[Federal Register Volume 83, Number 30 (Tuesday, February 13, 2018)]
[Notices]
[Pages 6218-6242]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-02636]
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NUCLEAR REGULATORY COMMISSION
[NRC-2018-0021]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954,
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued, and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued, from January 13, 2018, to January 29, 2018. The
last biweekly notice was published on January 30, 2018.
DATES: Comments must be filed by March 15, 2018. A request for a
hearing must be filed by April 16, 2018.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0021. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
9127; email: [email protected]. For
[[Page 6219]]
technical questions, contact the individual listed in the FOR FURTHER
INFORMATION CONTACT section of this document.
Mail comments to: May Ma, Office of Administration, Mail
Stop: TWFN-3-D1, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-5411, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2018-0021, facility name, unit
number(s), plant docket number, application date, and subject when
contacting the NRC about the availability of information for this
action. You may obtain publicly-available information related to this
action by any of the following methods:
Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0021.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in this
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2018-0021, facility name, unit
number(s), plant docket number, application date, and subject in your
comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment
period such that failure to act in a timely way would result, for
example in derating or shutdown of the facility. If the Commission
takes action prior to the expiration of either the comment period or
the notice period, it will publish in the Federal Register a notice of
issuance. If the Commission makes a final no significant hazards
consideration determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and petition for leave to intervene
(petition) with respect to the action. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309. The NRC's regulations are accessible
electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert
[[Page 6220]]
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of the amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission no later
than 60 days from the date of publication of this notice. The petition
must be filed in accordance with the filing instructions in the
``Electronic Submissions (E-Filing)'' section of this document, and
should meet the requirements for petitions set forth in this section,
except that under 10 CFR 2.309(h)(2) a State, local governmental body,
or federally recognized Indian Tribe, or agency thereof does not need
to address the standing requirements in 10 CFR 2.309(d) if the facility
is located within its boundaries. Alternatively, a State, local
governmental body, Federally-recognized Indian Tribe, or agency thereof
may participate as a non-party under 10 CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
the discretion of the presiding officer, be permitted to make a limited
appearance pursuant to the provisions of 10 CFR 2.315(a). A person
making a limited appearance may make an oral or written statement of
his or her position on the issues but may not otherwise participate in
the proceeding. A limited appearance may be made at any session of the
hearing or at any prehearing conference, subject to the limits and
conditions as may be imposed by the presiding officer. Details
regarding the opportunity to make a limited appearance will be provided
by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for leave to intervene (petition), any
motion or other document filed in the proceeding prior to the
submission of a request for hearing or petition to intervene, and
documents filed by interested governmental entities that request to
participate under 10 CFR 2.315(c), must be filed in accordance with the
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR
46562, August 3, 2012). The E-Filing process requires participants to
submit and serve all adjudicatory documents over the internet, or in
some cases to mail copies on electronic storage media. Detailed
guidance on making electronic submissions may be found in the Guidance
for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit
paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to (1) request a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign submissions and access the E-Filing
system for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition or
other adjudicatory document (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on the NRC's public website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a
digital ID certificate and a docket has been created, the participant
can then submit adjudicatory documents. Submissions must be in Portable
Document Format (PDF). Additional guidance on PDF submissions is
available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the
time the document is submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must
[[Page 6221]]
apply for and receive a digital ID certificate before adjudicatory
documents are filed so that they can obtain access to the documents via
the E-Filing system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public website at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m.
and 6 p.m., Eastern Time, Monday through Friday, excluding government
holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
stating why there is good cause for not filing electronically and
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff. Participants filing adjudicatory documents in this
manner are responsible for serving the document on all other
participants. Filing is considered complete by first-class mail as of
the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service. A presiding officer, having granted an
exemption request from using E-Filing, may require a participant or
party to use E-Filing if the presiding officer subsequently determines
that the reason for granting the exemption from use of E-Filing no
longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the
Commission or the presiding officer. If you do not have an NRC-issued
digital ID certificate as described above, click cancel when the link
requests certificates and you will be automatically directed to the
NRC's electronic hearing dockets where you will be able to access any
publicly available documents in a particular hearing docket.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or personal phone
numbers in their filings, unless an NRC regulation or other law
requires submission of such information. For example, in some
instances, individuals provide home addresses in order to demonstrate
proximity to a facility or site. With respect to copyrighted works,
except for limited excerpts that serve the purpose of the adjudicatory
filings and would constitute a Fair Use application, participants are
requested not to include copyrighted materials in their submission.
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County, North
Carolina
Date of amendment request: November 15, 2017. A publicly available
version is in ADAMS under Accession No. ML17331A484.
Description of amendment request: The amendments would revise fire
protection license condition 2.B.(6) to allow, as a performance-based
method, certain currently-installed thermal insulation materials to be
retained and allow future use of these insulation materials in limited
applications subject to appropriate engineering reviews and controls,
as a deviation from the National Fire Protection Association Standard
805, Chapter 3, Section 3.3, Prevention.
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.
A fire hazards evaluation was performed for the areas of the
plant where the identified insulation materials are installed. The
fire hazards evaluation demonstrates that these materials do not
contribute appreciably to the spread of fire, nor represent a
secondary combustible beyond those currently analyzed in the Fire
Probabilistic Risk Analysis (FPRA) due to the limited applications
where these materials are installed. Therefore, it is concluded that
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The identified installations of the insulation materials were
evaluated against the fire scenarios supporting the FPRA. In all
instances, the supporting analyses and existing fire scenarios were
found to be bounding. Expanded zones of fire influence would not
fail additional FPRA targets, or there were no FPRA credited targets
in the area. Therefore, it is concluded that this change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The limited installations of the insulation materials do not
compromise post-fire safe shutdown capability as previously
designed, reviewed, and considered. Essential fire protection safety
functions are maintained and are capable of being performed. Because
the insulation materials do not compromise post-fire safe shutdown
capability as previously designed, reviewed, and considered, it is
concluded that this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel,
550 South Tryon Street, M/C DEC45A, Charlotte, NC 28202.
NRC Branch Chief: Undine Shoop.
