession No. ML17354A964.
Description of amendment request: The requested amendment proposes
changes to Combined License Appendix C (and to plant-specific Tier 1
information) and associated Tier 2 information to allow a pneumatic
test to be used in lieu of a hydrostatic test for the Main Control Room
Emergency Habitability System (VES) consistent with American Society of
Mechanical Engineers (ASME) Section III.
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 allow for pneumatic testing of the VES ASME
Section III components and piping. ASME Section III, ND-6000
contains the requirements for pressure testing of piping and
components. ASME Section III, ND-6112.1(a) allows for a pneumatic
test to be used in lieu of a hydrostatic test when components,
appurtenances or systems cannot be readily dried and traces of the
testing medium cannot be tolerated. Due to the design and layout of
the VES, it may be difficult to dry the system following a
hydrostatic test. Traces of water could result in sending a slug of
water through the system or rust to form. Allowing for pneumatic
testing continues to meet the ASME Section III code. The proposed
changes do not affect the operation of the VES. The VES maintains
its design function to maintain control room habitability.
The proposed changes do not affect the operation of any systems
or equipment that initiate an analyzed accident or alter any
structures, systems, and components (SSCs) accident initiator or
initiating sequence of events. Therefore, the probabilities of
accidents previously evaluated are not affected.
The proposed changes do not affect the prevention and mitigation
of other abnormal events (e.g., anticipated operational occurrences,
earthquakes, floods and turbine missiles), or their safety or design
analyses. Therefore, the consequences of the accidents evaluated in
the Updated Final Safety Analysis Report are not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not affect the operation of any systems
or equipment that may initiate a new or different kind of accident,
or alter any SSC such that a new accident initiator or initiating
sequence of events is created.
The proposed changes do not 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.
Therefore, the requested 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 allow for pneumatic testing of the VES ASME
Section III components and piping. The VES ASME Section III
components and piping continue to meet the ASME Section III code.
The proposed changes do not have any effect on the ability of the
safety-related SSCs to perform their design basis functions. The
proposed changes do not affect the ability of the VES to maintain
control room habitability.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed changes, and no margin of
safety is reduced. Therefore, the requested amendment does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and based on
this review it appears that the three standards of 10 CFR 50.92 (c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazard 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: January 31, 2018. A publicly-available
version is in ADAMS under Accession No. ML18031B131.
Description of amendment request: The requested amendment proposes
changes to the Technical Specification (TS) 3.4.6, Pressurizer Safety
Valve, Applicability to require the pressurizer safety valves (PSVs) to
be operable when the TS 3.4.14, Low Temperature Overpressure Protection
(LTOP), is not required to be operable. A conforming change to the TS
3.4.6 Actions is also proposed. Additional TS changes necessary to
support PSVs operability are proposed for consistency with the TS 3.4.6
change.
The request also proposes moving TS Limiting Condition for
Operation Notes regarding reactor coolant pump starts from TS 3.4.4,
Reactor Coolant System (RCS) Loops, 3.4.8, Minimum RCS Flow, and
3.4.14, Low Temperature Overpressure Protection (LTOP), to TS 3.4.3,
RCS Pressure/Temperature (P/T) Limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not affect the operation of any systems
or equipment that initiate an analyzed accident or alter any
structures, systems, and components (SSCs) accident initiator or
initiating sequence of events.
The proposed changes do not affect the physical design of SSCs
related to the TS on Engineered Safety Features Actuation System
(ESFAS), RCS P/T limits, RCS loops, RCS flow, pressurizer, PSVs,
LTOP, or Reactor Vessel head vent (RVHV), as described in the
Updated Final Safety Analysis Report (UFSAR). Therefore, the
operation of the listed functions and components is not affected.
Therefore, the proposed changes do not affect the probability of an
accident previously evaluated.
