[Federal Register Volume 84, Number 68 (Tuesday, April 9, 2019)]
[Notices]
[Pages 14142-14156]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2019-06449]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2019-0087]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
-----------------------------------------------------------------------
SUMMARY: Pursuant to 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 March 12, 2019 to March 25, 2019. The last
biweekly notice was published on March 26, 2019.
DATES: Comments must be filed by May 9, 2019. A request for a hearing
must be filed by June 10, 2019.
ADDRESSES: You may submit comments by any of the following methods:
Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2019-0087. Address
questions about NRC dockets IDs in Regulations.gov to Jennifer Borges;
telephone: 301-287-9127; email: [email protected]. For technical
questions, contact the individual(s) listed in the FOR FURTHER
INFORMATION CONTACT section of this document.
Mail comments to: Office of Administration, Mail Stop:
TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, ATTN: Program Management, Announcements and Editing Staff.
For additional direction on obtaining information and
submitting comments, see ``Obtaining Information and Submitting
Comments'' in the SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-5411, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2019-0087, 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-2019-0087.
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 ``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-2019-0087, 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. Background
Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the 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.
III. 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
[[Page 14143]]
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 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
[[Page 14144]]
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 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
application(s), see the application for amendment which is available
for public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: February 8, 2019. A publicly-available
version is in ADAMS under Accession No. ML19039A126.
Description of amendment request: The amendment would adopt TSTF-
564, ``Safety Limit MCPR [minimum critical power ratio],'' Revision 2,
which
[[Page 14145]]
revises the Fermi 2 technical specification safety limit on minimum
critical power ratio (SLMCPR) to reduce the need for cycle-specific
changes to the value while still meeting the regulatory requirement for
a safety limit. In addition, technical specification 5.6.5, ``Core
Operating Limits Report (COLR),'' is revised to require the current
SLMCPR value to be included in the COLR.
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 revises the TS SLMCPR and the list of
core operating limits to be included in the Core Operating Limits
Report (COLR). The SLMCPR is not an initiator of any accident
previously evaluated. The revised safety limit values continue to
ensure for all accidents previously evaluated that the fuel cladding
will be protected from failure due to transition boiling. The
proposed change does not affect plant operation or any procedural or
administrative controls on plant operation that affect the functions
of preventing or mitigating any accidents 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 previously evaluated?
Response: No.
The proposed amendment revises the TS SLMCPR and the list of
core operating limits to be included in the COLR. The proposed
change will not affect the design function or operation of any
structures, systems or components (SSCs). No new equipment will be
installed. As a result, the proposed change will not create any
credible new failure mechanisms, malfunctions, or accident
initiators not 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 amendment revises the TS SLMCPR and the list of
core operating limits to be included in the COLR. This will result
in a change to a safety limit, but will not result in a significant
reduction in the margin of safety provided by the safety limit. As
discussed in the application, changing the SLMCPR methodology to one
based on a 95% probability with 95% confidence that no fuel rods
experience transition boiling during an anticipated transient
instead of the current limit based on ensuring that 99.9% of the
fuel rods are not susceptible to boiling transition does not have a
significant effect on plant response to any analyzed accident. The
SLMCPR and the TS Llimiting Condition for Operation (LCO) on MCPR
continue to provide the same level of assurance as the current
limits and do not reduce a margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jon P. Christinidis, DTE Energy, Expert
Attorney--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-
1279.
NRC Branch Chief: David J. Wrona.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2, Pope County, Arkansas
Entergy Operations, Inc.; System Energy Resources, Inc.; Cooperative
Energy, A Mississippi Electric Cooperative; and Entergy Mississippi,
LLC, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New
York
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Entergy Louisiana, LLC, and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: January 31, 2019. A publicly-available
version is in ADAMS under Accession No. ML19032A256.
Description of amendment request: The amendments would revise the
technical specifications (TSs) for each of these facilities based on
Technical Specifications Task Force (TSTF) Traveler TSTF-529, ``Clarify
Use and Application Rules,'' Revision 4 (ADAMS Accession No.
ML16062A271). Specifically, the changes would revise and clarify the TS
usage rules for completion times, limiting conditions for operation
(LCOs), and surveillance requirements (SRs).
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 Section 1.3 and LCO 3.0.4 have no effect
on the requirement for systems to be Operable and have no effect on
the application of TS actions. The proposed change to SR 3.0.3 (or
equivalent) states that the allowance may only be used when there is
a reasonable expectation the surveillance will be met when
performed.
Since the proposed changes do not significantly affect system
Operability, the proposed changes will have no significant effect on
the initiating events for accidents previously evaluated and will
have no significant effect on the ability of the systems to mitigate
accidents previously evaluated.
Therefore, it is concluded that these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the TS usage rules do not affect the
design or function of any plant systems. The proposed changes do not
change the Operability requirements for plant systems or the actions
taken when plant systems are not operable.
Therefore, it is concluded that the changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes clarify the application of Section 1.3 and
LCO 3.0.4 and do not result in changes in plant operation. SR 3.0.3
(or equivalent) is revised to allow application of SR 3.0.3 when an
SR has not been previously performed if there is reasonable
expectation that the SR will be met when performed. This expands the
use of SR 3.0.3 while ensuring the affected system is capable of
performing its safety function. As a result, plant safety is either
improved or unaffected.
Therefore, it is concluded that the changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anna Vinson Jones, Senior Counsel, Entergy
Services,
[[Page 14146]]
Inc., 101 Constitution Avenue NW, Suite 200 East, Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Exelon Generation Company, LLC and Exelon FitzPatrick, LLC, Docket No.
50-333, James A. FitzPatrick Nuclear Power Plant (FitzPatrick), Oswego
County, New York
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station (Limerick), Units 1 and 2, Montgomery County,
Pennsylvania
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of amendment request: February 1, 2019, as supplemented by
letter dated March 7, 2019. Publicly-available versions are in ADAMS
under Accession Nos. ML19032A624 and ML19066A162, respectively.
Description of amendment request: The proposed amendments would
revise the technical specification (TS) requirements for these
facilities related to the safety limit minimum critical power ratio
(MCPR) and the core operating limits report (COLR). The proposed
amendments are based on Technical Specification Task Force (TSTF)
traveler TSTF-564, Revision 2, ``Safety Limit MCPR'' (ADAMS Accession
No. ML18297A361). The proposed amendments for Limerick and FitzPatrick
would also make changes to the MCPR and COLR requirements that are
outside the scope of TSTF-564, Revision 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is
presented below:
1. Do the proposed amendments involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendments revise the TS requirements for the
safety limit MCPR and the list of core operating limits to be
included in the COLR. The safety limit MCPR is not an initiator of
any accident previously evaluated. The revised safety limit values
will continue to ensure for all accidents previously evaluated that
the fuel cladding will be protected from failure due to transition
boiling. The proposed amendments for Limerick, Units 1 and 2, also
include a revision to point to MCPR limits specified in the COLR and
clarify references to other specifications. The proposed amendment
for FitzPatrick also revises the COLR methodology references by
deleting references that are no longer needed and clarifying the
remaining reference. The proposed changes do not affect plant
operation or any procedural or administrative controls on plant
operation that affect the functions of preventing or mitigating any
accidents previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed amendments create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendments revise the TS requirements for the
safety limit MCPR and the list of core operating limits to be
included in the COLR. The proposed amendments for Limerick, Units 1
and 2, also include a revision to point to MCPR limits specified in
the COLR and clarify references to other specifications. The
proposed amendment for FitzPatrick also revises the COLR methodology
references by deleting references that are no longer needed and
clarifying the remaining reference. The proposed change will not
affect the design function or operation of any structures, systems
or components. No new equipment will be installed. As a result, the
proposed changes will not create any credible new failure
mechanisms, malfunctions, or accident initiators not considered in
the design and licensing bases.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed amendments involve a significant reduction in
a margin of safety?
