[Federal Register Volume 82, Number 175 (Tuesday, September 12, 2017)]
[Notices]
[Pages 42844-42857]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-19214]
[[Page 42844]]
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NUCLEAR REGULATORY COMMISSION
[NRC-2017-0189]
Biweekly Notice: Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954,
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued, and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued, from August 15, 2017 to August 28, 2017. The
last biweekly notice was published on August 29, 2017.
DATES: Comments must be filed by October 12, 2017. A request for a
hearing must be filed by November 13, 2017.
ADDRESSES: You may submit comments by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0189. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: TWFN-8-D36M, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Paula Blechman, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-2242; email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2017-0189, facility name, unit
numbers, plant docket number, application date, and subject when
contacting the NRC about the availability of information for this
action. You may obtain publicly-available information related to this
action by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0189.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in this
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2017-0189, facility name, unit
numbers, 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 submissions into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated, or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment
period such that failure to act in a timely way would result, for
example in derating or shutdown of the facility. If the Commission
takes action prior to the expiration of either the comment period or
the notice period, it will publish in the Federal Register a notice of
issuance. If the Commission makes a final no significant hazards
consideration determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
[[Page 42845]]
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 Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of the amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission no later
than 60 days from the date of publication of this notice. The petition
must be filed in accordance with the filing instructions in the
``Electronic Submissions (E-Filing)'' section of this document, and
should meet the requirements for petitions set forth in this section,
except that under 10 CFR 2.309(h)(2) a State, local governmental body,
or Federally-recognized Indian Tribe, or agency thereof does not need
to address the standing requirements in 10 CFR 2.309(d) if the facility
is located within its boundaries. Alternatively, a State, local
governmental body, Federally-recognized Indian Tribe, or agency thereof
may participate as a non-party under 10 CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
the discretion of the presiding officer, be permitted to make a limited
appearance pursuant to the provisions of 10 CFR 2.315(a). A person
making a limited appearance may make an oral or written statement of
his or her position on the issues but may not otherwise participate in
the proceeding. A limited appearance may be made at any session of the
hearing or at any prehearing conference, subject to the limits and
conditions as may be imposed by the presiding officer. Details
regarding the opportunity to make a limited appearance will be provided
by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for leave to intervene (petition), any
motion or other document filed in the proceeding prior to the
submission of a request for hearing or petition to intervene, and
documents filed by interested governmental entities that request to
participate under 10 CFR 2.315(c), must be filed in accordance with the
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR
46562, August 3, 2012). The E-Filing process requires participants to
submit and serve all adjudicatory documents over the internet, or in
some cases to mail copies on electronic storage media. Detailed
guidance on making electronic submissions may be found in the Guidance
for Electronic Submissions to the NRC and on the NRC's Web site 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
[[Page 42846]]
[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 Web site 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 Web site 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 Web site at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m.
and 6 p.m., Eastern Time, Monday through Friday, excluding government
holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
stating why there is good cause for not filing electronically and
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff. Participants filing adjudicatory documents in this
manner are responsible for serving the document on all other
participants. Filing is considered complete by first-class mail as of
the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service. A presiding officer, having granted an
exemption request from using E-Filing, may require a participant or
party to use E-Filing if the presiding officer subsequently determines
that the reason for granting the exemption from use of E-Filing no
longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the
Commission or the presiding officer. If you do not have an NRC-issued
digital ID certificate as described above, click cancel when the link
requests certificates and you will be automatically directed to the
NRC's electronic hearing dockets where you will be able to access any
publicly-available documents in a particular hearing docket.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or personal phone
numbers in their filings, unless an NRC regulation or other law
requires submission of such information. For example, in some
instances, individuals provide home addresses in order to demonstrate
proximity to a facility or site. With respect to copyrighted works,
except for limited excerpts that serve the purpose of the adjudicatory
filings and would constitute a Fair Use application, participants are
requested not to include copyrighted materials in their submission.
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina
Date of amendment request: June 29, 2017. A publicly available
version is in ADAMS under Accession No. ML17180A538.
Description of amendment request: The amendments would adopt
changes, with variations, based on the NRC-approved safety evaluation
of Technical Specifications Task Force (TSTF) Traveler TSTF-542,
Revision 2, ``Reactor Pressure Vessel Water Inventory Control,'' dated
December 20, 2016 (ADAMS Package Accession No. ML16343B066). The
revisions would replace existing technical specification (TS)
requirements related to ``operations with a potential for draining the
reactor vessel'' (OPDRVs) with new requirements on reactor pressure
vessel water inventory control (RPV WIC) to protect Safety Limit
2.1.1.3, which requires reactor vessel water level to be greater than
the top of active irradiated fuel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change replaces existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.1.3. Draining of RPV [reactor pressure vessel] water
inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e.,
refueling) is not an accident previously evaluated and, therefore,
replacing the existing TS controls to prevent or mitigate such an
event with a new set of controls has no effect on any accident
previously evaluated. RPV water inventory control in Mode 4 or Mode
5 is not an initiator of any accident previously evaluated. The
existing OPDRV controls or the proposed RPV WIC controls are not
mitigating actions assumed in any accident previously evaluated.
[[Page 42847]]
The proposed change reduces the probability of an unexpected
draining event, which is not a previously evaluated accident, by
imposing new requirements on the limiting time in which an
unexpected draining event could result in the reactor vessel water
level dropping to the top of the active fuel (TAF). These controls
require cognizance of the plant configuration and control of
configurations with unacceptably short drain times. These
requirements reduce the probability of an unexpected draining event.
The current TS requirements are only mitigating actions and impose
no requirements that reduce the probability of an unexpected
draining event. The proposed change reduces the consequences of an
unexpected draining event, which is not a previously evaluated
accident, by requiring an Emergency Core Cooling System (ECCS)
subsystem to be operable at all times in Modes 4 and 5. The current
TS requirements do not require any water injection systems, ECCS or
otherwise, to be operable in certain conditions in Mode 5. The
change in requirement from two ECCS subsystems to one ECCS subsystem
in Modes 4 and 5 does not significantly affect the consequences of
an unexpected draining event because the proposed Actions ensure
equipment is available within the limiting drain time that is as
capable of mitigating the event as the current requirements. The
proposed controls provide escalating compensatory measures to be
established as calculated drain times decrease, such as verification
of a second method of water injection and additional confirmations
that containment and/or filtration would be available if needed. The
proposed change reduces or eliminates some requirements that were
determined to be unnecessary to manage the consequences of an
unexpected draining event, such as automatic initiation of an ECCS
subsystem and control room ventilation. These changes do not affect
the consequences of any accident previously evaluated since a
draining event in Modes 4 and 5 is not a previously evaluated
accident and the requirements are not needed to adequately respond
to a draining event.
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 replaces existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.1.3. The proposed change will not alter the design
function of the equipment involved. Under the proposed change, some
systems that are currently required to be operable during OPDRVs
would be required to be available within the limiting drain time or
to be in service depending on the limiting drain time. Should those
systems be unable to be placed into service, the consequences are no
different than if those systems were unable to perform their
function under the current TS requirements. The event of concern
under the current requirements and the proposed change is an
unexpected draining event. The proposed change does not create new
failure mechanisms, malfunctions, or accident initiators that would
cause a draining event or a new or different kind of accident not
previously evaluated or included in the design and licensing bases.