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1 (HNP), Wake County, North Carolina
Date of amendment request: October 19, 2017, as supplemented by
letter dated January 11, 2018. Publicly-available versions are in ADAMS
under Accession Nos. ML17292B648 and ML18011A911, respectively.
Description of amendment request: The amendment would revise the
HNP Updated Final Safety Analysis Report (UFSAR) to incorporate the
Tornado Missile Risk Evaluator (TMRE) Methodology contained in Nuclear
Energy Institute (NEI) 17-02, Revision 1, ``Tornado Missile Risk (TMRE)
Industry Guidance Document,'' September 2017 (ADAMS Accession No.
ML17268A036).
[[Page 6222]]
This methodology can only be applied to discovered conditions where
tornado missile protection is not currently provided, and cannot be
used to avoid providing tornado missile protection in the plant
modification process.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with Nuclear Regulatory
Commission (NRC) staff edits in square brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not involve an increase in the
probability of an accident previously evaluated. The relevant
accident previously evaluated is a Design Basis Tornado impacting
the HNP site. The probability of a Design Basis Tornado is driven by
external factors and is not affected by the proposed amendment.
There are no changes required to any of the previously evaluated
accidents in the UFSAR.
The proposed amendment does not involve a significant increase
in the consequences of a Design Basis Tornado. [The methodology as
proposed does not alter any input assumptions or results of the
accident analyses. Instead, it reflects a methodology to more
realistically evaluate the probability of unacceptable consequences
of a Design Basis Tornado. As such, there is no significant increase
in the consequence of an accident previously evaluated. A similar
consideration would apply in the event additional non-conforming
conditions are discovered in the future.]
Therefore, the proposed amendment, for both the conditions
described herein and any future application of the methodology, 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, including any future use of the
methodology, will involve no physical changes to the existing plant,
so no new malfunctions could create the possibility of a new or
different kind of accident. The proposed amendment makes no changes
to conditions external to the plant that could create the
possibility of a new or different kind of accident. The proposed
change will not create the possibility of a new or different kind of
accident due to new accident precursors, failure mechanisms,
malfunctions, or accident initiators not considered in the design
and licensing bases. The existing UFSAR accident analysis will
continue to meet requirements for the scope and type of accidents
that require analysis.
Therefore, the proposed amendment, for both the conditions
described herein and any future application of the methodology, 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 does not exceed or alter any controlling
numerical value for a parameter established in the UFSAR or
elsewhere in the HNP licensing basis related to design basis or
safety limits. The change does not impact any UFSAR Chapter 6 or 15
Safety Analyses, and those analyses remain valid. The change
maintains diversity and redundancy as required by regulation or
credited in the UFSAR. The change does not reduce defense-in-depth
as described in the UFSAR.
Therefore, the proposed amendment, for both the conditions
described herein and any future application of the methodology, does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's modified analysis and,
based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Attorney for licensee: Lara Nichols, Deputy General Counsel, Duke
Energy Corporation, 550 South Tyron Street, Mail Code DEC45A,
Charlotte, NC 28202.
NRC Branch Chief: Douglas A. Broaddus.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 6, 2017. A publicly-available
version is in ADAMS under Accession No. ML17340B321.
Description of amendment request: The amendment would revise
Technical Specification 3/4.3.2 Table 4.3-2, ``Engineered Safety
Features Actuation System [ESFAS] Instrumentation Surveillance
Requirements.'' The amendment would remove from Note 3 of the table the
exemption from testing ESFAS relays K114, K305, and K313 at power.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will remove the Technical Specification
Table 4.3-2 Note 3 exemption for testing relays K305, K313, and K114
at power. The Technical Specification Table 4.3-2 Note 3 exemption
allowed the K305, K313, and K114 to not be tested during power
operation. The K305 and K313 relays are associated with the Main
Steam Isolation Signal (MSIS). The K114 relays are associated with
the Containment Spray Actuation Signal (CSAS). The removal of the
exemption from testing during power operation means the impacted
relays will be tested more frequently improving the ability to
identify failed components.
The removal of the Technical Specification Table 4.3-2 Note 3
exemption for testing relays K305, K313, and K114 means these relays
will be tested more frequently. This testing frequency will be
consistent with the other Technical Specification Table 4.3-2
subgroup relays that do not have an exemption. The probability of an
operator choosing the wrong subgroup relay during testing is no
different for this change as it is for the existing Technical
Specification Table 4.3-2 subgroup relays that are already tested on
this same frequency. Thus, there will be no significant increase in
the probability of an operator error causing an accident.
The change will also eliminate a potential single failure
vulnerability associated with MSIS (relays K305 and K313) and CSAS
(relay K114). The elimination of the single failure potential will
lower the probability of an accident due to the spurious actuation
of the MSIS or CSAS.
The change uses a parallel 2 out of 2 with second 2 out of 2 to
ensure no single failure of one actuation path would prevent the
other actuation path from completing its function. This ensures no
additional failure mode would prevent required equipment from
actuating and increasing accident consequences.
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 will remove the Technical Specification
Table 4.3-2 Note 3 exemption for testing relays K305, K313, and
K114. The K305, K313, and K114 relays are part of the Engineered
Safety Features Actuation System (ESFAS). The ESFAS is used for
accident mitigation but an inadvertent actuation could cause an
accident. The K305 and K313 relays are associated with the MSIS. The
K114 relays are associated with the CSAS. The potential failures of
the main steam isolation and containment spray systems have been
evaluated in the Waterford 3 Updated Final Safety Analysis Report
(UFSAR). The potential accidents are as follows:
Loss of External Load which could be caused by closure
of the Main Steam Isolation Valves (MSIVs) (UFSAR Section 15.2,
Decrease in Heat Removal by the Secondary System).