Response: No.
The proposed amendments revise the TS safety limit MCPR and the
list of core operating limits to be included in the COLR. The
proposed amendments for Limerick, Units 1 and 2, also include a
revision to point to MCPR limits specified in the COLR and clarify
references to other specifications. The proposed amendment for
FitzPatrick also revises the COLR methodology references by deleting
references that are no longer needed and clarifying the remaining
reference. This will result in a change to a safety limit, but will
not result in a significant reduction in the margin of safety
provided by the safety limit. As discussed in the application,
changing the safety limit MCPR methodology to one based on a 95
percent probability with 95 percent confidence that no fuel rods
experience transition boiling during an anticipated transient
instead of the current limit based on ensuring that 99.9 percent of
the fuel rods are not susceptible to boiling transition does not
have a significant effect on plant response to any analyzed
accident.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-010, 50-237, and 50-249,
Dresden Nuclear Power Station, Units 1, 2, and 3, Grundy County,
Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
171, 50-277, and 50-278, Peach Bottom Atomic Power Station, Units 1, 2,
and 3, York and Lancaster Counties, Pennsylvania
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: March 1, 2019. Publicly-available
version is in ADAMS under Accession No. ML19063A685.
Description of amendment request: The amendments would revise the
[[Page 14147]]
emergency action levels (EALs) in the emergency plan for each site. The
proposed changes are based primarily on the resolution of emergency
preparedness frequently asked questions (EPFAQs) and industry best-
practices. Editorial changes are also proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed amendments involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes involving revisions to existing NRC-
approved [Nuclear Energy Institute guidance document] NEI 99-01,
Revision 6, EALs, as clarified by the NRC through the EPFAQ process,
for the affected facilities do not reduce the capability to meet the
emergency planning requirements established in 10 CFR 50.47 and 10
CFR 50, Appendix E. The proposed changes do not reduce the
functionality, performance, or capability of Exelon's ERO [emergency
response organization] to respond in mitigating the consequences of
any design basis accident.
The probability of a reactor accident requiring implementation
of Emergency Plan EALs has no relevance in determining whether the
proposed changes to the EALs reduce the effectiveness of the
Emergency Plans. As discussed in Section D, ``Planning Basis,'' of
NUREG-0654, Revision 1, ``Criteria for Preparation and Evaluation of
Radiological Emergency Response Plans and Preparedness in Support of
Nuclear Power Plants;''
``. . . The overall objective of emergency response plans is to
provide dose savings (and in some cases immediate life saving) for a
spectrum of accidents that could produce offsite doses in excess of
Protective Action Guides (PAGs). No single specific accident
sequence should be isolated as the one for which to plan because
each accident could have different consequences, both in nature and
degree. Further, the range of possible selection for a planning
basis is very large, starting with a zero point of requiring no
planning at all because significant offsite radiological accident
consequences are unlikely to occur, to planning for the worst
possible accident, regardless of its extremely low likelihood. . .
.''
Therefore, Exelon did not consider the risk insights regarding
any specific accident initiation or progression in evaluating the
proposed changes.
The proposed changes do not involve any physical changes to
plant equipment or systems, nor do they alter the assumptions of any
accident analyses. The proposed changes do not adversely affect
accident initiators or precursors nor do they alter the design
assumptions, conditions, and configuration or the manner in which
the plants are operated and maintained. The proposed changes do not
adversely affect the ability of Structures, Systems, or Components
(SSCs) to perform their intended safety functions in mitigating the
consequences of an initiating event within the assumed acceptance
limits.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed amendments create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes involving revisions to existing NRC-
approved NEI 99-01, Revision 6, EALs, as clarified by the NRC
through the EPFAQ process, for the affected facilities do not
involve any physical changes to plant systems or equipment. The
proposed changes do not involve the addition of any new plant
equipment. The proposed changes will not alter the design
configuration, or method of operation of plant equipment beyond its
normal functional capabilities. Exelon ERO functions will continue
to be performed as required. The proposed changes do not create any
new credible failure mechanisms, malfunctions, or accident
initiators.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from those that have been
previously evaluated.
3. Do the proposed amendments involve a significant reduction in
a margin of safety?
Response: No.
The proposed changes involving revisions to existing NRC-
approved NEI 99-01, Revision 6, EALs, as clarified by the NRC
through the EPFAQ process, for the affected facilities do not alter
or exceed a design basis or safety limit. There is no change being
made to safety analysis assumptions, safety limits, or limiting
safety system settings that would adversely affect plant safety as a
result of the proposed changes. There are no changes to setpoints or
environmental conditions of any SSC or the manner in which any SSC
is operated. Margins of safety are unaffected by the proposed
changes to the EALs based on further NRC clarification through the
EPFAQ. The applicable requirements of 10 CFR 50.47 and 10 CFR 50,
Appendix E will continue to be met.
Therefore, the proposed changes do 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
requested amendments involve 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. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois,
and Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station
(QCNPS), Units 1 and 2, Rock Island County, Illinois
Date of amendment request: December 5, 2018. A publicly-available
version is in ADAMS under Accession No. ML18339A009.
Description of amendment request: The amendments would revise the
technical specifications for both the single recirculation loop and two
recirculation loop Safety Limit Minimum Critical Power Ratio (SLMCPR)
limits for the DNPS and QCNPS units. The proposed decrease in these
limits improves operational flexibility through the recapture of
margins that are available as a result of the transition to Framatome,
Inc. using NRC-approved SLMCPR calculation methodology.
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 SLMCPR values have been determined using NRC-
approved methods discussed in AREVA Topical Report ANP-10307PA,
Revision 0, ``AREVA MCPR Safety Limit Methodology for Boiling Water
Reactors,'' dated June 2011. The proposed SLMCPRs for two
recirculation loop and single recirculation loop operation ensure
that the acceptance criterion continues to be met (i.e., at least
99.9 percent of all fuel rods in the core do not experience boiling
transition).
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
proposed license amendments do not involve any plant modifications
or operational changes that could affect system reliability or
performance, or that could affect the probability of operator error.