The proposed changes do not affect the physical design of SSCs
related to the TS on ESFAS, RCS P/T limits, RCS loops, RCS flow,
pressurizer, PSVs, LTOP, or RVHV to meet their design functions. The
design of the functions and components continue to meet the same
regulatory acceptance criteria, codes, and standards as stated in
the UFSAR. In addition, the proposed changes maintain the
capabilities of the ESFAS, RCS P/T
[[Page 10923]]
limits, RCS loops, RCS flow, pressurizer, PSVs, LTOP, or RVHV to
mitigate the consequences of an accident and to meet the applicable
regulatory acceptance criteria.
The proposed changes do not affect the prevention and mitigation
of other abnormal events (e.g., anticipated operational occurrences,
earthquakes, floods, and turbine missiles), or their safety or
design analyses. Therefore, the consequences of the accidents
evaluated in the UFSAR are not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not affect the operation of any systems
or equipment that may initiate a new or different kind of accident,
or alter any SSC such that a new accident initiator or initiating
sequence of events is created.
The proposed changes do not 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.
Therefore, the requested amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes maintain existing safety margins. The
proposed changes verify and maintain the physical design of SSCs
related to ESFAS, RCS P/T limits, RCS loops, RCS flow, pressurizer,
PSVs, LTOP, and RVHV 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.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed changes, and no margin of
safety is reduced. Therefore, the requested amendment does not
involve a significant reduction in 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 hazard 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 (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: February 2, 2018. A publicly-available
version is in ADAMS under Accession No. ML18037B114.
Description of amendment request: The requested amendment proposes
departures from the generic AP1000 Design Control Document (DCD) for
the plant-specific VEGP Combined License (COL) Appendix A Technical
Specifications (TS) and related departures from generic DCD Tier 2
information in the Updated Final Safety Analysis Report (UFSAR) (which
includes the plant-specific DCD Tier 2 information). Specifically, the
proposed changes would make administrative changes to COL Appendix A,
TS 5.6.3, for the core operating limits report required documentation
to include analytical methods which are described elsewhere in the TS
and in the UFSAR, and make an editorial change to COL Appendix A TS
5.7.2 for high radiation areas to correct a typographical error.
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 administrative and editorial changes
consistent with the requirements described elsewhere in the TS and
in the UFSAR, and do not adversely affect the operation of any
systems or equipment that initiate an analyzed accident or alter any
structures, systems, and components (SSCs) accident initiator or
initiating sequence of events. The proposed changes to the
analytical methods approved for maintaining core operating limits do
not result in any increase in probability of an analyzed accident
occurring, and prevent power oscillations and maintain the initial
conditions and operating limits required by the accident analysis,
and the analyses of normal operation and anticipated operational
occurrences, so that fuel design limits are not exceeded for events
resulting in positive reactivity insertion and reactivity feedback
effects, and so that the consequences of postulated accidents are
not changed. The proposed changes do not adversely affect the
ability of the automatic reactor trips to perform the required
safety function to trip the reactor when necessary to protect fuel
design limits, and do not adversely affect the probability of
inadvertent operation or failure of the automatic reactor trips.
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 administrative and editorial changes
consistent with the requirements described elsewhere in the TS and
in the UFSAR, and 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 to the
analytical methods approved for maintaining core operating limits do
not result in any increase in probability of an analyzed accident
occurring, and prevent power oscillations and maintain the initial
conditions and operating limits required by the accident analysis,
and the analyses of normal operation and anticipated operational
occurrences, so that fuel design limits are not exceeded for events
resulting in positive reactivity insertion and reactivity feedback
effects, and so that the consequences of postulated accidents are
not changed. The proposed changes do not adversely affect the
ability of the automatic reactor trips to perform the required
safety function to trip the reactor when necessary to protect fuel
design limits, and do not adversely affect the probability of
inadvertent operation or failure of the automatic reactor trips.