Loss of normal Feedwater Flow which could be caused by
the closure of the Main Feedwater Isolation Valves (UFSAR Section
15.2, Decrease in Heat Removal by the Secondary System).
[[Page 6223]]
Asymmetric Steam Generator Transient which could be
caused by the closure of one MSIV (UFSAR Section 15.9.1.1,
Asymmetric Steam Generator Transient).
Loss of component cooling to Reactor Coolant Pumps
(RCPs) which could be caused by the closure of the RCP Component
Coolant Water valve. This could lead to RCP seal assembly damage and
the possibility for a loss of coolant accident (UFSAR Section 15.6,
Decrease In Reactor Coolant System Inventory).
Inadvertent containment spray which could be caused by
actuation of one train of containment spray (UFSAR Section
6.2.1.1.3, Design Evaluation--Containment Pressure--Temperature
Analysis).
The removal of the exemption from testing during power operation
means the impacted relays will be tested more frequently thereby
improving the ability to identify failed components; however, they
will be tested at power. The ESFAS K305, K313, and K114 relay test
logic is designed to test the relays at power and not actuate the
end devices which could adversely impact the plant. Any failures
that could actuate plant equipment would continue to be bounded by
the existing UFSAR accidents; therefore, no new accident is being
created.
The ESFAS is used for accident mitigation. The removal of the
exemption from testing during power operation means the impacted
relays will be tested more frequently thereby improving the ability
to identify failed components. This lowers the possibility of the
ESFAS equipment not being available when needed. This also means
that with the ESFAS equipment available, this change does not create
the possibility of a 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 change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will remove the Technical Specification
Table 4.3-2 Note 3 exemption for testing relays K305, K313, and
K114. The removal of the exemption from testing during power
operation means the impacted relays will be tested more frequently
thereby improving the ability to identify failed components. The
more frequent testing will improve the margin of safety.
The change will also eliminate a potential single failure
vulnerability associated with MSIS (relays K305 and K313) and CSAS
(relay K114). The elimination of the single failure potential will
improve the margin of safety by reducing the potential of an
accident due to the spurious actuation of the MSIS or CSAS.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel,
Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East,
Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station (LSCS), Units 1 and 2, LaSalle County, Illinois
Date of amendment request: December 13, 2017. A publicly-available
version is in ADAMS under Accession No. ML17360A159.
Description of amendment request: The amendments would revise
technical specifications (TSs) to adopt Technical Specification Task
Force (TSTF)-542, Reactor Pressure Vessel Water Inventory Control (RPV
WIC).
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 replaces existing TS requirements related to
OPDRVs [operations with a potential for draining the reactor vessel]
with new requirements on RPV WIC water inventory control] that will
protect Safety Limit 2.1.1.3. Draining of RPV 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.
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 secondary
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 amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No.
The proposed change replaces existing TS [technical
specification] 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.
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 replaces existing TS requirements related to
OPDRVs with new requirements on RPV WIC. The current requirements do
not have a stated safety basis
[[Page 6224]]
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: 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 No. 50-220, Nine Mile Point
Nuclear Station, Unit 1, Oswego County, New York
Date of amendment request: December 15, 2017. A publicly available
version is in ADAMS under Accession No. ML17349A027.
Description of amendment request: The amendment would revise the
Nine Mile Point Nuclear Station, Unit 1, Technical Specifications (TSs)
by replacing existing requirements related to ``operations with a
potential for draining the reactor vessel'' (OPDRVs) with new
requirements on reactor pressure vessel water (RPV) inventory control
(WIC). The proposed changes are based on Technical Specifications Task
Force (TSTF) Improved Standard Technical Specifications Change Traveler
TSTF-542, Revision 2, ``Reactor Pressure Vessel Water Inventory
Control'' (ADAMS Accession No. ML16074A448).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes replace existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will ensure RPV water
level remains above -10 inches indicator scale. Draining of RPV
water inventory in the cold shutdown and refueling conditions is not
an accident previously evaluated; 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 the cold shutdown or refueling condition
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 -10 inches indicator scale. 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 a Core Spray subsystem to be operable at all times in the
cold shutdown and refueling conditions. The change in requirement
from two Core Spray subsystems to one Core Spray subsystem in the
cold shutdown or refueling conditions 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 a Core
Spray subsystem and control room ventilation. These changes do not
affect the consequences of any accident previously evaluated since a
draining event in the cold shutdown or refueling condition 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 maintain RPV water
level above -10 inches indicator scale. 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 change 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 accident 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 maintain RPV water level above -10 inches
indicator scale. 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 -10 inches indicator scale 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 maintain
RPV water level above -10 inches indicator scale 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.
[[Page 6225]]
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of amendment request: November 10, 2017. A publicly-available
version is in ADAMS under Accession No. ML17314A024.
Description of amendment request: The amendment would make changes
to the organization, staffing, and training requirements contained in
Section 6.0, ``Administrative Controls,'' of the Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Technical Specifications (TSs) and
define two new positions for Certified Fuel Handler and Non-Certified
Operator in Section 1.0, ``Definitions,'' to reflect the permanently
defueled condition.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes would not take effect until TMI-1 has
permanently ceased operation and certified a permanently defueled
condition. The proposed changes would revise the TMI-1 TS by
deleting or modifying certain portions of the TS administrative
controls described in Section 6.0 of the TS that are no longer
applicable to a permanently shutdown and defueled facility.
Additionally, the ``Certified Fuel Handler'' and ``Non-Certified
Operator'' would be added to Section 1.0 of the TS to define these
positions that are applicable to permanently shutdown and defueled
facility. These changes are administrative in nature.
The proposed changes do not involve any physical changes to
plant Structures, Systems, and Components (SSCs) or the manner in
which SSCs are operated, maintained, modified, tested, or inspected.
The proposed changes do not involve a change to any safety limits,
limiting safety system settings, limiting control settings, limiting
conditions for operation, surveillance requirements, or design
features.