As such, the proposed changes do not affect any postulated accident
precursors. Since no individual precursors of an accident are
affected, the proposed license amendments do not involve a
significant increase in the probability of a previously analyzed
event.
The consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. The basis for the SLMCPR calculations is to ensure
that during normal operation and during anticipated operational
occurrences, at least 99.9 percent of all fuel
[[Page 14148]]
rods in the core do not experience boiling transition if the safety
limit is not exceeded.
Based on these considerations, the proposed changes do not
involve a significant increase in the consequences of a previously
analyzed accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Creation of the possibility of a new or different kind of
accident requires creating one or more new accident precursors. New
accident precursors may be created by modifications of plant
configuration, including changes in allowable modes of operation.
The SLMCPR is a TS numerical value calculated for two recirculation
loop operation and single recirculation loop operation to ensure at
least 99.9 percent of all fuel rods in the core do not experience
boiling transition if the safety limit is not exceeded. SLMCPR
values are calculated using NRC-approved methodology identified in
the TSs. The proposed SLMCPR values do not involve any new modes of
plant operation or any plant modifications and do not directly or
indirectly affect the failure modes of any plant systems or
components. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The SLMCPR provides a margin of safety by ensuring that at least
99.9 percent of the fuel rods do not experience boiling transition
during normal operation and anticipated operational occurrences if
the MCPR Safety Limit is not exceeded. Revision of the SLMCPR values
in TS 2.1.1.2, using an NRC-approved methodology, will ensure that
the current level of fuel protection is maintained by continuing to
ensure that the fuel design safety criterion is met (i.e., that no
more than 0.1 percent of the rods are expected to be in boiling
transition if the MCPR Safety Limit is not exceeded). The SLMCPRs
are verified to be bounding by cycle specific analyses prior to
power operations for each operating cycle. Therefore, the proposed
amendments do not result in 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.
Attorneys for licensee: Tamra (Tami) 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 No. 50-244, R. E. Ginna Nuclear
Power Plant, Wayne County, New York
Date of amendment request: February 15, 2018. A publicly-available
version is in ADAMS under Accession No. ML19045A282.
Description of amendment request: The amendment would revise
Technical Specification (TS) 5.5.15, ``Containment Leakage Rate Testing
Program,'' to reflect an increase to the existing Type A integrated
leak rate test program test interval from 10 years to 15 years, in
accordance with Nuclear Energy Institute (NEI) Report NEI 94-01,
Revision 2-A, ``Industry Guideline for Implementing Performance-Based
Option of 10 CFR part 50, Appendix J.'' The proposed change would also
reflect adoption of both the use of American National Standards
Institute/American Nuclear Society (ANSI/ANS) 56.8-2002, ``Containment
System Leakage Testing Requirements,'' and a more conservative
allowable test interval extension of 9 months for Type A leakage tests
in accordance with NEI 94-01, Revision 2-A. The amendment would also
make an administrative change to remove the exception under TS 5.5.15
for the one-time 15-year Type A test internal being performed after May
31, 1996, and performed prior to May 31, 2011, as this has already
occurred.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed activity involves the revision of the R. E. Ginna
Nuclear Power Plant (GNPP) Technical Specification (TS) 5.5.15,
``Primary Containment Leakage Rate Testing Program,'' to allow the
extension of the Type A Integrated Leakage Rate Test (ILRT)
containment test interval to 15 years. Per the guidance provided in
Nuclear Energy Institute (NEI) 94-01, Industry Guideline for
Implementing Performance-Based Option of 10 CFR 50, Appendix J,
Revision 2-A, the current Type A test interval of 10 years would be
extended on a permanent basis to no longer than 15 years from the
last Type A test.
The proposed interval extensions do not involve either a
physical change to the plant or a change in the manner in which the
plant is operated or controlled. The containment is designed to
provide an essentially leak tight barrier against the uncontrolled
release of radioactivity to the environment for postulated
accidents. As such, the containment and the testing requirements
invoked to periodically demonstrate the integrity of the containment
exist to ensure the plant's ability to mitigate the consequences of
an accident, and do not involve the prevention or identification of
any precursors of an accident.
The change in Type A test frequency to once-per-fifteen-years,
measured as an increase to the total integrated plant risk for those
accident sequences influenced by Type A testing, based on the
probabilistic risk assessment (PRA) is 0.29 person-Roentgen
equivalent man (rem)/year. Electric Power Research Institute (EPRI)
Report No. 1009325, Revision 2A states that a very small population
dose is defined as an increase of less than 1.0 person-rem per year
or less than 1 percent of the total population dose, whichever is
less restrictive for the risk impact assessment of the extended ILRT
intervals. This is consistent with the Nuclear Regulatory Commission
(NRC) Final Safety Evaluation which endorsed NEI 94-01 and EPRI
Report No. 1009325, Revision 2A. Moreover, the risk impact when
compared to other severe accident risks is negligible. Therefore,
the proposed extension does not involve a significant increase in
the probability of an accident previously evaluated.
In addition, as documented in NUREG-1493, ``Performance-Based
Containment Leak-Test Program,'' dated September 1995, Types B and C
tests have identified a very large percentage of containment leakage
paths, and the percentage of containment leakage paths that are
detected only by Type A testing is very small. The GNPP Type A test
history supports this conclusion.
The integrity of the containment is subject to two types of
failure mechanisms that can be categorized as: (1) Activity based,
and (2) time based. Activity based failure mechanisms are defined as
degradation due to system and/or component modifications or
maintenance. Local leak rate test requirements and administrative
controls such as configuration management and procedural
requirements for system restoration ensure that containment
integrity is not degraded by plant modifications or maintenance
activities. The design and construction requirements of the
containment combined with the containment inspections performed in
accordance with American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel (B&PV) Code, Section XI, ``Rules for
Inservice Inspection of Nuclear Power Plant Components,''
Containment Maintenance Rule Inspections, Containment Coatings
Program and TS requirements serve to provide a high degree of
assurance that the containment would not degrade in a manner that is
detectable only by a Type A test (ILRT). Based on the above, the
proposed test interval extensions do not significantly increase the
consequences of an accident previously evaluated.
This proposed amendment also deletes the exception previously
granted to allow one-time extension of the ILRT test frequency for
GNPP. Specifically, TS 5.5.15, item a. is deleted, as it requires
the first Type A test performed after May 31, 1996, to be performed
by May 31, 2011. This exception was included in the TS for one-time
testing activities that would have already taken place by the time
this amendment is approved; therefore, deletion is solely an
administrative action that has no effect on any component and no
impact on how the unit is operated.
[[Page 14149]]
Therefore, the proposed changes do not result in a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment to the GNPP TS 5.5.15 involves the
extension of the GNPP Type A containment test interval from 10 years
to 15 years. 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; thereby, do not involve any accident precursors or
initiators.
The proposed change does not involve a physical modification to
the plant (i.e., no new or different type of equipment will be
installed) nor does it alter the design, configuration, or change
the manner in which the plant is operated or controlled beyond the
standard functional capabilities of the equipment.