Thus, based on the above, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change replaces existing TS requirements related to
OPDRVs with new requirements on RPV WIC. The current requirements do
not have a stated safety basis and no margin of safety is
established in the licensing basis. The safety basis for the new
requirements is to protect Safety Limit 2.1.1.3. New requirements
are added to determine the limiting time in which the RPV water
inventory could drain to the top of the fuel in the reactor vessel,
should an unexpected draining event occur. Plant configurations that
could result in lowering the RPV water level to the TAF within one
hour are now prohibited. New escalating compensatory measures based
on the limiting drain time replace the current controls. The
proposed TS establish a safety margin by providing defense-in-depth
to ensure that the Safety Limit is protected and to protect the
public health and safety. While some less restrictive requirements
are proposed for plant configurations with long calculated drain
times, the overall effect of the change is to improve plant safety
and to add safety margin.
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: Kathryn B. Nolan, Deputy General Counsel,
550 South Tryon Street, M/C DEC45A, Charlotte, NC 28202.
NRC Branch Chief: Undine Shoop.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant (PNP), Van Buren County, Michigan
Date of amendment request: July 27, 2017. A publicly-available
version is in ADAMS under Accession No. ML17208A428.
Description of amendment request: The proposed amendment would
revise certain staffing and training requirements, reports, programs,
and editorial changes in the Technical Specifications (TSs) Table of
Contents; Section 1.0, ``Use and Application''; and Section 5.0,
``Administrative Controls,'' that will no longer be applicable once PNP
is permanently defueled.
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 would not take effect until the PNP
Certified Fuel Handler Training and Retraining Program has been
approved by the NRC, and PNP has permanently ceased operation and
entered a permanently defueled condition. The proposed changes would
revise the PNP TS by modifying the definitions, in TS Section 1.0,
and administrative controls, in TS Section 5.0, to correspond to the
permanently defueled condition. Additionally, certain portions of
the administrative control sections are deleted because they are no
longer applicable to a permanently defueled facility.
The proposed deletion and modification of provisions of the
administrative controls do not directly affect the design of
structures, systems, and components (SSCs) necessary for safe
storage of spent nuclear fuel or the methods used for handling and
storage of such fuel in the spent fuel pool (SFP). The proposed
changes to the administrative controls are administrative in nature
and do not affect any accidents applicable to the safe management of
spent nuclear fuel or the permanently shutdown and defueled
condition of the reactor. Thus, the consequences of an accident
previously evaluated are not increased.
In a permanently defueled condition, the only credible accidents
are the fuel handling accident (FHA), the failure of tanks
containing radioactive liquids, and a spent fuel cask drop accident.
The probability of occurrence of previously evaluated accidents is
not increased, because extended operation in a permanently defueled
condition will be the only operation allowed. This mode of operation
is bounded by the existing analyses. Additionally, the occurrence of
postulated accidents associated with reactor operation are no longer
credible in a permanently defueled reactor. This significantly
reduces the scope of applicable accidents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment has no impact on facility systems,
structures, and components (SSCs) affecting the safe storage of
spent nuclear fuel, or on the methods of operation of such SSCs, or
on the handling
[[Page 42848]]
and storage of spent nuclear fuel itself. The proposed amendment
does not result in different or more adverse failure modes or
accidents than previously evaluated because the reactor will be
permanently shutdown and defueled, and PNP will no longer be
authorized to operate the reactor or retain or place fuel in the
reactor vessel.
The proposed amendment does not affect systems credited in the
PNP accident analysis for a[n] FHA, or for mitigating accident
releases from the failure of tanks containing radioactive liquids or
from a spent fuel cask drop. The proposed changes will continue to
require proper control and monitoring of safety significant
parameters and activities.
The proposed amendment does not result in any new mechanisms
that could damage the remaining relevant safety barriers that
support maintaining the plant in a permanently shutdown and defueled
condition (e.g., fuel cladding and SFP cooling). Since extended
operation in a defueled condition will be the only operation
allowed, and this condition is bounded by 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 accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment involves deleting and/or modifying
certain TS requirements once the PNP has been permanently shutdown
and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50
license for PNP will no longer authorize operation of the reactor or
emplacement or retention of fuel into the reactor vessel following
submittal of the certifications required by 10 CFR 50.82(a)(1).
Therefore, the occurrence of postulated accidents associated with
reactor operation are no longer credible.
The only remaining credible accidents are the fuel handling
accident (FHA), the failure of tanks containing radioactive liquids,
and a spent fuel cask drop accident. The proposed amendment does not
adversely affect the inputs or assumptions of any of the design
basis analyses that impact these analyzed conditions.
The proposed changes are limited to those portions of the TS
that are not related to the SSCs that are important to the safe
storage of spent nuclear fuel. The requirements that are proposed to
be revised or deleted from the PNP TS are not credited in the
existing 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 are no longer possible because the
reactor will be permanently shutdown and defueled, and PNP will no
longer be authorized to operate the reactor or retain or place fuel
in the reactor vessel.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White
Plains, NY 10601.
NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: July 18, 2017. A publicly-available
version is in ADAMS under Accession No. ML17199F854.
Description of amendment request: The proposed change would revise
the design value for the spent fuel storage pool in Technical
Specification (TS) 4.3.2, ``Drainage,'' to an appropriate value,
consistent with the original design basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
No physical changes to the facility will occur as a result of
this proposed amendment. The proposed changes will not alter the
physical design. The proposed change will revise the current TS
4.3.2 value for the SFP [spent fuel pool] level design to be
consistent with the original design basis value and the applicable
regulatory requirements. The proposed value will continue to ensure
that inadvertent draining of the SFP will not result in the
uncovering of spent fuel, as well as provide adequate shielding for
personnel protection.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not alter the physical design, safety
limits, or safety analysis assumptions associated with the operation
of the plant. Accordingly, the change does not introduce any new
accident initiators, nor does it reduce or adversely affect the
capabilities of any plant structure, system, or component to perform
their safety function. The proposed change will revise the current
TS 4.3.2 value for the SFP level design to be consistent with the
original design basis value and the applicable regulatory
requirements. The proposed value will continue to ensure that
inadvertent draining of the SFP will not result in the uncovering of
spent fuel, as well as provide adequate shielding for personnel
protection.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change conforms to NRC regulatory guidance
regarding the content of plant Technical Specifications. The
proposed change does not alter the physical design, safety limits,
or safety analysis assumptions associated with the operation of the
plant.
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 the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: David J. Wrona.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: July 19, 2017. A publicly-available
version is in ADAMS under Accession No. ML17200D096.