These proposed changes do not adversely affect any other SSC
design functions or methods of operation in a manner that results in
a new failure mode, malfunction, or sequence of events that affect
safety-related or nonsafety-related equipment. Therefore, this
activity does not allow for a new fission product release path,
result in a new fission product barrier failure mode, or create a
new sequence of events that results in significant fuel cladding
failures.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes are administrative and editorial changes
consistent with the requirements described elsewhere in the TS and
in the UFSAR, and maintain existing safety margins through continued
application of the existing requirements of the UFSAR. The proposed
changes maintain the initial conditions and operating limits
required by the accident analysis, and the analyses of normal
operation and anticipated operational occurrences, so that the
existing fuel design limits specified in the UFSAR are not exceeded
for events resulting in positive reactivity insertion and reactivity
feedback effects, and so that the consequences of postulated
accidents are not changed. Therefore, the proposed changes satisfy
the same safety functions in accordance with the
[[Page 10924]]
same requirements 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.
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 hazard consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(Watts Bar), Units 1 and 2, Rhea County, Tennessee
Date of amendment request: January 5, 2018. A publicly-available
version is in ADAMS under Accession No. ML18008A257.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.6.3, ``Containment Isolation Valves,''
and Surveillance Requirement 3.6.3.5 to change the frequency in
accordance with the Watts Bar Containment Leakage Rate Testing Program,
which is described in TS 5.7.2.19. The proposed change would allow leak
rate testing of the containment purge system containment isolation
valves to be performed at least once every 30 months, as prescribed in
Regulatory Guide 1.163.
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 deletes the augmented testing requirement
for these containment isolation valves and allows the surveillance
intervals to be set in accordance with the Containment Leakage Rate
Testing Program. This change does not affect the system function or
design. The purge valves are not an initiator of any previously
analyzed accident. Leakage rates do not affect the probability of
the occurrence of any accident. Operating history has demonstrated
that the valves do not degrade and cause leakage as previously
anticipated. Because these valves have been demonstrated to be
reliable, these valves can be expected to perform the containment
isolation function as assumed in the accident analyses. The proposed
changes do not affect the source term, containment isolation, or
radiological release assumptions used in evaluating the radiological
consequences of an accident previously evaluated. Further, the
proposed changes do not increase the types or amounts of radioactive
effluent that may be released offsite, nor significantly increase
individual or cumulative occupational/public radiation exposures.
The proposed changes do not significantly increase the probability
of an accident and are consistent with safety analysis assumptions
and resultant consequences.
Therefore, the changes do not increase the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This change does not involve a physical alteration to the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing any normal plant operation. The
change does not alter assumptions made in the safety analyses or
licensing basis. Extending the test intervals has no influence on,
nor does it contribute in any way to, the possibility of a new or
different kind of accident or malfunction from those previously
analyzed. No change has been made to the design, function, or method
of performing leakage testing. Leakage acceptance criteria have not
changed. No new accident modes are created by extending the testing
intervals. No safety-related equipment or safety functions are
altered as a result of this change.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The only margin of safety that has the potential of being
impacted by the proposed change involves the offsite dose
consequences of postulated accidents, which are directly related to
the containment leakage rate. The proposed change does not alter the
method of performing the tests nor does it change the leakage
acceptance criteria. Sufficient data has been collected to
demonstrate these resilient seals do not degrade at an accelerated
rate. Because of this demonstrated reliability, this change will
provide sufficient surveillance to determine an increase in the
unfiltered leakage prior to the leakage exceeding that assumed in
the accident analysis.
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.
Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant,
Unit 2, Rhea County, Tennessee
Date of amendment request: October 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17284A452.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.3.1, Table 3.3.1-1, ``Reactor Trip
System (RPS) Instrumentation,'' to increase the values for the nominal
trip setpoint and the allowable value for Function 14.a, ``Turbine
Trip--Low Fluid Oil Pressure.'' The proposed changes are due to the
planned replacement and relocation of the pressure switches from the
low pressure auto-stop trip fluid oil header to the high pressure
turbine electrohydraulic control (EHC) oil header. The changes are
needed due to the higher EHC system operating pressure.