The changes do not directly affect the design of SSCs necessary
for safe storage of spent irradiated fuel or the methods used for
handling and storage of such fuel in the Spent Fuel Pool (SFP). The
proposed changes are administrative in nature and do not affect any
accidents applicable to the safe management of spent irradiated fuel
or the permanently shutdown and defueled condition of the reactor.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence 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 the TS definitions and administrative
controls have no impact on facility plant Structures, Systems, and
Components (SSCs) affecting the safe storage of spent irradiated
fuel, or on the methods of operation of such SSCs, or on the actual
handling and storage of spent irradiated fuel. The proposed changes
do not result in different or more adverse failure modes or
accidents than previously evaluated because the reactor will be
permanently shutdown and defueled and TMI-1 will no longer be
authorized to operate the reactor.
The proposed changes do not affect systems credited in the
accident analyses at TMI-1. The proposed changes will continue to
require proper control and monitoring of safety significant
parameters and activities.
The proposed changes do not result in any new mechanisms that
could initiate damage to the remaining relevant safety barriers in
support of 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 therefore bounded by the existing analyses,
such a condition does not create the possibility of a new or
different kind of accident.
The proposed changes do not alter the protection system design,
create new failure modes, or change any modes of operation. The
proposed changes do not involve a physical alteration of the plant,
and no new or different kind of equipment will be installed.
Consequently, there are no new initiators that could result in a new
or different kind of accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes involve TS administrative controls once the
TMI-1 facility has been permanently shutdown and defueled. As
specified in 10 CFR 50.82(a)(2), the 10 CFR 50 license for TMI-1
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). As a result, the
occurrence of certain design basis postulated accidents are no
longer considered credible when the reactor is permanently defueled.
The proposed changes are limited to those portions of the
administrative TSs that are related to the safe storage and
maintenance of spent irradiated fuel. The proposed TS changes do not
affect plant design, hardware, system operation, or procedures for
accident mitigation systems. There is no change in the established
safety margins for these systems. The requirements that are proposed
to be added, revised and/or deleted from the TMI-1 TS are not
credited in the existing accident analysis for the applicable
postulated accidents; therefore, they do not contribute to the
margin of safety associated with the accident analysis. Certain
postulated design basis accidents (DBAs) involving the reactor are
no longer possible because the reactor will be permanently shutdown
and defueled and TMI-1 will no longer be authorized to operate the
reactor.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: December 20, 2017. A publicly-available
version is in ADAMS under Accession No. ML17355A019.
Description of amendment request: The amendment would revise
technical specification (TS) requirements related to direct current
(DC) electrical systems, specifically limiting conditions for operation
3.8.4, 3.8.5, and 3.8.6. The proposed amendment would also add a new
Battery and Monitoring Maintenance Program to TS Section 5.5,
``Programs and Manuals.'' The proposed changes are consistent with
Technical Specifications Task Force (TSTF) Traveler TSTF-500, Revision
2, ``DC Electrical Rewrite--Update to TSTF-360.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes restructure the Technical Specifications
(TS) for the direct current (DC) electrical power system and are
consistent with TSTF-500, Revision 2, ``DC Electrical Rewrite--
Update to TSTF-360.'' The proposed changes modify TS Actions
relating to battery and battery charger
[[Page 6226]]
inoperability. The DC electrical power system, including associated
battery chargers, is not an initiator of any accident sequence
analyzed in the Updated Safety Analysis Report (USAR). Rather, the
DC electrical power system supports equipment used to mitigate
accidents. The proposed changes to restructure TS and change
surveillances for batteries and chargers to incorporate the updates
included in TSTF-500, Revision 2, will maintain the same level of
equipment performance required for mitigating accidents assumed in
the USAR. Operation in accordance with the proposed TS would ensure
that the DC electrical power system is capable of performing its
specified safety function as described in the USAR. Therefore, the
mitigating functions supported by the DC electrical power system
will continue to provide the protection assumed by the analysis. The
relocation of preventive maintenance surveillances, and certain
operating limits and actions, to a licensee-controlled battery
monitoring and maintenance program will not challenge the ability of
the DC electrical power system to perform its design function.
Appropriate monitoring and maintenance that are consistent with
industry standards will continue to be performed. In addition, the
DC electrical power system is within the scope of 10 CFR 50.65,
``Requirements for monitoring the effectiveness of maintenance at
nuclear power plants,'' which will ensure the control of maintenance
activities associated with the DC electrical power system.
The integrity of fission product barriers, plant configuration,
and operating procedures as described in the USAR will not be
affected by the proposed changes. Therefore, the consequences of
previously analyzed accidents will not increase by implementing
these changes. 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 involve restructuring the TS for the DC
electrical power system. The DC electrical power system, including
associated battery chargers, is not an initiator to any accident
sequence analyzed in the USAR. Rather, the DC electrical power
system supports equipment used to mitigate accidents. The proposed
changes to restructure the TS and change surveillances for batteries
and chargers to incorporate the updates included in TSTF-500,
Revision 2, ``DC Electrical Rewrite--Update to TSTF-360,'' will
maintain the same level of equipment performance required for
mitigating accidents assumed in the USAR. Administrative and
mechanical controls are in place to ensure the design and operation
of the DC systems continues to meet the plant design basis described
in the USAR. Therefore, operation of the facility in accordance with
this proposed change will 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 through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The equipment margins will be maintained in
accordance with the plant-specific design bases as a result of the
proposed changes. The proposed changes will not adversely affect
operation of plant equipment. These changes will not result in a
change to the setpoints at which protective actions are initiated.
Sufficient DC capacity to support operation of mitigation equipment
is ensured. The changes associated with the new battery maintenance
and monitoring program will ensure that the station batteries are
maintained in a highly reliable manner. The equipment fed by the DC
electrical sources will continue to provide adequate power to
safety-related loads in accordance with analysis assumptions.
TS changes made in accordance with TSTF-500, Revision 2, ``DC
Electrical Rewrite--Update to TSTF-360,'' maintain the same level of
equipment performance stated in the USAR and the current TSs.