Description of amendment request: The amendments would replace
existing technical specification (TS) requirements related to
``operations with a potential for draining the reactor vessel''
(OPDRVs) with new requirements on reactor pressure vessel (RPV) water
inventory control (WIC) to protect Safety Limit 2.1.4. Safety Limit
2.1.4 requires RPV water level to be greater than the top of active
irradiated fuel. The proposed changes are based on Technical
Specifications Task Force (TSTF) Traveler TSTF-542, ``Reactor Pressure
Vessel Water Inventory Control,'' Revision 2 (ADAMS Package Accession
No. ML16250A231).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 42849]]
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes replace existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.4. Draining of RPV water inventory in OPERATIONAL
CONDITION 4 (i.e., cold shutdown) and OPERATIONAL CONDITION 5 (i.e.,
refueling), is not an accident previously evaluated and, therefore,
replacing the existing TS controls to prevent or mitigate such an
event with a new set of controls has no effect on any accident
previously evaluated. RPV water inventory control in OPERATIONAL
CONDITION 4 or 5 is not an initiator of any accident previously
evaluated. The existing OPDRV controls or the proposed RPV WIC
controls are not mitigating actions assumed in any accident
previously evaluated.
The proposed changes reduce the probability of an unexpected
draining event (which is not a previously evaluated accident) by
imposing new requirements on the limiting time in which an
unexpected draining event could result in the reactor vessel water
level dropping to the top of the active fuel (TAF). These controls
require cognizance of the plant configuration and control of
configurations with unacceptably short drain times. These
requirements reduce the probability of an unexpected draining event.
The current TS requirements are only mitigating actions and impose
no requirements that reduce the probability of an unexpected
draining event.
The proposed changes reduce the consequences of an unexpected
draining event (which is not a previously evaluated accident) by
requiring an Emergency Core Cooling System (ECCS) subsystem to be
operable at all times in OPERATIONAL CONDITIONS 4 and 5. The current
TS requirements do not require any water injection systems, ECCS or
otherwise, to be Operable in certain conditions in OPERATIONAL
CONDITION 5. The change in requirement from two ECCS subsystems to
one ECCS subsystem in OPERATIONAL CONDITIONS 4 and 5 does not
significantly affect the consequences of an unexpected draining
event because the proposed Actions ensure equipment is available
within the limiting drain time that is as capable of mitigating the
event as the current requirements. The proposed controls provide
escalating compensatory measures to be established as calculated
drain times decrease, such as verification of a second method of
water injection and additional confirmations that containment and/or
filtration would be available if needed.
The proposed changes reduce or eliminate some requirements that
were determined to be unnecessary to manage the consequences of an
unexpected draining event, such as automatic initiation of an ECCS
subsystem and control room ventilation. These changes do not affect
the consequences of any accident previously evaluated since a
draining event in OPERATIONAL CONDITIONS 4 and 5 is not a previously
evaluated accident and the requirements are not needed to adequately
respond to a draining event.
Therefore, the proposed changes do 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 replace existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.4. The proposed changes will not alter the design function
of the equipment involved. Under the proposed changes, some systems
that are currently required to be operable during OPDRVs would be
required to be available within the limiting drain time or to be in
service depending on the limiting drain time. Should those systems
be unable to be placed into service, the consequences are no
different than if those systems were unable to perform their
function under the current TS requirements.
The event of concern under the current requirements and the
proposed changes is an unexpected draining event. The proposed
changes do not create new failure mechanisms, malfunctions, or
accident initiators that would cause a draining event or a new or
different kind of accident not previously evaluated or included 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. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes replace existing TS requirements related to
OPDRVs with new requirements on RPV WIC. The current requirements do
not have a stated safety basis and no margin of safety is
established in the licensing basis. The safety basis for the new
requirements is to protect Safety Limit 2.1.4. New requirements are
added to determine the limiting time in which the RPV water
inventory could drain to the TAF in the reactor vessel should an
unexpected draining event occur. Plant configurations that could
result in lowering the RPV water level to the TAF within one hour
are now prohibited. New escalating compensatory measures based on
the limiting drain time replace the current controls. The proposed
TS establish a safety margin by providing defense-in-depth to ensure
that the Safety Limit is protected and to protect the public health
and safety. While some less restrictive requirements are proposed
for plant configurations with long calculated drain times, the
overall effect of the change is to improve plant safety and to add
safety margin.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: January 23, 2017, as supplemented by
letter dated July 3, 2017. Publicly-available versions are in ADAMS
under Accession Nos. ML17025A399 and ML17184A176, respectively.
Description of amendment request: The license amendment request was
originally noticed in the Federal Register on March 28, 2017 (82 FR
15383). The notice is being reissued in its entirety to include the
revised scope, description of the amendment request, and proposed no
significant hazards consideration determination. As a result of the
revised scope, updates to the ``Basis for proposed no significant
hazards consideration determination'' section of this notice are
delineated by brackets.
The amendments would modify the Technical Specifications (TSs) by
limiting the MODE of applicability for the Reactor Protection System
(RPS), Startup, and Operating Rate of Change of Power--High, functional
unit trip. Additionally, the proposed amendments add new Limiting
Condition for Operation (LCO) 3.0.5 and relatedly modifies LCO 3.0.1
and LCO 3.0.2, to provide for placing inoperable equipment under
administrative control for the purpose of conducting testing required
to demonstrate OPERABILITY.
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.
Limiting the MODE 1 applicability for RPS functional unit,
Startup and Operating Rate of Change of Power--High, to Power Range
Neutron Flux Power <=15% of RATED THERMAL POWER, is an
administrative change in nature and does not alter the manner in
which the functional unit is operated or maintained. The proposed
changes do not represent any physical
[[Page 42850]]
change to plant [structures, systems, and components (SSC(s))], or
to procedures established for plant operation. The subject RPS
functional unit is not an event initiator nor is it credited in the
mitigation of any event or credited in the [probabilistic risk
assessment (PRA)]. As such, the initial conditions associated with
accidents previously evaluated and plant systems credited for
mitigating the consequences of accidents previously evaluated remain
unchanged.
The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1
and Unit 2 TS and related modification to [LCO 3.0.1 and] LCO 3.0.2
is consistent with the guidance provided in NUREG-1432, Volume 1
[ADAMS Accession No. ML12102A165] (Reference 6.1 [of the amendment
request]) and thereby has been previously evaluated by the
Commission with a determination that the proposed change does not
involve a significant hazards consideration.
Therefore, facility operation in accordance with the proposed
license amendments would 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.
Limiting the MODE 1 applicability for the RPS functional unit,
Startup and Operating Rate of Change of Power--High, to Power Range
Neutron Flux Power <=15% of RATED THERMAL POWER, is an
administrative change in nature and does not involve the addition of
any plant equipment, methodology or analyses. The proposed changes
do not alter the design, configuration, or method of operation of
the subject RPS functional unit or of any other SSC. More
specifically, the proposed changes neither alter the power rate-of-
change trip function nor its ability to bypass and reset as
required. The subject RPS functional unit remains capable of
performing its design function.
The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1
and Unit 2 TS and related modification to [LCO 3.0.1 and] LCO 3.0.2
is consistent with the guidance provided in NUREG-1432, Volume 1
(Reference 6.1 [of the amendment request]) and thereby has been
previously evaluated by the Commission with a determination that the
proposed change does not involve a significant hazards
consideration.