This proposed amendment also deletes the exception previously
granted to allow one-time extension of the ILRT test frequency for
GNPP. Specifically, TS 5.5.15, item a. is deleted, as it requires
the first Type A test performed after May 31, 1996, to be performed
by May 31, 2011. This exception was included in the TS for one-time
testing activities that would have already taken place by the time
this amendment is approved; therefore, deletion is solely an
administrative action that has no effect on any component and no
impact on how the unit is operated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment to TS 5.5.15 involves the extension of
the GNPP Type A containment test interval to 15 years. This
amendment does not alter the manner in which safety limits, limiting
safety system set points, or limiting conditions for operation are
determined. The specific requirements and conditions of the TS
Containment Leak Rate Testing Program exist to ensure that the
degree of containment structural integrity and leak-tightness that
is considered in the plant safety analysis is maintained. The
overall containment leak rate limit specified by TS is maintained.
The proposed change involves the extension of the interval
between Type A containment leak rate tests for GNPP. The proposed
surveillance interval extension is bounded by the 15-year ILRT
interval currently authorized within NEI 94-01, Revision 2-A.
Industry experience supports the conclusion that Types B and C
testing detects a large percentage of containment leakage paths and
that the percentage of containment leakage paths that are detected
only by Type A testing is small. The containment inspections
performed in accordance with Option B to 10 CFR 50, Appendix J and
the overlapping inspection activities performed as part of ASME
Section Xl, and the TS serve to provide a high degree of assurance
that the containment would not degrade in a manner that is
detectable only by Type A testing. The combination of these factors
ensures that the margin of safety in the plant safety analysis is
maintained. The design, operation, testing methods and acceptance
criteria for Types A, B, and C containment leakage tests specified
in applicable codes and standards would continue to be met, with the
acceptance of this proposed change, since these are not affected by
changes to the Type A test intervals.
In addition, this proposed amendment also deletes the exception
previously granted to allow one-time extension of the ILRT test
frequency for GNPP. Specifically, TS 5.5.15, item a. is deleted, as
it requires the first Type A test performed after May 31, 1996, to
be performed by May 31, 2011. This exception was included in the TS
for one-time testing activities that would have already taken place
by the time this amendment is approved; therefore, deletion is
solely an administrative action that has no effect on any component
and no impact on how the unit is operated.
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.
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: James G. Danna.
FirstEnergy Nuclear Operating Company (FENOC), et al., Docket No. 50-
346, Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), Ottawa
County, Ohio
Date of amendment request: February 5, 2019. A publicly-available
version is in ADAMS under Accession No. ML19036A523.
Description of amendment request: By letter dated April 25, 2018
(ADAMS Accession No. ML18115A007), FENOC notified the NRC that DBNPS
will permanently cease power operations by May 31, 2020. The proposed
amendment would revise the DBNPS renewed facility operating license
(RFOL) and technical specifications (TSs) following the permanent
cessation of power operations to reflect the post-shutdown and
permanently defueled condition. The proposed amendment would eliminate
TS requirements and license conditions which would not be applicable
once DBNPS ceases power operations and can no longer place fuel in the
reactor vessel. The proposed amendment would also eliminate obsolete
license conditions. In addition, the proposed amendment would revise
several license conditions and TS requirements, including limiting
conditions for operation (LCOs), usage rules, definitions, surveillance
requirements (SRs), and administrative controls. FENOC also proposed to
revise the licensing bases for DBNPS, including the design bases
accident (DBA) 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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes would not take effect until DBNPS has
certified to the NRC that it has permanently ceased operation and
entered a permanently defueled condition. Because the 10 CFR part 50
license for DBNPS will no longer authorize operation of the reactor,
or emplacement or retention of fuel into the reactor vessel with the
certifications required by 10 CFR part 50.82(a)(1) submitted, as
specified in 10 CFR part 50.82(a)(2), the occurrence of postulated
accidents associated with reactor operation is no longer credible.
The remaining [Updated Final Safety Analysis Report] UFSAR
Chapter 15 postulated design basis accident (DBA) events that could
potentially occur at a permanently defueled facility would be a fuel
handling accident (FHA) in the spent fuel pool (SFP), the waste gas
decay tank rupture (WGDTR), and external causes. The FHA analyses
for DBNPS shows that, following 95 days of decay time after reactor
shutdown and provided the SFP water level requirements of TS LCO
3.7.14 are met, the dose consequences are acceptable without relying
on structures, systems, and components (SSCs) to remain functional
for accident mitigation during and following the event other than
the passive SFP structure. The remaining DBAs that support the
permanently shutdown and defueled condition do not rely on any
active safety systems for mitigation.
The probability of occurrence of previously evaluated accidents
is not increased, since safe storage and handling of fuel will be
the only operations performed, and therefore, bounded by the
existing analyses. Additionally, the occurrence of postulated
accidents associated with reactor operation
[[Page 14150]]
will no longer be credible in a permanently defueled reactor. This
significantly reduces the scope of applicable accidents.
The deletion of TS definitions and rules of usage and
application requirements that will not be applicable in a defueled
condition has no impact on facility SSCs or the methods of operation
of such SSCs. The deletion of design features and safety limits not
applicable to the permanently shut down and defueled status of DBNPS
has no impact on the remaining applicable DBAs.
The removal of LCOs or SRs that are related only to the
operation of the nuclear reactor or only to the prevention,
diagnosis, or mitigation of reactor-related transients or accidents
do not affect the applicable DBAs previously evaluated since these
DBAs are no longer applicable in the permanently defueled condition.
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 changes to delete or modify certain DBNPS RFOL, TS,
and current licensing bases (CLB) have no impact on facility SSCs
affecting the safe storage of spent irradiated fuel, or on the
methods of operation of such SSCs, or on the handling and storage of
spent irradiated fuel itself. The removal of TS that are related
only to the operation of the nuclear reactor, or only to the
prevention, diagnosis, or mitigation of reactor related transients
or accidents, cannot result in different or more adverse failure
modes or accidents than previously evaluated because the reactor
will be permanently shutdown and defueled.
The proposed modification or deletion of requirements of the
DBNPS RFOL, TS, and CLB do not affect systems credited in the
accident analysis for the remaining credible DBAs at DBNPS. The
proposed RFOL and PDTS [permanently defueled TSs] will continue to
require proper control and monitoring of safety significant
parameters and activities. The TS regarding SFP water level and
spent fuel storage is retained to preserve the current requirements
for safe storage of irradiated fuel. The proposed amendment does not
result in any new mechanisms that could initiate damage to the
remaining relevant safety barriers for defueled plants (fuel
cladding, spent fuel racks, SFP integrity, and SFP water level).