Therefore, the proposed changes do not involve a significant
reduction 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: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: David J. Wrona.
NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Manitowoc County, Wisconsin
Date of amendment request: August 31, 2017. A publicly-available
version is in ADAMS under Accession No. ML17243A201.
Description of amendment request: The proposed amendment would
modify the licensing basis, by the addition of a License Condition, to
allow for the implementation of the provisions of 10 CFR part 50.69,
``Risk-Informed Categorization and Treatment of Structures, Systems,
and Components (SSCs) for Nuclear Power Plants.'' The provisions of 10
CFR 50.69 allow adjustment of the scope of equipment subject to special
treatment controls (e.g., quality assurance, testing, inspection,
condition monitoring, assessment, and evaluation). For equipment
determined to be of low safety significance, alternative treatment
requirements can be implemented in accordance with this regulation. For
equipment determined to be of high safety significance, requirements
will not be changed or will be enhanced.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The process used to evaluate SSCs
for changes to NRC special treatment requirements and the use of
alternative requirements ensures the ability of the SSCs to perform
their design function. The potential change to special treatment
requirements does not change the design and operation of the SSCs.
As a result, the proposed change does not significantly affect any
initiators to accidents previously evaluated or the ability to
mitigate any accidents previously evaluated. The consequences of the
accidents previously evaluated are not affected because the
mitigation functions performed by the SSCs assumed in the safety
analysis are not being modified. The SSCs required to safely shut
down the reactor and maintain it in a safe shutdown condition
following an accident will continue to perform their design
functions.
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 will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The proposed change does not change
the functional requirements, configuration, or method of operation
of any SSC. Under the proposed change, no additional plant equipment
will be installed.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will permit the use of a risk-informed
categorization process to modify the scope of SSCs subject to NRC
special treatment requirements and to implement alternative
treatments per the regulations. The proposed change does not affect
any Safety Limits or operating parameters used to establish the
safety
[[Page 6227]]
margin. The safety margins included in analyses of accidents are not
affected by the proposed change. The regulation requires that there
be no significant effect on plant risk due to any change to the
special treatment requirements for SSCs and that the SSCs continue
to be capable of performing their design basis functions, as well as
to perform any beyond design basis functions consistent with the
categorization process and results.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Steven Hamrick, Managing Attorney--Nuclear
Florida Power & Light Company, LAW/WAS, 801 Pennsylvania Ave. NW #220,
Washington, DC 20004.
NRC Branch Chief: David J. Wrona.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
No. 1, Rockingham County, New Hampshire
Date of amendment request: December 1, 2017. A publicly-available
version is in ADAMS under Accession No. ML17339A428.
Description of amendment request: The amendment would revise
certain 18-month surveillance requirements previously performed while
shut down to be performed during power operations. The amendment would
also revise the administrative controls portion of the technical
specifications (TSs) to replace plant-specific titles with generic
titles and modify TSs 6.1.2, 6.2.2, 6.2.4, and Table 6.2-1 to be
consistent with NUREG-1431, ``Standard Technical Specifications,
Westinghouse Plants.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The technical specification (TS) surveillance requirements and
administrative controls associated with the proposed changes to the
TS are not initiators of any accidents previously evaluated, so the
probability of accidents previously evaluated is unaffected by the
proposed changes. The proposed change does not alter the design,
function, or operation of any plant structure, system, or component
(SSC). The capability of any operable TS-required SSC to perform its
specified safety function is not impacted by the proposed change. As
a result, the outcomes of accidents previously evaluated are
unaffected. Therefore, the proposed changes do not result in a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed change does not challenge the integrity or
performance of any safety-related systems. No plant equipment is
installed or removed, and the changes do not alter the design,
physical configuration, or method of operation of any plant SSC.
No physical changes are made to the plant, so no new causal
mechanisms are introduced. Therefore, the proposed changes to the TS
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not challenge the integrity or
performance of any safety-related systems. No plant equipment is
installed or removed, and the changes do not alter the design,
physical configuration, or method of operation of any plant SSC. No
physical changes are made to the plant, so no new causal mechanisms
are introduced. Therefore, the proposed changes to the TS do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The ability of any operable SSC to perform its designated safety
function is unaffected by the proposed changes. The proposed changes
do not alter any safety analyses assumptions, safety limits,
limiting safety system settings, or method of operating the plant.
The changes do not adversely affect plant operating margins or the
reliability of equipment credited in the safety analyses. With the
proposed change, each DC electrical train remains fully capable of
performing its safety function. Therefore, the proposed changes do
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Steve Hamrick, Acting Managing Attorney,
Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-
0420.
NRC Branch Chief: James G. Danna.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: July 28, 2017, as supplemented by
January 23, 2108, letter. Publicly-available versions are in ADAMS
under Accession No. ML17209A755, and ML18023A440, respectively.
Description of amendment request: The requested amendment proposes
changes to combined license Appendix A, plant-specific Technical
Specifications (TS) to make them consistent with the remainder of the
design, licensing basis, and the TS. The U.S. Nuclear Regulatory
Commission (NRC) staff previously noticed this amendment request in the
Federal Register on December 5, 2017 (82 FR 57473). However, due to
administrative errors that were inadvertently introduced, the NRC staff
is noticing this amendment request again.
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:
An evaluation to determine whether or not a significant hazards
consideration is involved with the proposed amendment was completed
by focusing on the three standards set forth in 10 CFR 50.92,
``Issuance of amendment,'' as discussed below. However, to provide
for ease of review, similar changes have been grouped into
categories to facilitate the significant hazards evaluations
required by 10 CFR 50.92. Generic significant hazards evaluations
are provided for the More Restrictive Changes and a specific
significant hazards evaluation for each Clarification or Less
Restrictive change. In regards to obvious editorial or
administrative changes (e.g., formatting, page rolls, punctuation,
etc.), an explicit discussion was not always provided, but is
considered to be addressed by the applicable generic significant
hazards evaluation.