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.
Limiting the MODE 1 applicability for RPS functional unit,
Startup and Operating Rate of Change of Power--High, to Power Range
Neutron Flux Power <=15% of RATED THERMAL POWER is an administrative
change in nature. The proposed changes neither involve changes to
any safety analyses assumptions, safety limits, or limiting safety
system settings nor do they adversely impact plant operating margins
or the reliability of equipment credited in safety analyses.
The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1
and Unit 2 TS and related modification to [LCO 3.0.1 and] LCO 3.0.2
is consistent with the guidance provided in NUREG-1432, Volume 1
(Reference 6.1 [of the amendment request]) and thereby has been
previously evaluated by the Commission with a determination that the
proposed change does not involve a significant hazards
consideration.
Therefore, operation of the facility in accordance with the
proposed amendment will not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
NRC Branch Chief: Undine Shoop.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
Date of amendment request: June 29, 2017. A publicly-available
version is in ADAMS under Accession No. ML17195A569.
Description of amendment request: The amendments would modify the
Technical Specification (TS) requirements for mode change limitations
in TS 3.0.4 and TS 4.0.4 based on Technical Specifications Tasks Force
(TSTF) Improved Standard Technical Specifications Change Traveler,
TSTF-359, Revision 9, ``Increase Flexibility in MODE Restraints''
(ADAMS Accession No. ML031190607).
The NRC issued a notice of opportunity for comment in the Federal
Register on August 2, 2002 (67 FR 50475), on possible amendments
concerning TSTF-359, including a model safety evaluation and model no
significant hazards consideration determination, using the consolidated
line item improvement process (CLIIP). Subsequently, on April 4, 2003,
the NRC published a Notice of Availability for TSTF-359, Revision 8, in
the Federal Register (68 FR 16579). That notice announced the
availability of this TS improvement through the CLIIP. The NRC
subsequently made two modifications in response to comments, as well as
one editorial change, which have been incorporated into TSTF-359,
Revision 9. The changes proposed in the licensee's submittal are,
therefore, based on TSTF-359, Revision 9.
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 allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS Action. Being
in a TS Action is not an initiator of any accident previously
evaluated. Therefore, the probability of an accident previously
evaluated is not significantly increased. The consequences of an
accident while relying on Actions as allowed by the proposed LCO
3.0.4 are no different than the consequences of an accident while
relying on Actions for other reasons, such as equipment
inoperability. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS while in a TS Action, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS while in a TS Action. The TS
allow operation of the plant without the full complement of
equipment through the Actions for not meeting the TS Limiting
Conditions for Operation (LCO). The risk associated with this
allowance is managed by the imposition of Actions that must be
performed within the prescribed completion times. The net effect of
being in a TS Action on the margin of safety is not considered
[[Page 42851]]
significant. The proposed change does not alter the required actions
or completion times of the TS. The proposed change allows TS Actions
to be entered and the associated required actions and completion
times to be used in new circumstances. This use is predicated upon
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing Actions
in similar circumstances without assessing and managing risk. The
net change to the margin of safety is insignificant. Therefore, this
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
NRC Branch Chief: Undine Shoop.
National Institute of Standards and Technology (NIST), Docket No. 50-
184, Center for Neutron Research Test Reactor, Montgomery County,
Maryland
Date of amendment request: March 2, 2017 (two letters), as
supplemented by letters dated March 29, 2017, and May 25, 2017.
Publicly-available versions are in ADAMS under Accession Nos.
ML17068A163, ML17068A164, ML17097A243, and ML17153A172, respectively.
Description of amendment request: The proposed amendment would
modify the NIST test reactor's technical specifications (TSs) to remove
limitations in the present version of the TSs that prohibit use of a
test procedure and to change the organizational chart in the TSs. In
addition, the proposed amendment would modify the NIST test reactor's
license to allow transfer of instrumentation calibration and testing
sources from the NIST's material license to the reactor license.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
No, the proposed amendment would not increase the probability or
consequences of an accident previously evaluated. The proposed
amendment removes conformance conflicts within the Technical
Specifications that would occur when operating the reactor as
permitted under TSs 2.2(4). The conflicts are removed from the TSs
by adding exception statements. When the reactor is operated under
the NRC approved conditions in TSs 2.2(4), steady state thermal
hydraulic analysis shows that operation at less than 500 kW
[kilowatt] with natural circulation results in a critical heat flux
ratio and onset of flow instability ratio greater than 2. Transient
analysis of reactivity insertion accidents shows that the fuel
cladding temperature remains far below the safety limit. The limit
of 10 kw was chosen since that was deemed adequate for any
operational situation requiring natural circulation operation, such
as testing of an unknown core loading.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
No, the proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The proposed amendment removes conformance conflicts
within the Technical Specifications that would occur when operating
the reactor as permitted under TSs 2.2(4). The conflicts are removed
from the TSs by adding exception statements. The accident analysis
was discussed in the document, NIST Response to NRC Request for
Information (TAC No. MD3410), August 19, 2008, ADAMS Accession
Number ML082890338. The request from the NRC was: ``. . . Provide
justification for 500 kW power operations under natural convection
flow by demonstrating that no credible accidents would result in
exceeding the safety limit. . . ,'' the following was the response
by NIST. ``This analysis shows that there is ample margin between
the maximum clad temperature in any credible accident and the safety
limit of 450 [deg]C [degrees Centigrade].'' The details of the
analysis are presented in the above reference.
The intent with this amendment is to allow, without apparent TSs
nonconformance, operation analyzed and evaluated by the NRC. This
will allow the use of testing similar to that which was performed in
the commissioning of NIST test reactor.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
No, the proposed amendment would not involve a significant
reduction in a margin of safety. This amendment will allow testing
when commissioning a core configuration that is unknown in the most
conservative manner appropriate. It removes apparent TS conflicts
that would force the licensee into situations that would be less
conservative and with less 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: Melissa J. Lieberman, Deputy Chief Counsel
for NIST, National Institute of Standards and Technology, 100 Bureau
Drive, Gaithersburg, MD 20899.
NRC Branch Chief: Alexander Adams, Jr.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
Date of amendment request: April 21, 2017, as supplemented by
letter dated August 15, 2017. Publicly-available versions are in ADAMS
under Accession Nos. ML17111A958, and ML17227A775, respectively.