Since extended operation in a defueled condition and safe fuel
handling will be the only operation allowed, and therefore bounded
by the existing analyses, such a condition does not create the
possibility of 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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes are to delete or modify certain RFOL, TS,
and CLB once the DBNPS facility has been permanently shutdown and
defueled. Because the 10 CFR part 50 license for DBNPS will no
longer authorize operation of the reactor, or emplacement or
retention of fuel into the reactor vessel, the occurrence of
postulated accidents associated with reactor operation is no longer
credible. The remaining postulated DBA events that could potentially
occur at a permanently defueled facility would be a[n] FHA, WGDTR,
and external causes. The proposed amendment does not adversely
affect the inputs or assumptions of any of the design basis
analyses.
The proposed changes are limited to those portions of the RFOL,
TS, and CLB that are not related to the safe storage of irradiated
fuel. The requirements that are proposed to be revised or deleted
from the RFOL, TS, and CLB are not credited in the updated
applicable accident analysis for the remaining applicable postulated
accidents, and as such, do not contribute to the margin of safety
associated with the accident analysis. Postulated design basis
accidents involving the reactor will no longer be possible because
the reactor will be permanently shutdown and defueled, and DBNPS
will no longer be authorized to operate the reactor.
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: Rick Giannantonio, General Counsel,
FirstEnergy Corporation, Mail Stop A-GO-15, 76 South Main Street,
Akron, OH 44308.
NRC Branch Chief: David J. Wrona.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: February 26, 2019. A publicly-available
version is in ADAMS under Accession No. ML19060A060.
Description of amendment request: The proposed amendment would
expand the criteria within technical specification (TS) 3.2.1
surveillance requirements to apply a revised penalty factor to measured
transient FQ(Z) in response to Westinghouse Nuclear Safety
Advisory Letter, NSAL-15-1, ``Heat Flux Hot Channel Factor Technical
Specification Surveillance.''
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 of occurrence or consequences of an accident
previously evaluated?
Response: No.
The proposed amendment to add an additional surveillance
requirement, to apply the penalty factor of 1.02 or a factor
specified in the COLR [core operating limit report], whichever is
greater, to the transient FQ(Z) calculation, ensures that
the assumptions and inputs to the safety analyses remain valid and
does not result in actions that would increase the probability or
consequences of any accident previously evaluated.
The design of the protection systems will be unaffected. The
reactor protection system and engineered safety feature actuation
system will continue to function in a manner consistent with the
plant design basis. All design, material and construction standards
that were applicable prior to the request are maintained.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident-
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Operation in accordance with the revised TS and its limits
precludes new challenges to systems or structures that might
introduce a new type of accident. All design and performance
criteria will continue to be met and no new single failure
mechanisms will be created. The proposed change for resolution of
Westinghouse NSAL-15-1 does not involve the alteration of plant
equipment or introduce unique operational modes or accident
precursors. Therefore it does not create the potential for a
different kind of accident.
Therefore, the proposed changes do not create the possibility of
a new or, different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Operation in accordance with the revised TS and its limits
preserves the margins assumed in the safety analyses. This ensures
that all design and performance criteria associated with the safety
analysis will continue to be met and that the margin of safety is
not affected.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
[[Page 14151]]
NRC Branch Chief: David J. Wrona.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: February 26, 2019. A publicly-available
version is in ADAMS under Accession No. ML19063A498.
Description of amendment request: The proposed amendment would
adopt Technical Specification Task Force (TSTF) Traveler TSTF-563,
``Revise Instrument Testing Definitions to Incorporate the Surveillance
Frequency Control Program.'' TSTF-563 revises the TS definitions of
Channel Calibration, Channel Operational Test, and Trip Actuating
Device Operational Test.
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 of occurrence or consequences of an accident
previously evaluated?
Response: No.
The proposed change revises the TS [technical specification]
definitions of Channel Calibration, COT [channel operational test],
and TADOT [trip actuating device operational test] to allow the
frequency for testing the components or devices in each step to be
determined in accordance with the TS Surveillance Frequency Control
Program. All components in the channel continue to be tested. The
frequency at which a channel test is performed is not an initiator
of any accident previously evaluated, so the probability of an
accident is not affected by the proposed change. The channels
surveilled in accordance with the affected definitions continue to
be required to be operable and the acceptance criteria of the
surveillances are unchanged. As a result, any mitigating functions
assumed in the accident analysis will continue to be performed.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident-
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the TS definitions of Channel
Calibration, COT, and TADOT to allow the frequency for testing the
components or devices in each step to be determined in accordance
with the TS Surveillance Frequency Control Program. The design
function or operation of the components involved are not affected
and there is no physical alteration of the plant (i.e., no new or
different type of equipment will be installed). No credible new
failure mechanisms, malfunctions, or accident initiators not
considered in the design and licensing bases are introduced. The
changes do not alter assumptions made in the safety analysis. The
proposed changes are consistent with the safety analysis
assumptions.
Therefore, the proposed changes do not create the possibility of
a new or, different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the TS definitions of Channel
Calibration, COT, and TADOT to allow the frequency for testing the
components or devices in each step to be determined in accordance
with the TS Surveillance Frequency Control Program. The Surveillance
Frequency Control Program assures sufficient safety margins are
maintained, and that design, operation, surveillance methods, and
acceptance criteria specified in applicable codes and standards (or
alternatives approved for use by the NRC) will continue to be met as
described in the plants' licensing basis. The proposed change does
not adversely affect existing plant safety margins, or the
reliability of the equipment assumed to operate in the safety
analysis. As such, there are no changes being made to safety
analysis assumptions, safety limits, or limiting safety system
settings that would adversely affect plant safety as a result of the
proposed change. Margins of safety are unaffected by the method of
determining surveillance test intervals under an NRC-approved
licensee-controlled program.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
NRC Branch Chief: David J. Wrona.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
No. 1, Rockingham wCounty, New Hampshire
Date of amendment request: October 4, 2018. A publicly-available
version is in ADAMS under Accession No. ML18277A377.
Description of amendment request: The amendment would revise the
Seabrook Station, Unit No. 1 (Seabrook), Technical Specifications (TSs)
and Surveillance Requirements (SRs) associated with the control rods.
The amendment would adopt changes provided in Technical Specifications
Task Force (TSTF) Traveler TSTF-234, ``Add Action for More than One
[D]RPI [Digital Rod Position Indicator] Inoperable,'' and TSTF-547,
``Clarification of Rod Position Requirements,'' and make various other
changes to align the Seabrook TSs more closely with NUREG-1431,
``Standard Technical Specifications--Westinghouse Plants.'' In all, the
amendment would revise SR 4.1.1.1.1, SR 4.1.1.2, TS 3.1.3.1, SR
4.1.3.1.1, TS 3.1.3.2, SR 4.1.3.2, TS 3.1.3.3, SR 4.1.3.3, TS 3.1.3.5,
SR 4.1.3.5, TS 3.1.3.6, SR 4.1.3.6, TS 3.10.5, SR 4.10.5, and TS
6.8.1.6.b.
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.
Control and shutdown rods are assumed to insert into the core to
shut down the reactor in evaluated accidents. Rod insertion limits
ensure that adequate negative reactivity is available to provide the
assumed shutdown margin (SDM). Rod alignment limits maintain an
appropriate power distribution and reactivity insertion profile.