Valuation for More Restrictive Changes
This generic category include changes that impose additional
requirements, decrease allowed outage times, increase the Frequency
of Surveillances, impose additional Surveillances, increase the
scope of Specifications to include additional plant equipment,
broaden the Applicability of Specifications, or provide additional
actions. These changes have been evaluated to not be detrimental to
plant safety.
More restrictive changes are proposed only when such changes are
consistent with the current Vogtle Electric Generating Plant, Units
3 and 4 (VEGP) licensing basis; the applicable VEGP safety analyses;
and good engineering practice such that the availability and
reliability of the affected equipment is not reduced.
Changes to the Technical Specifications (TS) requirements
categorized as More Restrictive are annotated with an ``MR'' in
Section 2 Discussion of Change (DOC). This affects TS changes L05
and L08.
Southern Nuclear Operating Company (SNC) proposes to amend the
VEGP TS. SNC
[[Page 6228]]
has evaluated each of the proposed TS changes identified as More
Restrictive in accordance with the criteria set forth in 10 CFR
50.92, ``Issuance of amendment,'' and has determined that the
proposed changes do not involve a significant hazards consideration.
This significant hazards consideration is applicable to each More
Restrictive change identified in Section 2.
The basis for the determination that the proposed changes do not
involve a significant hazards consideration is an evaluation of
these changes against each of the criteria in 10 CFR 50.92(c). The
criteria and conclusions of the evaluation are presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes provide more stringent TS requirements.
These more stringent requirements impose greater operational control
and conservatism, and as a result, do not result in operations that
significantly increase the probability of initiating an analyzed
event, and do not alter assumptions relative to mitigation of an
accident or transient event. The more restrictive requirements
continue to ensure process variables, structures, systems, and
components are maintained consistent with the safety analyses and
licensing basis. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or
changes in methods governing normal plant operation. The proposed
changes do impose different Technical Specification requirements.
However, these changes are consistent with the assumptions in the
safety analyses and licensing basis. Therefore, the proposed changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The imposition of more restrictive requirements either has no
effect on or increases a margin of plant safety. As provided in the
discussion of change, each change in this category is, by
definition, providing additional restrictions to enhance plant
safety. The changes maintain requirements within the safety analyses
and licensing basis. Therefore, the proposed changes do not involve
a significant reduction in a margin of safety.
Evaluation for Clarification Changes
This category consists of technical changes which revise
existing requirements such that the design and operation of a system
correctly reflects how the LCO is applied and how the Action or
Surveillance Requirement (SR) is carried out. This adds detail and
clarity to the Technical Specifications (TS) in operating the
applicable portions of the as designed and licensed plant.
Technical changes to the TS requirements categorized as
``Clarification'' are identified with an ``CL'' and an individual
number in Section 2 Discussion of Change (DOC).
Southern Nuclear Operating Company (SNC) proposes to amend the
Vogtle Electric Generating Plant, Units 3 and 4 (VEGP), Technical
Specifications. SNC has evaluated each of the proposed technical
changes identified as ``Clarification'' individually in accordance
with the criteria set forth in 10 CFR 50.92 and has determined that
the proposed changes do not involve a significant hazards
consideration.
The basis for the determination that the proposed changes do not
involve a significant hazards consideration is an evaluation of
these changes against each of the criteria in 10 CFR 50.92(c). The
criteria and conclusions of the evaluation are presented below.
L09 SNC proposes to amend TS 3.3.19 Diverse Actuation System
Manual Controls, Note (c) in Table 3.3.19-1 to ``With upper
internals in place.''
SNC has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment by focusing on
the three standards set forth in 10 CFR 50.92, ``Issuance of
amendment,'' as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant or a change in the methods governing normal plant
operations. The change applies to a Diverse Actuation System (DAS)
Manual Controls Mode 6 note for operability of the Automatic
Depressurization System (ADS) Stage 4 valves that involves revising
the note from reactor internals in place to upper internals in
place. In accordance with Limiting Condition for Operation (LCO)
3.4.13 ADS--Shutdown, Reactor Coolant System (RCS) Open
Applicability and TS 3.3.9, Engineered Safeguards Actuation System
Instrumentation, Function 7, the ADS Stage 4 valves are not required
to be operable in MODE 6 with the upper internals removed. However,
the reactor internals would still be present. The change involves
clarification of the note (with no change in required system or
device function), such that the appropriate configuration in Mode 6
would be in place and would not conflict with TS 3.4.13 or TS 3.3.9.
The revised note is not an initiator to any accident previously
evaluated. As a result, the probability of an accident previously
evaluated is not affected.
The consequences of an accident as a result of the revised note
and associated requirements and actions are no different than the
consequences of the same accident during the existing ones. As a
result, the consequences of an accident previously evaluated are not
affected by this change.
The proposed change does not alter or prevent the ability of
structures, systems, and components from performing their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits. The proposed change does not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change clarifies TS requirements for the DAS manual
control ADS Stage 4 valves such that they would be in agreement with
the requirements set forth for the ADS in RCS Shutdown Mode 6.
However, the proposed change does not involve a physical alteration
of the plant as described in the [Updated Final Safety Analysis
Report (UFSAR)]. No new equipment is being introduced, and equipment
is not being operated in a new or different manner. There are no
setpoints, at which protective or mitigative actions are initiated,
affected by this change. This change will not alter the manner in
which equipment operation is initiated, nor will the function
demands on credited equipment be changed. No change is being made to
the procedures relied upon to respond to an off-normal event as
described in the UFSAR as a result of this change. As such, no new
failure modes are being introduced. The change does not alter
assumptions made in the safety analysis and licensing basis.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not reduce a margin of safety because
it has no effect on any assumption of the safety analyses. While the
condition for the manual control of ADS Stage 4 actuation switches
in Mode 6 has changed, no action is made less restrictive than
currently approved for any associated actuated device inoperability.
As such, there is no significant reduction in a margin of safety.