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 reflects a design change to the turbine
control system that results in the use of an increased control oil
pressure system, necessitating a change to the value at which a low
fluid oil pressure initiates a reactor trip on turbine trip. The low
fluid oil pressure is an input to the reactor trip instrumentation
in response to a turbine trip event. The value at which the low
fluid oil initiates a reactor trip is not an accident initiator. A
change in the nominal control oil pressure does not introduce any
mechanisms that would increase the probability of an accident
previously analyzed. The reactor trip on turbine trip function is
initiated by the same protective signal as used for the existing
auto stop low fluid oil system trip signal. There is no change in
form or function of this signal and the probability or consequences
of previously analyzed accidents are not impacted.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
[[Page 10925]]
Response: No.
The EHC fluid oil pressure rapidly decreases in response to a
turbine trip signal. The value at which the low fluid oil pressure
switches initiates a reactor trip is not an accident initiator. The
proposed TS change reflects the higher pressure that will be sensed
after the pressure switches are relocated from the auto stop low
fluid oil system to the EHC high pressure header. Failure of the new
switches would not result in a different outcome than is considered
in the current design basis. Further, the change does not alter
assumptions made in the safety analysis but ensures that the
instruments perform as assumed in the accident analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Response: No.
The change involves a parameter that initiates an anticipatory
reactor trip following a turbine trip. The safety analyses do not
credit this anticipatory trip for reactor core protection. The
original pressure switch configuration and the new pressure switch
configuration both generate the same reactor trip signal. The
difference is that the initiation of the trip will now be adjusted
to a different system of higher pressure. This system function of
sensing and transmitting a reactor trip signal on turbine trip
remains the same. There is no impact to safety analysis acceptance
criteria as described in the plant licensing basis because no change
is made to the accident analysis assumptions.
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. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: September 14, 2017, as supplemented by
letter dated December 12, 2017.
Brief description of amendments: The amendments modified Technical
Specifications (TSs) to allow temporary changes to TSs 3.5.2,
``Emergency Core Cooling Systems (ECCS)--Operating''; 3.6.6,
``Containment Spray System''; 3.7.5, ``Auxiliary Feedwater (AFW)
System''; 3.7.6, ``Component Cooling Water (CCW) System''; 3.7.7,
``Nuclear Service Water System (NSWS)''; 3.7.9, ``Control Room Area
Ventilation System (CRAVS)''; 3.7.11, ``Auxiliary Building Filtered
Ventilation Exhaust System (ABFVES)''; and 3.8.1, ``[Alternating
Current] Sources--Operating,'' to permit the ``A'' Train NSWS to be
inoperable for a total of 14 days to address a non-conforming condition
on the ``A'' Train supply piping from the Standby Nuclear Service Water
Pond.
Date of issuance: February 15, 2018.
Effective date: These license amendments are effective as of its
date of issuance and shall be implemented within 120 days of issuance.
Amendment Nos.: 308 (Unit 1) and 287 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML18030A682; 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 Licenses and TSs.
Date of initial notice in Federal Register: December 19, 2017 (82
FR 60226).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 15, 2018.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: March 27, 2017.
Brief description of amendment: The amendment revised the Technical
Specification (TS) requirements in order to address Generic Letter
2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay
Heat Removal, and Containment Spray Systems,'' dated January 11, 2008,
as described in Technical Specifications Task Force (TSTF) Traveler
TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas
Accumulation.''
Date of issuance: February 16, 2018.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 246. A publicly-available version is in ADAMS under
Accession No. ML18025A213; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-21: The amendment
revised the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: June 6, 2017 (82 FR
26132).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 16, 2018.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
1 (ANO-1), Pope County, Arkansas
Date of amendment request: August 14, 2017.
Brief description of amendment: The amendment revised the ANO-1
Technical Specification (TS)
[[Page 10926]]
requirements for unavailable barriers by adding Limiting Condition for
Operation (LCO) 3.0.9, which allows a delay time for entering a
supported system TS when the inoperability is solely due to an
unavailable barrier. The change is consistent with Technical
Specification Task Force (TSTF)-427, Revision 2, ``Allowance for Non
Technical Specification Barrier Degradation Supported System
OPERABILITY.'' In addition, the amendment corrected a typographical
omission on TS page 3.0-3, which was editorial in nature.