Description of amendment request: The amendment request proposes to
depart from approved AP1000 Design Control Document (DCD) Tier 2
information (text, tables and figures) as incorporated into the Updated
Final Safety Analysis Report (UFSAR) as plant-specific DCD information,
and also proposes to depart from involved plant-specific Tier 1
information (and associated Combined License (COL) Appendix C
information). Specifically, the amendment request proposes changes to
COL Appendix C (and plant-specific Tier 1) Table 2.2.4-1 and Figure
2.2.4-1 to add two main feedwater thermal relief valves and two start-
up feedwater thermal relief valves. The proposed COL Appendix C (and
plant-specific DCD Tier 1) changes require additional changes to
corresponding Tier 2 information in UFSAR Chapters 3 and 10. Because
this proposed change requires a departure from Tier 1 information in
the Westinghouse Electric Company's AP1000 DCD, the licensee also
requested an exemption from the requirements of the Generic DCD Tier 1
in accordance with 10 CFR 52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The changes to Combined License (COL) Appendix C (and plant-
specific Tier 1) Table 2.2.4-1 and Figure 2.2.4-1, and associated
Updated Final Safety Analysis Report (USFAR) design information do
not adversely impact previously evaluated accidents. The addition of
the thermal relief valves to the feedwater lines does not adversely
impact the ability to isolate the main and startup feedwater lines
following a steam or
[[Page 42852]]
feedwater line break or steam generator tube rupture. The new
thermal relief valves are normally closed and required to open to
prevent potential overpressure conditions when ambient temperatures
increase in the area. Thermal relief valves added into the feedwater
lines operate mechanically and are not activated upon a new
engineered safety features (ESF) signal in response to design basis
accidents. Isolation capabilities of the main and startup feedwater
lines are not adversely affected as ESF signals are not changed. The
proposed change does not reduce the temperature of feedwater and
does not increase feedwater flow during any operational mode as main
feedwater and startup feedwater isolation and control valves are not
changed by this activity. Performance of overpressure relief
supports the safety-related functions of the isolation and control
valves in the main and startup feedwater lines when isolation is
required.
No safety-related structure, system, component (SSC) or function
is adversely affected by this change. The change does not involve an
interface with any SSC accident initiator or initiating sequence of
events, and thus, the probabilities of the accidents evaluated in
the plant-specific UFSAR are not affected. The proposed changes do
not involve a change to the predicted radiological releases due to
postulated accident conditions, thus, the consequences of the
accidents evaluated in the UFSAR are not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to COL Appendix C (and plant-specific Tier
1) Table 2.2.4-1 and Figure 2.2.4-1, and associated UFSAR design
information do not reduce the temperature of feedwater and do not
increase feedwater flow during any operational mode such that it
would result in a new or different kind of accident from accidents
previously evaluated. Conclusions of existing analyses are not
changed by this activity as existing feedwater isolation and control
valves functions are not changed.
The proposed changes to add thermal relief valves to the main
and startup feedwater lines do not adversely affect any safety-
related equipment, and do not add any new interfaces to safety-
related SSCs that adversely affect safety functions. No system or
design function or equipment qualification is adversely affected by
these changes as the changes do not modify any SSCs that prevent
safety functions from being performed by the existing main feedwater
and startup feedwater valves. The changes do not introduce a new
failure mode, malfunction or sequence of events that could adversely
affect safety or safety-related equipment as feedwater isolation
capabilities are not changed. Performance of overpressure relief
supports the safety-related functions of the isolation and control
valves in the main and startup feedwater lines when isolation is
required.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to COL Appendix C (and plant-specific Tier
1) Table 2.2.4-1 and Figure 2.2.4-1, and associated UFSAR design
information add thermal relief valves to the main feedwater and
startup feedwater lines. These valves are designed to the same codes
and standards as the existing piping to which they are connected,
including ASME Code Section III, Class C, seismic Category I. The
proposed changes do not affect any other safety-related equipment or
fission product barriers. The requested changes will not affect any
design code, function, design analysis, safety analysis input or
result, or design/safety margin. No safety analysis or design basis
acceptance limit/criterion is challenged or exceeded by the
requested changes. There are not any changes to operation of the
main feedwater and startup feedwater isolation and control valves
when isolation of the lines is required. Operation of the relief
valves supports isolation capabilities for the main and feedwater
isolation and control valves.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County,
Georgia
Date of amendment request: July 14, 2017. A publicly-available
version is in ADAMS under Accession No. ML17195B047.
Description of amendment request: The requested amendment proposes
to depart from Tier 2 information in the Updated Final Safety Analysis
Report (UFSAR) (which includes the plant-specific design control
document (DCD) Tier 2 information) and involves related changes to
plant-specific Tier 1 (and associated Combined License (COL) Appendix
C) information, and COL Appendix A Technical Specifications.
Specifically, the requested amendment proposes changes to add a second
normal residual heat removal system (RNS) suction relief valve in
parallel to the current RNS suction relief valve, with the necessary
piping changes. Additionally, a change is proposed to Tier 1 Figure
2.2.1-1, for penetration P19, to accurately depict the orientation of
the class break of containment isolation valve RNS-PL-V061.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with NRC staff's edits in
square brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to Combined License (COL) Appendix C (and
plant-specific Tier 1) Figures 2.2.1-1 and 2.3.6-1, Tables 2.3.6-1,
2.3.6-2 and 2.3.6-4, COL Appendix A, Technical Specification 3.4.14
and associated Updated Final Safety Analysis Report (UFSAR) design
information to identify a new normal residual heat removal system
(RNS) relief valve, RNS-PL-V020, do not adversely impact accidents
previously evaluated in the safety analysis. Transients that are
capable of overpressurizing the reactor coolant system (RCS) are
categorized as either mass or heat input transients. The relief
valves must be capable of passing flow greater than that required
for the limiting low-temperature overpressure protection (LTOP)
transients while maintaining RCS pressure less than the lowest
pressure represented by the pressure/temperature limit curve, 110%
of the design pressure of the RNS, or the acceptable RNS relief
valve inlet pressure. The restrictions added to COL Appendix A,
Technical Specification 3.4.14 to close chemical and volume control
system (CVS) makeup line containment isolation valve, CVS-PL-V091,
limit flow capacity when the RCS is aligned to the RNS to support
LTOP functions and provide reliable operation of the RNS relief
valves during mass and heat input transients. When CVS-PL-V091 is
open, the RCS is depressurized and an RCS vent of >=4.15 square
inches is established. Transient conditions including mass input and
heat input are not changed and probability of events is not
increased as the added RNS relief valve, RNS-PL-V020, supports LTOP
functions as required by Technical Specification 3.4.14. The current
3-inch RNS relief valve is sufficient to terminate identified
transients; however, the added 1-inch RNS relief valve reduces
chatter in the current valve during low flow scenarios.
Responses to mass and heat input transients are not changed as
LTOP functions to prevent overpressurization of the RCS are not
changed by this activity. The added RNS relief valve, RNS-PL-V020,
is designed in accordance with the same requirements as the current
RNS relief valve, RNS-PL-V021, but with a lower flow capacity and
functions at a lower setpoint pressure. Overpressure protection
provided by the RNS is not
[[Page 42853]]
changed. The change does not adversely impact the capability of the
RNS to protect the RCS from exceeding pressure and temperature
limits in accordance with 10 CFR 50, Appendix G or 110% of the
design pressure of the RNS. Changes in piping to accommodate the
addition of the valve and reduce inlet piping losses do not impact
the consequences or probabilities of previously evaluated accidents.
The class break correction for valve RNS-PL-V061, in COL Appendix C
(and plant-specific Tier 1) Figure 2.2.1-1 does not impact accidents
previously evaluated.