Control and shutdown rods are initiators to several accidents
previously evaluated, such as rod ejection. The proposed change does
not change the limiting conditions for operation for the rods or
make any technical changes to the surveillance requirements
governing the rods. Therefore, the proposed change has no
significant effect on the probability of any accident previously
evaluated.
Adding new TS Actions to provide a limited time to repair rod
control system failures has no effect on the SDM assumed in the
accident analysis as the proposed Actions require verification that
SDM is maintained. The effects on power distribution will not cause
a significant increase in the consequences of any accident
previously evaluated as all TS requirements on power distribution
continue to be applicable.
The proposed change to resolve the conflicts in the TS ensures
that the intended Actions are followed when equipment is inoperable.
Actions taken with inoperable equipment are not assumptions in the
accidents previously evaluated and have no significant effect on the
consequences.
The capability of any operable TS-required equipment to perform
its specified safety function is not impacted by the proposed
change. As a result, the outcomes of accidents previously evaluated
are unaffected. Therefore, the proposed changes do not result in a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
[[Page 14152]]
Response: No.
The proposed change does not challenge the integrity or
performance of any safety-related systems. No plant equipment is
installed or removed, and the changes do not alter the design,
physical configuration, or method of operation of any plant system
or component. No physical changes are made to the plant, so no new
causal mechanisms are introduced. Therefore, the proposed changes to
the TS do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The ability of the control rods to perform their designated
safety function is unaffected by the proposed changes. The proposed
changes do not alter any safety analyses assumptions, safety limits,
limiting safety system settings, or method of operating the plant.
The proposed change to provide time to repair rods that are operable
but immovable does not result in a significant reduction in the
margin of safety because all rods must be verified to be operable,
and all other banks must be within the insertion limits. The changes
do not adversely affect plant operating margins or the reliability
of equipment credited in the safety analyses. Therefore, the
proposed changes do not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis, and based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Debbie Hendell, Managing Attorney, Florida
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: James G. Danna.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: February 27, 2019. A publicly-available
version is in ADAMS under Accession No. ML19058A221.
Description of amendment request: The proposed change is consistent
with Technical Specifications Task Force (TSTF) Traveler TSTF-546,
Revision 0, ``Revise APRM [Average Power Range Monitor] Channel
Adjustment Surveillance Requirement'' (ADAMS Accession No.
ML17205A444). The amendment would alter Surveillance Requirement (SR)
4.3.1.1 of Technical Specification 3.3.1, ``Reactor Protection System
Instrumentation.'' The change would revise the SR to verify that
calculated (i.e., calorimetric heat balance) power is no more than 2
percent greater than the APRM channel output. The SR requires the APRM
channel to be adjusted such that calculated power is no more than 2
percent greater than the APRM indicated power when operating at >=24
percent of rated thermal power. This change would revise the SR to
distinguish between APRM indications that are consistent with the
accident analyses and those that provide additional margin.
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 APRM system and the RPS are not initiators of any accidents
previously evaluated. As a result, the proposed change does not
affect the probability of any accident previously evaluated. The
APRM system and the Reactor Protection System (RPS) functions act to
mitigate the consequences of accidents previously evaluated. The
reliability of APRM system and the RPS is not significantly affected
by removing the gain adjustment requirement on the APRM channels
when the APRMs are calibrated conservatively with respect to the
calculated heat balance. This is because the actual core thermal
power at which the reactor will automatically trip is lower, thereby
increasing the margin to the core thermal limits and the limiting
safety system settings assumed in the safety analyses. The
consequences of an accident during the adjustment of the APRM
instrumentation are no different from those during the existing
surveillance testing period or the existing time allowed to restore
the instruments to operable status. As a result, the ability of the
APRM system and the RPS to mitigate any accident previously
evaluated is not significantly affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed change does not alter the protection system design,
create new failure modes, or change any modes of operation. The
proposed change does not involve a physical alteration of the plant;
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 margin of safety provided by the APRM system and the RPS is
to ensure that the reactor is shut down automatically when plant
parameters exceed the setpoints for the system. Any reduction in the
margin of safety resulting from the adjustment of the APRM channels
while continuing operation is considered to be offset by delaying a
plant shutdown (i.e., a transient) for a short time with the APRM
system, the primary indication of core power and an input to the
RPS, not calibrated. Additionally, the short time period required
for adjustment is consistent with the time allowed by Technical
Specifications to restore the core power distribution parameters to
within limits and is acceptable based on the low probability of a
transient or design basis accident occurring simultaneously with
inaccurate APRM channels.
The proposed change does not alter setpoints or limits
established or assumed by the accident analyses. The Technical
Specifications continue to require operability of the RPS functions,
which provide core protection for postulated reactivity insertion
events occurring during power operating conditions consistent with
the plant 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: Steven Fleischer, PSEG Services Corporation,
80 Park Plaza, T-5, Newark, NJ 07102.
NRC Branch Chief: James G. Danna.
Tennessee Valley Authority (TVA), Docket Nos. 50-390 and 50-391, Watts
Bar Nuclear Plant, Units 1 and 2, Rhea County, Tennessee
Date of amendment request: October 12, 2018. A publicly-available
version is in ADAMS under Accession No. ML18288A352.
Description of amendment request: The amendments would revise the
Technical Specifications (TS) by the adoption, with administrative and
technical variations, of Technical Specification Task Force (TSTF)
Traveler TSTF-425, Revision 3, ``Relocate Surveillance Frequencies to
Licensee Control--Risk Informed Technical Specification Task Force
(RITSTF) Initiative 5b.'' TSTF-425, Revision 3, provides for the
relocation of specific surveillance frequencies to a licensee-
controlled program. Additionally, the change would add a new program,
the Surveillance Frequency Control Program (SFCP), to TS Section 5.0,
``Administrative Controls.''
[[Page 14153]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change relocates the specified frequencies for
periodic surveillance requirements to licensee control under a new
SFCP. Surveillance frequencies are not an initiator to any accident
previously evaluated. As a result, the probability of any accident
previously evaluated is not significantly increased. The systems and
components required by the technical specifications for which the
surveillance frequencies are relocated are still required to be
operable, meet the acceptance criteria for the surveillance
requirements, and be capable of performing any mitigation function
assumed in the accident analysis. As a result, the consequences of
any accident previously evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the change does not impose any new or
different requirements. The change does not alter assumptions made
in the safety analysis. The proposed change is consistent with the
safety analysis assumptions and current plant operating practice.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria
for structures, systems, [and] components, specified in applicable
codes and standards (or alternatives approved for use by the NRC)
will continue to be met as described in the plant licensing basis
(including the final safety analysis report and bases to TS),
because these are not affected by changes to the surveillance
frequencies. Similarly, there is no effect to safety analysis
acceptance criteria as described in the plant licensing basis. To
evaluate a change in the relocated surveillance frequency, TVA will
perform a probabilistic risk evaluation using the guidance contained
in NRC approved NEI [Nuclear Energy Institute] 04-10, Revision 1, in
accordance with the TS SFCP. This methodology provides reasonable
acceptance guidelines and methods for evaluating the risk increase
of proposed changes to surveillance frequencies consistent with
Regulatory Guide 1.177.