L10 SNC proposes to amend current TS 3.5.4, ``Passive Residual
Heat Removal Heat Exchanger PRHR HX--Operating,'' Surveillance
Requirement (SR) 3.5.4.6 to: Verify both PRHR HX air operated outlet
valves stroke open and both IRWST gutter isolation valves stroke
closed.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant or a change in the methods governing normal plant
operations. The change involves correcting an existing surveillance
requirement (with no change in required system or device
[[Page 6229]]
function), such that the surveillance requirement complies with the
In-Containment Refueling Water Storage Tank (IRWST) Gutter Isolation
valve design and the Passive Residual Heat Removal (PRHR) Heat
Exchanger (HX) outlet isolation valve design. Revised surveillance
requirement presentation and compliance with TS actions are not an
initiator to any accident previously evaluated. As a result, the
probability of an accident previously evaluated is not affected.
The consequences of an accident as a result of the revised
surveillance requirement are no different than the consequences of
the same accident during the existing one. As a result, the
consequences of an accident previously evaluated are not affected by
this change.
The proposed change does not alter or prevent the ability of
structures, systems, and components from performing their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits. The proposed change does not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change clarifies the surveillance requirement such
that it agrees with the IRWST and PRHR HX isolation valve design.
However, the proposed change does not involve a physical alteration
of the plant as described in the UFSAR. No new equipment is being
introduced, and equipment is not being operated in a new or
different manner. There are no setpoints, at which protective or
mitigative actions are initiated, affected by this change. This
change will not alter the manner in which equipment operation is
initiated, nor will the function demands on credited equipment be
changed. No change is being made to the procedures relied upon to
respond to an off-normal event as described in the UFSAR as a result
of this change. As such, no new failure modes are being introduced.
The change does not alter assumptions made in the safety analysis
and licensing basis. Therefore, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not reduce a margin of safety because
it has no effect on any assumption of the safety analyses. While the
surveillance requirement has changed for the IRWST and PRHR HX
isolation valves, no action is made less restrictive than currently
approved for any associated actuated device inoperability. As such,
there is no significant reduction in a margin of safety.
10 CFR 50.92 Evaluations for Less Restrictive Changes
This category consists of technical changes which revise
existing requirements such that more restoration time is provided,
fewer compensatory measures are needed, unnecessary Surveillance
Requirements (SR) are deleted, or less restrictive surveillance
requirements are required. This would also include unnecessary
requirements which are deleted from the Technical Specifications
(TS) and other technical changes that do not fit a generic category.
These changes are evaluated individually.
Technical changes to the TS requirements categorized as ``Less
Restrictive'' are identified with an ``LR'' and an individual number
in Section 2 Discussion of Change (DOC).
Southern Nuclear Operating Company (SNC) proposes to amend the
Vogtle Electric Generating Plant, Units 3 and 4 (VEGP), Technical
Specifications. SNC has evaluated each of the proposed technical
changes identified as ``Less Restrictive'' individually in
accordance with the criteria set forth in 10 CFR 50.92 and has
determined that the proposed changes do not involve a significant
hazards consideration.
The basis for the determination that the proposed changes do not
involve a significant hazards consideration is an evaluation of
these changes against each of the criteria in 10 CFR 50.92(c). The
criteria and conclusions of the evaluation are presented below.
L01 SNC proposes to amend TS 1.1 Definitions--Shutdown Margin
by:
Changing Shutdown Margin (SDM) definition c. ``In MODE 2 with
keff<1.0 and MODES 3, 4, and 5, the worth of fully inserted Gray Rod
Cluster Assemblies (GRCAs) will be included in the SDM
calculation.'' to ``In MODE 2 with keff<1.0 and in MODES 3, 4, and
5, the worth of the verified fully inserted Gray Rod Cluster
Assemblies (GRCAs) which have passed the acceptance criteria for
GRCA bank worth measurements performed during startup physics
testing may be included in the SDM calculation.''
SNC has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment by focusing on
the three standards set forth in 10 CFR 50.92, ``Issuance of
amendment,'' as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant or a change in the methods governing normal plant
operations. The change proposed involves re-defining whether the
worth of the Gray Rod Cluster Assemblies (GRCAs) should be included
in MODE 2 with keff<1.0 and Modes 3, 4, and 5 when calculating the
appropriate Shutdown Margin (SDM). The worth of the GRCAs for MODE 2
with keff<1.0 and Modes 3, 4, and 5 is not credited in the safety
analyses as stated in the NRC Safety Evaluation Report (SER)
``Westinghouse Electric Company's Final Topical Report Safety
Evaluation For WCAP-16943, ``Enhanced Gray Rod Cluster Assembly
Rodlet Design,'' Section 3.0 for ensuring adequate SDM exists.
The change involves revising the existing SDM definition (with
no change in required system or device function), such that a more
appropriate, albeit less restrictive, definition would be applied
when calculating SDM. The revised SDM definition is not an initiator
of any accident previously evaluated. As a result, the probability
of an accident previously evaluated is not affected.
The consequences of an accident as a result of the revised
definition requirements are no different than the consequences of
the same accident during the existing one. As a result, the
consequences of an accident previously evaluated are not affected by
this change.
The proposed change does not alter or prevent the ability of
structures, systems, and components from performing their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits. The proposed change does not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed change does not involve a physical alteration of
the plant as described in the UFSAR. No new equipment is being
introduced, and equipment is not being operated in a new or
different manner. There are no setpoints, at which protective or
mitigative actions are initiated, affected by this change.