No safety-related structure, system, component (SSC) or function
is adversely affected by this change. The change does not involve an
interface with any structure, system, or component (SSC) accident
initiator or initiating sequence of events, and thus, the
probabilities of the accidents evaluated in the plant-specific UFSAR
are not affected. The proposed changes do not involve a change to
the predicted radiological releases due to postulated accident
conditions, thus, the consequences of the accidents evaluated in the
UFSAR are not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Conclusions of existing analyses are not changed by the proposed
change as LTOP functions provided by both the current and added RNS
relief valves continue to provide the assumed protection for LTOP
events. RCS pressure is maintained within limits by the use of both
RNS relief valves. The closure of CVS-PL-V091 limits flow and
reduces the impact of mass and heat input transients when RNS relief
valves are relied upon for overpressure protection.
The proposed change to add the smaller RNS relief valve, RNS-PL-
V020, does not adversely affect safety-related equipment, and does
not add any new interfaces to safety-related SSCs that adversely
affect safety functions. The added RNS relief valve, functions in
the same manner as the current RNS relief valve, but has a lower
capacity and lifts at a lower pressure. The added RNS relief valve
also discharges to the liquid radwaste system (WLS) containment
sump. No system or design function or equipment qualification is
adversely affected by these changes as the change does not modify
any SSCs that prevent safety functions from being performed by the
RNS and the current relief valve. The changes do not introduce a new
failure mode, malfunction or sequence of events that could adversely
affect safety or safety-related equipment. Piping changes to
accommodate the installation of the new valve do not create the
potential for a new or different kind of accident as the piping
requirements are consistent with those of the current relief valve,
and subject to the same pipe rupture evaluation requirements. LTOP
functions are not changed. The class break correction for valve RNS-
PL-V061 does not impact accident analysis or create a new or
different kind of accident as the function of the affected equipment
and piping is not changed.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not affect safety-related equipment or
fission product barriers. LTOP functions are not adversely impacted
as both the current and added RNS relief valves continue to provide
protection from overpressurization. The added RNS relief valve is
designed in accordance with [American Society of Mechanical
Engineers (ASME)] Code Section III, Class 2, requirements consistent
with the current RNS relief valve. Modified piping is constructed
consistent with current design requirements for RNS piping. The
addition of the valve adds safety margin in regards to transients as
the new valve lifts at a lower set pressure than the current valve,
causing flow rates to be lower through the RNS piping. Therefore,
margin of safety is not reduced. The requested changes will not
affect any design code, function, design analysis, safety analysis
input or result, or design/safety margin. No safety analysis or
design basis acceptance limit/criterion is challenged or exceeded by
the requested changes. Transient conditions, including mass input
and heat input, are not changed and margin of safety is not reduced
as the added RNS relief valve supports LTOP functions in the same
manner as the current RNS relief valve.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
Date of amendment request: May 31, 2017. A publicly-available
version is in ADAMS under Accession No. ML17151A296.
Description of amendment request: The requested amendment proposes
to depart from approved AP1000 Design Control Document (DCD) Tier 2
information (text, tables, and figures) as incorporated into the
Updated Final Safety Analysis Report (UFSAR) as plant-specific DCD
information, and from involved plant-specific Technical Specifications
as incorporated in Appendix A of the combined license. Specifically,
the proposed changes support the addition of chemicals necessary to
achieve proper reactor coolant system (RCS) water quality by allowing
an unborated water source through the chemical mixing tank to be
unisolated for <=1 hour for chemical addition to the pressurizer to be
performed with reactor coolant pumps (RCPs) not in operation. In order
to perform chemical addition to the pressurizer without the mixing
provided by forced reactor coolant system (RCS) flow, administrative
controls are established such that coolant introduced into the RCS is
at a boron concentration greater than or equal to that required to meet
the shutdown margin (SDM) boron concentration.
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.
Updated Final Safety Analysis Report (UFSAR) 15.4.6, Chemical
and Volume Control System Malfunction that Results in a Decrease in
the Boron Concentration in the Reactor Coolant, addresses
inadvertent boron dilution events. The principal means of positive
reactivity insertion to the core is the addition of unborated,
primary-grade water from the demineralized water transfer and
storage system (DWS) into the reactor coolant system (RCS) through
the reactor makeup portion of the chemical and volume control system
(CVS).
These events are primarily evaluated with one or more reactor
coolant pumps (RCPs) in operation providing adequate mixing. The
changes proposed by this amendment request do not involve operations
where the RCPs are in operation. Therefore, there is no increase in
the probability or consequences of inadvertent boron dilution events
with RCPs operating.
UFSAR Subsection 15.4.6 also describes that when a reactor
coolant pump is not operating, the demineralized water isolation
valves are closed and an uncontrolled boron dilution transient
cannot occur. The proposed amendment adds provisions to allow a
specific CVS unborated water source flow path to be opened through
the chemical mixing tank to the RCS pressurizer when RCPs are not in
operation for the purpose of chemical addition to the pressurizer.
The administrative control provisions proposed provide adequate
assurance that any injection to the RCS pressurizer would only occur
such that injected water is limited to
[[Page 42854]]
boron concentrations greater than the required concentrations to
meet the SDM. With no reduction in SDM, there would be no means of
positive reactivity insertion to the core leading to an adverse
reactivity event. As such, there is no significant increase in the
probability of a previously evaluated boron dilution event as a
result of this change.
Since the proposed change does not lead to any positive
reactivity insertion, there are no increased consequences of an
accident previously evaluated.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The administrative control provisions proposed provide adequate
assurance that any injection to the pressurizer would only occur
such that injected water is limited to boron concentrations greater
than the required concentrations to meet the SDM. With no reduction
in SDM, there would be no means of positive reactivity insertion to
the core leading to an adverse reactivity event. Failure modes
involving procedural controls and operator actions are considered in
evaluating inadvertent boron dilution events. The possibility of a
new or different kind of failure, malfunction, or sequence of events
has been evaluated with these proposed changes; events are precluded
with the proposed administrative controls and defense in depth
features inherent in the AP1000 design.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety is established by maintaining the required
SDM during shutdown activities. The proposed changes to the UFSAR
and Technical Specifications do not adversely affect the safety-
related functions of the RCS or CVS in maintaining adequate SDM.
Provisions are proposed for a specific CVS unborated water source
flow path to be opened through the chemical mixing tank to the RCS
pressurizer when RCPs are not in operation; however, this activity
is performed under administrative controls that preclude the
potential for a reduction in SDM.
The changes do not affect containment penetrations or any other
safety-related equipment or fission product barriers. The requested
changes will not affect any design code, function, design analysis,
safety analysis input or result, or design/safety margin. No safety
analysis or design basis acceptance limit/criterion is challenged or
exceeded by the requested changes. The existing design and operation
of the associated systems are adequate to preclude an inadvertent
boron dilution from occurring when RCPs are not in operation.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazard consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
Date of amendment request: July 28, 2017. A publicly-available
version is in ADAMS under Accession No. ML17209A185.