Therefore, the proposed 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 Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: February 1, 2019. A publicly-available
version is in ADAMS under Accession No. ML19032A632.
Description of amendment request: The amendments would adopt
Technical Specification Task Force Traveler TSTF-563, ``Revise
Instrument Testing Definitions to Incorporate the Surveillance
Frequency Control Program.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequence of an accident previously evaluated?
Response: No.
The proposed change revises the TS [Technical Specification]
definitions of Channel Calibration, COT [Channel Operational Test],
and TADOT [Trip Actuation Device Operational Test] to allow the
frequency for testing the components or devices in each step to be
determined in accordance with the TS Surveillance Frequency Control
Program. All components in the channel continue to be tested. The
frequency at which a channel test is performed is not an initiator
of any accident previously evaluated, so the probability of an
accident is not affected by the proposed change. The channels
surveilled in accordance with the affected definitions continue to
be required to be operable and the acceptance criteria of the
surveillances are unchanged. As a result, any mitigating functions
assumed in the accident analysis will continue to be performed.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the TS definitions of Channel
Calibration, COT, and TADOT to allow the frequency for testing the
components or devices in each step to be determined in accordance
with the TS Surveillance Frequency Control Program. The design
function or operation of the components involved are not affected
and there is no physical alteration of the plant (i.e., no new or
different type of equipment will be installed). No credible new
failure mechanisms, malfunctions, or accident initiators not
considered in the design and licensing bases are introduced. The
changes do not alter assumptions made in the safety analysis. The
proposed changes are consistent with the safety analysis
assumptions.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change revises the TS definitions of Channel
Calibration, COT, and TADOT to allow the frequency for testing the
components or devices in each step to be determined in accordance
with the TS Surveillance Frequency Control Program. The Surveillance
Frequency Control Program assures sufficient safety margins are
maintained, and that design, operation, surveillance methods, and
acceptance criteria specified in applicable codes and standards (or
alternatives approved for use by the Nuclear Regulatory Commission
(NRC)) will continue to be met as described in the plants' licensing
basis. The proposed change does not adversely affect existing plant
safety margins, or the reliability of the equipment assumed to
operate in the safety analysis. As such, there are no changes being
made to safety analysis assumptions, safety limits, or limiting
safety system settings that would adversely affect plant safety as a
result of the proposed change. Margins of safety are unaffected by
method of determining surveillance test intervals under an NRC-
approved licensee-controlled program.
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.
[[Page 14154]]
NRC Branch Chief: Undine Shoop.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Unit 1, Coffey County, Kansas
Date of amendment request: January 23, 2019, as supplemented by
letter dated March 11, 2019. Publicly-available versions are in ADAMS
under Accession Nos. ML19036A772 and ML19078A131, respectively.)
Description of amendment request: The amendment would revise
technical specification (TS) requirements in Section 1.3, ``Completion
Times,'' and Section 3.0, ``Limiting Condition for Operation (LCO)
Applicability,'' regarding LCO and surveillance requirement (SR) usage.
The proposed changes are consistent with the NRC-approved Technical
Specifications Task Force (TSTF) Traveler TSTF-529, Revision 4,
``Clarify Use and Application Rules,'' using the consolidated line item
improvement process (ADAMS Accession No. ML16062A271). The model safety
evaluation was approved by the NRC in a letter dated April 21, 2016
(ADAMS Package Accession No. ML16060A441).
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 Section 1.3 and LCO 3.0.4 have no effect
on the requirement for systems to be Operable and have no effect on
the application of TS actions. The proposed change to SR 3.0.3
states that the allowance may only be used when there is a
reasonable expectation the surveillance will be met when performed.
Since the proposed change does not significantly affect system
Operability, the proposed change will have no significant effect on
the initiating events for accidents previously evaluated and will
have no significant effect on the ability of the systems to mitigate
accidents previously evaluated.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to the TS usage rules does not affect the
design or function of any plant systems. The proposed change does
not change the Operability requirements for plant systems or the
actions taken when plant systems are not operable.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change clarifies the application of Section 1.3 and
LCO 3.0.4 and does not result in changes in plant operation. SR
3.0.3 is revised to allow application of SR 3.0.3 when an SR has not
been previously performed if there is reasonable expectation that
the SR will be met when performed. This expands the use of SR 3.0.3
while ensuring the affected system is capable of performing its
safety function. As a result, plant safety is either improved or
unaffected.
Therefore, it is concluded that this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 1200 17th Street NW, Washington, DC 20036.
NRC Branch Chief: Robert J. Pascarelli.
IV. Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
Dominion Energy Nuclear Connecticut, Inc., Docket No. 50-423, Millstone
Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: April 4, 2018, as supplemented by letter
dated October 22, 2018.
Brief description of amendment: The amendment revised ACTION 18 in
Technical Specifications Table 3.3-3, Functional Unit 7.e, ``Control
Building Inlet Ventilation Radiation,'' for Millstone Power Station,
Unit No. 3, to allow continued fuel handling and reactor operation with
inoperable inlet radiation monitoring instrumentation provided that one
train of the control room emergency ventilation system is operating in
the emergency mode. The technical specification change specifies that
one train of the control room emergency ventilation system be placed in
the emergency mode of operation within 7 days if one radiation monitor
channel is inoperable, or immediately, if both radiation monitor
channels are inoperable.
Date of issuance: March 21, 2019.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 272. A publicly-available version is in ADAMS under
Accession No. ML19042A277; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-49: The amendment
revised the Renewed Facility Operating License and Technical
Specifications.
Date of initial notice in Federal Register: July 17, 2018 (83 FR
33266). The supplemental letter dated October 22, 2018, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards
[[Page 14155]]
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 21, 2019.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket Nos. 50-003 and 50-247, Indian
Point Nuclear Generating Unit Nos. 1 and 2 (Indian Point 1 and Indian
Point 2), Westchester County, New York
Date of amendment request: June 20, 2018. A publicly-available
version is in ADAMS under Accession No. ML18179A173.
Brief description of amendments: The amendments deleted certain
license conditions from the Indian Point 1 and Indian Point 2 Operating
Licenses that impose specific requirements on the decommissioning trust
agreement. With approval of these amendments, the provisions of 10 CFR
50.75(h), which specify the regulatory requirements for decommissioning
trust funds, apply to the licensee, Entergy Nuclear Operations, Inc.,
for Indian Point 1 and Indian Point 2.
Date of issuance: March 21, 2019.
Effective date: As of the date of issuance, and shall be
implemented within 60 days of issuance.
Amendment Nos.: 61 (Unit No. 1) and 289 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML19065A101;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Provisional Operating License No. DPR-5 and Renewed Facility
Operating License No. DPR-26: The amendments revised the Operating
Licenses.