Date of issuance: February 26, 2018.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 259. A publicly-available version is in ADAMS under
Accession No. ML18033A175; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: October 24, 2017 (82 FR
49236).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 26, 2018.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2 (ANO-2), Pope County, Arkansas
Date of amendment request: August 14, 2017.
Brief description of amendment: The amendment revised the ANO-2
Technical Specification (TS) requirements for unavailable barriers by
adding Limiting Condition for Operation (LCO) 3.0.9, which allows a
delay time for entering a supported system TS when the inoperability is
solely due to an unavailable barrier. The change is consistent with
Technical Specification Task Force (TSTF)-427, Revision 2, ``Allowance
for Non Technical Specification Barrier Degradation Supported System
OPERABILITY.''
Date of issuance: February 26, 2018.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 309. A publicly-available version is in ADAMS under
Accession No. ML18051A589; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: October 24, 2017 (82 FR
49237).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 26, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant (Calvert Cliffs), Units 1 and 2, Calvert
County, Maryland
Date of amendment request: March 28, 2017.
Brief description of amendments: The amendments revised the Calvert
Cliffs, Units 1 and 2, Technical Specifications (TSs) to change the low
level of the refueling water tank to reflect a needed increase in the
required borated water volume and change the allowable value of the
refueling water tank level-low function.
Date of issuance: February 15, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days of the end of CC1R24 refueling outage for Calvert
Cliffs, Unit 1, and within 60 days of the end of CC2R23 refueling
outage for Calvert Cliffs, Unit 2.
Amendment Nos.: 323 (Unit 1) and 301 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML18029A195; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: June 19, 2017 (82 FR
27887).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 15, 2018.
No significant hazards consideration comments received: No.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of amendment request: June 29, 2017.
Brief description of amendments: The amendments revised the
Technical Specification (TS) requirements for mode change limitations
in TS 3.0.4 and TS 4.0.4 based on Technical Specifications Tasks Force
(TSTF) Improved Standard Technical Specifications Change Traveler,
TSTF-359, Revision 9, ``Increase Flexibility in Mode Restraints''
(ADAMS Accession No. ML031190607).
Date of issuance: February 20, 2018.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 278 (Unit 3) and 273 (Unit 4). A publicly-available
version is in ADAMS under Accession No. ML18018A559; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: September 12, 2017 (82
FR 42850).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 20, 2018.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: February 13, 2017, as supplemented by
letter dated August 11, 2017.
Brief description of amendments: The amendments adopted the NRC-
endorsed Nuclear Energy Institute (NEI) 99-01, Revision 6,
``Development of Emergency Action Levels for Non-Passive Reactors.''
Date of issuance: February 16, 2018.
Effective date: As of the date of issuance and shall be implemented
within a 365-day period after issuance.
Amendment Nos.: Salem--322 (Unit No. 1) and 303 (Unit No. 2); Hope
Creek--210. A publicly-available version is in ADAMS under Accession
No. ML17355A570; documents related to these amendments are listed in
the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-70, DPR-75, and NPF-57:
The amendments revised the emergency action level technical bases
documents.
Date of initial notice in Federal Register: March 28, 2017 (82 FR
15384). The supplemental letter dated
[[Page 10927]]
August 11, 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 February 16, 2018.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: May 26, 2016, as supplemented by letter
dated October 26, 2017.
Brief description of amendments: The amendments correct a non-
conservative Technical Specification (TS) Surveillance Requirement
acceptance criterion for the diesel generator steady-state frequency in
Limiting Condition for Operation 3.8.1, ``AC [Alternating Current]
Sources--Operating.''
Date of issuance: February 12, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 341--Unit 1 and 334--Unit 2. A publicly-available
version is in ADAMS under Accession No. ML18026A810; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-77 and DPR-79.
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: August 2, 2016 (81 FR
50740). The supplemental letter dated October 26, 2017, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 12, 2018.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, on March 6, 2018.
For the Nuclear Regulatory Commission.
Gregory F. Suber,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2018-04827 Filed 3-12-18; 8:45 am]
BILLING CODE 7590-01-P