Description of amendment request: The requested amendment proposes
to depart from approved AP1000 Design Control Document (DCD) Tier 2
information as incorporated into the Updated Final Safety Analysis
Report (UFSAR) as plant-specific DCD information, and also proposes to
depart from involved plant-specific Tier 1 information and the
associated combined license (COL) Appendix C information. Specifically,
the amendment, if approved, would revise the COL documents mentioned
previously to reflect the proposed changes to update Reactor Coolant
System (RCS) requirements for the reactor vessel head vent (RVHV) mass
flow rate. Pursuant to the provisions of 10 CFR 52.63(b)(1), an
exemption from elements of the design as certified in the 10 CFR part
52, Appendix D, design certification rule is also requested for the
plant-specific DCD Tier 1 material departures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
UFSAR Subsections 15.2.7, 15.5.1, and 15.5.2 describe analyses
performed for an increase in reactor coolant inventory due to a loss
of normal feedwater flow, and for malfunctions of the chemical and
volume control system and the core makeup tanks. In each of these
evaluated accidents, it is assumed that the operators are alerted to
the event due to a high pressurizer water level and take subsequent
action to open the reactor vessel head vent valves. When the head
vent is opened, the pressurizer water level increase slows and
eventually decreases.
Changing the required mass flow rate from 8.2 lbm/sec at a
Reactor Coolant System (RCS) pressure of 1250 psia [pounds per
square inch absolute] to 9.0 lbm/sec [pounds mass per second] at an
RCS pressure of 2500 psia for the reactor vessel head vent (RVHV)
flow path does not change the probability of these events occurring.
The valves are used to mitigate the events. They are not an
initiator of these accidents, or any other accident previously
evaluated. Changing the required mass flow rate does not change the
consequences of these accidents. The proposed flow rate change is
made to be consistent with the latest AP1000 safety analysis. This
change does not lead to an increase in the probability of a loss of
coolant accident, nor does it cause the RVHV to exceed the
capability of the normal makeup system. The changes described above
continue to ensure the design is capable of providing adequate flow
rate for emergency letdown and the prevention of long term
pressurizer overfill.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes impact the acceptance criteria for RVHV
mass flow rate. The required mass flow rate is changed from 8.2 lbm/
sec at an RCS pressure of 1250 psia to 9.0 lbm/sec at an RCS
pressure of 2500 psia to align with the events evaluated in the
current safety analysis. The proposed changes do not result in a new
accident initiator and do not impact a current accident initiator.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes impact the acceptance criteria for RVHV
mass flow rate. The required mass flow rate is changed from 8.2 lbm/
sec at an RCS pressure of 1250 psia to 9.0 lbm/sec at an RCS
pressure of 2500 psia. The proposed changes are made to reflect the
updated AP1000 plant safety analysis; the changes are conservative
and bound the expected performance of the as-built equipment.
COL Appendix C (plant-specific Tier 1) is proposed to be updated
to reflect the new mass flow rate through the RVHV line and the
associated system pressure. COL Appendix C (plant-specific Tier 1)
is updated to reflect the latest safety analysis, which credits an
emergency letdown mass flow rate of 9.0 lbm/sec at an RCS pressure
of 2500
[[Page 42855]]
psia. At these conditions, long term pressurizer overfill is
prevented. RCS calculations show that the expected mass flow rate
through the emergency letdown path is 12.34 lbm/sec. Therefore, the
safety analysis calculation, and the corresponding mass flow rate
and RCS pressure values used in the proposed changes, is
conservative and bounded by the expected mass flow rate.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue, North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: July 31, 2017. A publicly-available
version is in ADAMS under Accession No. ML17212A842.
Description of amendment request: The amendment would revise the
staffing and staff augmentation times described in the South Texas
Project Emergency Plan. The proposed amendment would increase the
Emergency Response Organization (ERO) response times and would modify
minimum staffing functions and requirements of the ERO and Operations
Support Center staff. The changes also include formatting,
clarification, and editorial modifications.
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 has no effect on normal plant operation
or on any accident initiator or precursors and does not impact the
function of plant structures, systems, or components. The proposed
changes do not alter or prevent the ability of the Emergency
Response Organization to perform their intended functions to
mitigate the consequences of an accident or event.
Therefore, the proposed STPEGS [South Texas Project Electric
Generating Station] Emergency Plan change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not impact any accident analysis.
The change does not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed), a
change in the method of plant operation, or new operator actions.
The proposed change does not introduce failure modes that could
result in a new accident, and the change does not alter assumptions
made in the safety analysis. The proposed change revises the on-
shift staffing and staff augmentation response times in the STPEGS
Emergency Plan. The proposed changes do not alter or prevent the
ability of the Emergency Response Organization to perform their
intended functions to mitigate the consequences of an accident or
event.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed change is
associated with the STPEGS Emergency Plan staff and staff
augmentation and does not impact operation of the plant or its
response to transients or accidents. The change does not affect the
Technical Specifications. The proposed change does not involve a
change in the method of plant operation and no accident analyses
will be affected by the proposed change. Safety analysis acceptance
criteria are not affected by the proposed change. The revised STPEGS
Emergency Plan will continue to provide the necessary response staff
with the proposed change. Therefore, the proposed change is
determined to not adversely affect the ability to meet the
requirements of 10 CFR 50.54(q)(2), 10 CFR 50 Appendix E, or the
emergency planning standards described in 10 CFR 50.47(b).
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kym Harshaw, General Counsel, STP Nuclear
Operating Company, P.O. Box 289, Wadsworth, TX 77483.
NRC Branch Chief: Robert J. Pascarelli.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
[[Page 42856]]
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant, Units 1 and 2 (Brunswick), Brunswick County,
North Carolina
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1 (Harris), Wake County, North Carolina
Duke Energy Progress, LLC, Docket No. 50-261, H.B. Robinson Steam
Electric Plant Unit No. 2 (Robinson), Darlington County, South Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3 (Oconee), Oconee County,
South Carolina
Date of amendment request: April 29, 2016, as supplemented by
letters dated October 3, 2016, and January 16, 2017.
Brief description of amendments: The amendments (1) consolidated
the Emergency Operations Facilities (EOFs) for Brunswick, Harris, and
Robinson with the Duke Energy Progress, LLC (Duke Energy) corporate EOF
in Charlotte, North Carolina; (2) decreased the frequency for a
multisite drill at Oconee from once per 6 years to once per 8 years;
(3) allowed the multisite drill performance with sites other than the
Catawba Nuclear Station, McGuire Nuclear Station, or Oconee; (4)
changed the Brunswick, Harris, and Robinson augmentation times to be
consistent with those of the sites currently supported by the Duke
Energy corporate EOF; and (5) decreased the frequency of the
unannounced augmentation drill at Brunswick from twice per year to once
per year.
Date of issuance: August 21, 2017.
Effective date: As of the date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment Nos.: 279 and 307 for Brunswick, Units 1 and 2; 160 for
Harris, Unit 1; 254 for Robinson Unit No. 2; and 405, 407, and 406 for
Oconee, Units 1, 2, and 3. A publicly-available version is in ADAMS
under Accession No. ML17188A387; documents related to these amendments
are listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-71 and DPR-62 for
Brunswick, Units 1 and 2; NPF-63 for Harris, Unit 1; DPR-23 for
Robinson Unit No. 2; and DPR-38, DPR-47, and DPR-55 for Oconee, Units
1, 2, and 3: The amendments revised the emergency plans.