Date of initial notice in Federal Register: September 11, 2018 (83
FR 45984).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 21, 2019.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., Cooperative
Energy, A Mississippi Electric Cooperative, and Entergy Mississippi,
LLC, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: March 26, 2018.
Brief description of amendment: The amendment revised the Updated
Final Safety Analysis Report descriptions for the replacement of the
Turbine First Stage Pressure output signals with Power Range Neutron
Monitoring System output signals.
Date of issuance: March 12, 2019.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No: 217. A publicly-available version is in ADAMS under
Accession No. ML18215A196; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-29: The amendment
revised the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: June 5, 2018 (83 FR
26115).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 12, 2019.
No significant hazards consideration comments received: No.
Entergy Operations, Inc.; System Energy Resources, Inc.; Cooperative
Energy, A Mississippi Electric Cooperative; and Entergy Mississippi,
LLC, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS),
Claiborne County, Mississippi
Date of amendment request: April 27, 2018, as supplemented by
letter dated October 10, 2018.
Brief description of amendment: The amendment revised the GGNS
Emergency Plan to adopt an Emergency Action Level scheme based on
Nuclear Energy Institute (NEI) guidance in NEI 99-01, Revision 6,
``Development of Emergency Action Levels for Non-Passive Reactors,''
dated November 2012, which was endorsed by the NRC by letter dated
March 28, 2013.
Date of issuance: March 12, 2019.
Effective date: As of the date of issuance and shall be implemented
within 365 days of issuance.
Amendment No: 216. A publicly-available version is in ADAMS under
Accession No. ML19025A023; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-29: The amendment
revised the GGNS Emergency Plan.
Date of initial notice in Federal Register: June 5, 2018 (83 FR
26104). The supplemental letter dated October 10, 2018, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 12, 2019.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-010, 50-237, and 50-249,
Dresden Nuclear Power Station, Units 1, 2, and 3, Grundy County,
Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: January 31, 2018, as supplemented by
letters dated July 27 and November 29, 2018.
Brief description of amendments: The amendments revise the
emergency response organization positions identified in the emergency
plan for each site.
Date of issuance: March 21, 2019.
Effective date: As of the date of issuance and shall be implemented
on or before December 31, 2019.
Amendment Nos.: Braidwood 201/201, Byron 206/206, Clinton 223,
Dresden 46/261/254, LaSalle 236/222, and Quad Cities 274/269. A
publicly-available version is in ADAMS under Accession No. ML19036A586.
Documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Facility Operating License Nos. NPF-72, NPF-77, NPF-37, NPF-66,
NPF-62, DPR-2, DPR-19, DPR-25, NPF-11, NPF-18, DPR-29, and DPR-30:
Amendments revised the emergency plans.
Date of initial notice in Federal Register: April 10, 2018 (83 FR
15417).
The Commission's related evaluation of the amendments is contained
in a safety evaluation dated March 21, 2019.
No significant hazards consideration comments received: No.
[[Page 14156]]
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: May 14, 2018, as supplemented by letter
dated November 20, 2018.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to increase the minimum load required
for the Emergency Diesel Generator (EDG) partial-load rejection
Surveillance Requirement (SR). Additionally, the amendments modified
the EDG voltage and frequency limits for the SR and established a
recovery period for the EDG(s) to return to steady-state conditions.
Date of issuance: March 18, 2019.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1, 285 and Unit 2, 279. A publicly-available
version is in ADAMS under Accession No. ML18354A673; 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: July 3, 2018 (83 FR
31185). The supplemental letter dated November 20, 2018, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 18, 2019.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket No. 52-025, Vogtle Electric
Generating Plant (VEGP), Unit 3, Burke County, Georgia
Date of amendment request: October 19, 2018.
Description of amendment: The amendment authorizes the Southern
Nuclear Operating Company to depart from certified AP1000 Design
Control Document (DCD) Tier 2* material that has been incorporated into
the Updated Final Safety Analysis Report (UFSAR). Specifically, the
proposed departure consists of changes to Tier 2* information in the
UFSAR (which includes the plant-specific DCD information) to change the
vertical reinforcement information provided in the VEGP Unit 3 column
line 1 wall from elevation 135'-3'' to 137'-0'' .
Date of issuance: March 13, 2019.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 156 for Unit 3. Publicly-available versions are in
an ADAMS package under Accession No. ML19044A500 which includes the
Safety Evaluation that references documents, located in that ADAMS
package, related to this amendment.
Facility Combined Licenses No. NPF-91: Amendment revised the
Facility Combined License.
Date of initial notice in Federal Register: November 20, 2018 (83
FR 58607).
The Commission's related evaluation of the amendment is contained
in the Safety Evaluation dated March 13, 2019.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant,
Unit 2, Rhea County, Tennessee
Date of amendment request: March 5, 2018, as supplemented by
letters dated April 27 and October 11, 2018.
Brief description of amendment: The amendment revised License
Condition 2.C.(4), concerning the use of the PAD4TCD computer program.
While the current License Condition permits the use of PAD4TCD for Unit
2, Cycles 1 and 2 only, the revision allows the use of PAD4TCD until
the Unit 2 steam generators (SGs) are replaced with SGs equivalent to
the existing SGs at Unit 1.
Date of issuance: March 20, 2019.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 26. A publicly-available version is in ADAMS under
Accession No. ML19046A286; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-96: Amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: December 4, 2018 (83 FR
62623). The supplemental letters dated April 27 and October 11, 2018,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 20, 2019.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar
Nuclear Plant (Watts Bar), Units 1 and 2, Rhea County, Tennessee
Date of amendment request: August 1, 2018, as supplemented by
letter dated March 4, 2019.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) to adopt, with minor variation, Technical
Specification Task Force (TSTF) Traveler TSTF-266-A, Revision 3,
``Eliminate the Remote Shutdown System Table of Instrumentation and
Controls.'' Specifically, the comparable TS Table 3.3.4-1, ``Remote
Shutdown System Instrumentation and Controls,'' was deleted from Watts
Bar, Units 1 and 2, TS 3.3.4, ``Remote Shutdown System.''
Date of issuance: March 18, 2019.
Effective date: As of the date of issuance and shall be implemented
by March 24, 2019.
Amendment Nos.: 124 and 25. A publicly-available version is in
ADAMS under Accession No. ML19066A009; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Facility Operating License Nos. NPF-90 and NPF-96: Amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: February 12, 2019 (84
FR 3510). The supplemental letter dated March 4, 2019, requested
expedited completion of the NRC review of 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 (NSHC) determination as published in the Federal
Register.
The Commission's related evaluation of the amendments and final
NSHC determination are contained in a Safety Evaluation dated March 18,
2019.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 29th day of March 2019.
For the Nuclear Regulatory Commission.
Craig G. Erlanger,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2019-06449 Filed 4-8-19; 8:45 am]
BILLING CODE 7590-01-P