Date of initial notice in Federal Register: July 5, 2016 (81 FR
43650). The supplemental letters dated October 3, 2016, and January 16,
2017, provided additional information that expanded the scope of the
application as originally noticed and changed the NRC staff's original
proposed no significant hazards consideration determination as
published in the Federal Register. Accordingly, the NRC published a
second proposed no significant hazards consideration determination in
the Federal Register on February 14, 2017 (82 FR 10594). This notice
superseded the original notice in its entirety.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 21, 2017.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: July 28, 2016, as supplemented
by letters dated February 23, 2017, and June 21, 2017.
Brief description of amendment: The amendment revised the current
emergency action level scheme to one based on Nuclear Energy Institute
(NEI) guidance in NEI 99-01, Revision 6, ``Development of Emergency
Action Levels for Non-Passive Reactors'' (ADAMS Accession No.
ML12326A805). Revision 6 of NEI 99-01 was endorsed by the NRC in a
letter dated March 28, 2013.
Date of issuance: August 28, 2017.
Effective date: As of its date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment No.: 244. A publicly-available version is in ADAMS under
Accession No. ML17188A230; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-21: The amendment
revised the Operating License.
Date of initial notice in Federal Register: September 27, 2016 (81
FR 66305). The supplemental letters dated February 23, 2017, and June
21, 2017, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 28, 2017.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of application for amendments: October 7, 2016, as
supplemented by letter dated March 20, 2017.
Brief description of amendments: The amendments revised the Updated
Final Safety Analysis Report (UFSAR) to identify the TORMIS Computer
Code as the methodology used for assessing tornado-generated missile
protection of unprotected plant structures, systems and components
(SSCs) and to describe the results of the Byron Station site-specific
tornado hazard analysis.
Date of issuance: August 10, 2017.
Effective date: As of the date of issuance and shall be implemented
within 60 days of the date of issuance. The UFSAR changes shall be
filed with the NRC in the next periodic update to the UFSAR scheduled
for December 15, 2018.
Amendment Nos.: 199 for NPF-37 and 199 for NPF-66. A publicly-
available version is in ADAMS under Accession No. ML17188A155;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Facility Operating License Nos. NPF-37, and NPF-66: The amendments
revised the current licensing basis as described in the UFSAR.
Date of initial notice in Federal Register: December 6, 2016 (81 FR
87969). The March 20, 2017, supplement contained clarifying information
and did not change the scope of the proposed action or affect the NRC
staff's initial proposed finding of no significant hazards
consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 10, 2017.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station (Beaver Valley), Unit Nos. 1 and 2,
Beaver County, Pennsylvania
Date of amendment request: June 30, 2017.
Brief description of amendments: The amendments modified
requirements on control and shutdown rods, and rod and bank position
indication for Beaver Valley, Unit No. 2. The changes are consistent
with Technical Specifications Task Force (TSTF) Traveler TSTF-547,
Revision 1, ``Clarification of Rod
[[Page 42857]]
Position Requirements.'' Additional supporting changes to Beaver
Valley, Unit Nos. 1 and 2, Technical Specifications were also made.
Date of Issuance: August 16, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 299 (Unit No. 1) and 188 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML17221A280;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-66 and NPF-73:
Amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: July 11, 2017 (82 FR
32017).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 16, 2017.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: September 21, 2015, as supplemented by
letters dated November 13, December 15 (two letters), and December 18,
2015; February 16, March 8, March 9, March 24, March 28, April 4, April
5, April 14, April 22 (two letters), April 27, May 11, May 20 (two
letters), May 27, June 9, June 17, June 20, June 24, July 13 (two
letters), July 27, July 29 (two letters), August 3 (three letters),
September 12, September 21, September 23, October 13, October 28, and
October 31, 2016; and January 20, February 3, March 3, and June 12,
2017.
Brief description of amendments: The amendments revised Renewed
Facility Operating Licenses (RFOLs) and Technical Specifications (TSs)
to authorize an increase of maximum reactor core thermal power level
for Browns Ferry Nuclear Plant, Units 1, 2, and 3 to 3,952 megawatt
thermal (MWt). These license amendments represent an increase of
approximately 14.3 percent above the current licensed thermal power
level of 3,458 MWt, which is an increase of approximately 20 percent
above the original licensed thermal power level of 3,293 MWt. The NRC
considers the requested increase in power level to be an extended power
uprate.
Date of issuance: August 14, 2017.
Effective date: As of the date of issuance and shall be implemented
prior to startup from the refueling outages of fall 2018 (Unit 1),
spring 2019 (Unit 2), and spring 2018 (Unit 3).
Amendment Nos.: 299 (Unit 1), 323 (Unit 2), and 283 (Unit 3). A
publicly-available version is in ADAMS under Accession No. ML17032A120;
documents related to these amendments are listed in the Safety
Evaluation (SE) enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the RFOLs and TSs.
Date of initial notice in Federal Register: July 5, 2016 (81 FR
43666). The supplemental letters dated April 22 (two letters), April
27, May 11, May 20 (two letters), May 27, June 9, June 17, June 20,
June 24, July 13, (two letters); July 27, July 29 (two letters), August
3 (three letters), September 12, September 21, September 23, October
13, October 28, and October 31, 2016; and January 20, February 3, March
3, and June 12, 2017, provided additional information that clarified
the application, did not expand the scope of the application as
originally noticed, and did not change the 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 the SE dated August 14, 2017.
No significant hazards consideration comments received: Yes, refer
to Section 6.0, ``Public Comments,'' of the SE.
Wolf Creek Nuclear Operating Corporation (WCNOC), Docket No. 50-482,
Wolf Creek Generating Station (WCGS), Coffey County, Kansas
Date of amendment request: September 30, 2016, as supplemented by
letters dated March 16 and April 26, 2017.
Brief description of amendment: The amendment revised the emergency
action level (EAL) scheme used at WCGS. The currently approved EAL
scheme is based on Nuclear Management and Resources Council/National
Environmental Studies Project (NUMARC/NESP)-007, Revision 2,
``Methodology for Development of Emergency Action Levels,'' January
1992. The amendment allows WCNOC to adopt an EAL scheme, which is based
on the guidance established in Nuclear Energy Institute (NEI) 99-01,
Revision 6, ``Development of Emergency Action Levels for Non-Passive
Reactors,'' November 2012. Revision 6 of NEI 99-01 has been endorsed by
the NRC by letter dated March 28, 2013.
Date of issuance: August 28, 2017.
Effective date: As of its date of issuance and shall be implemented
by September 30, 2018.
Amendment No.: 218. A publicly-available version is in ADAMS under
Accession No. ML17166A409; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Operating License.
Date of initial notice in Federal Register: December 6, 2016 (81 FR
87974). The supplemental letters dated March 16 and April 26, 2017,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 28, 2017.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 30th day of August 2017.
For the Nuclear Regulatory Commission.
Eric J. Benner,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2017-19214 Filed 9-11-17; 8:45 am]
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