This change will not alter the manner in which equipment
operation is initiated, nor will the function demands on credited
equipment be changed. No change is being made to the procedures
relied upon to respond to an off-normal event as described in the
UFSAR as a result of this change. As such, no new failure modes are
being introduced. The change does not alter assumptions made in the
safety analysis and licensing basis. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change removes the requirement to include the worth
of the GRCAs when calculating the SDM because they are not credited
for SDM in MODE 2 with keff<1.0 and in MODES 3, 4, and 5. The
proposed change does not involve a physical alteration of the plant
as described in the UFSAR. No new equipment is being introduced, and
equipment is not being operated in a new or different manner. There
are no setpoints, at which protective or mitigative actions are
initiated, affected by this change. This change will not alter the
manner in which equipment operation is initiated, nor will the
function demands on credited equipment be changed. No change is
[[Page 6230]]
being made to the procedures relied upon to respond to an off-normal
event as described in the UFSAR as a result of this change. As such,
no new failure modes are being introduced. The change does not alter
assumptions made in the safety analysis and licensing basis.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not reduce a margin of safety because
it has no effect on any assumption of the safety analyses. While the
SDM calculation defined is made less restrictive by eliminating the
worth of the GRCAs in MODE 2 with keff<1.0 and in MODES 3, 4, and 5,
no credit is taken in the safety analyses for including their worth
as discussed in the NRC Safety Evaluation Report (SER)
``Westinghouse Electric Company's Final Topical Report Safety
Evaluation For WCAP-16943, ``Enhanced Gray Rod Cluster Assembly
Rodlet Design,'' Section 3.0. As such, there is no significant
reduction in a margin of safety.
L02 SNC proposes to amend TS 3.1.4 Rod Group Alignment Limits
by:
L02A. Change Limiting Condition of Operation (LCO) from ``All
shutdown and control rods shall be OPERABLE.'' to ``Each rod cluster
control assembly (RCCA) shall be OPERABLE.''
L02B. Change LCO AND statement from ``Individual indicated rod
positions shall be within 12 steps of their group step counter
demand position.'' to ``Individual indicated rod positions of each
RCCA and Gray Rod Cluster Assembly shall be within their 12 steps of
their group step counter demand position.''
L02C. Delete LCO 3.1.4 note.
L02D. Change Action Condition A from ``one or more rod(s)
inoperable.'' to where it now applies to ``One or more RCCA(s)
inoperable.''
L02E. Acronym defined in change to Required Action B.1
Completion Time from ``1 hour with the OPDMS not monitoring
parameters'' to ``1 hour with the On-Line Power Distribution
Monitoring System not monitoring parameters.''
L02F. Add Required Action B.2.3.1 where the Required Action will
be to ``Perform SR 3.2.5.1'' with a Completion Time of ``Once per 12
hours,'' OR perform B.2.3, which is renumbered as B.2.3.2.1.
L02G. Delete Required Action B.2.4 Note, and renumber the
Required Action to B.2.3.2.2.
L02H. Delete Required Action B.2.5 Note, and renumber the
Required Action to B.2.3.2.3.
L02I. Renumber Required Action B.2.6 to B.2.4.
L02J. Change SR 3.1.4.2 Note from ``Not applicable to GRCAs'' to
``Not applicable to Axial Offset (AO) Control Bank RCCAs.''
L02K. Change SR 3.1.4.2 from ``Verify rod freedom of movement
(trippability) by moving each rod not fully inserted in the core
>=10 steps in either direction.'' to ``Verify rod freedom of
movement (trippability) by moving each RCCA not fully inserted in
the core >=10 steps in either direction.''
L02L. Delete the Note to SR 3.1.4.3
L02M. Change SR 3.1.4.3 from ``Verify rod drop time of each rod
. . .'' to ``Verify rod drop time of each RCCA . . .''.
SNC has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment by focusing on
the three standards set forth in 10 CFR 50.92, ``Issuance of
amendment,'' as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant or a change in the methods governing normal plant
operations. The proposed changes involve revising the existing LCO
3.1.4 operability to be applicable to RCCAs with accompanying
changes in actions and surveillance requirements (with no change in
required system or device function), such that more appropriate,
albeit less restrictive, actions would be applied. The proposed
changes involve excluding the Gray Rod Cluster Assemblies (GRCAs) in
the LCO 3.1.4 Rod Group Alignments LCO since their trip reactivity
worth is not credited in the shutdown margin assessments in MODES 1
and 2, nor required by the design basis to be operable. Only the rod
cluster control assemblies (RCCAs) are required to be operable. The
maximum rod misalignment is an initial assumption in the safety
analyses that directly affects core power distributions and
assumption of available shutdown margin (SDM). Since the GRCAs do
not have a function to maintain the reactor sub-critical unless they
are fully inserted, and the reactor is shut down, operability does
not apply to GRCAs like it does for RCCAs in MODES 1 and 2. The
design basis function of the GRCAs when the reactor is critical does
not include a provision of trip reactivity.
The revised LCO, associated actions and surveillance
requirements are not an initiator to any accident previously
evaluated. As a result, the probability of an accident previously
evaluated is not affected.
The consequences of an accident as a result of the revised LCO
requirements, associated actions, and surveillance requirements are
no different than the consequences of the same accident during the
existing ones. As a result, the consequences of an accident
previously evaluated are not affected by this change.
The proposed change does not alter or prevent the ability of
structures, systems, and components from performing their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits. The proposed change does not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated. The proposed change does not
involve a physical alteration of the plant as described in the
UFSAR. No new equipment is being introduced, and equipment is not
being operated in a new or different manner. There are no setpoints,
at which protective or mitigative actions are initiated, affected by
this change.
This change will not alter the manner in which equipment
operation is initiated, nor will the function demands on credited
equipment be changed. No change is being made to the procedures
relied upon to respond to an off-normal event as described in the
UFSAR as a result of this change. As such, no new failure modes are
being introduced. The change does not alter assumptions made in the
safety analysis and licensing basis. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change involves revising the existing LCO 3.1.4
operability to be applicable to RCCAs with accompanying changes in
actions and surveillance requirements (with no change in required
system or device function), such that more appropriate, albeit less
restrictive, actions would be applied. The proposed change does not
involve a physical alteration of the plant as described in the
UFSAR. No new equipment is being introduced, and equipment is not
being operated in a new or different manner. There are no setpoints,
at which protective or mitigative actions are initiated, affected by
this change. This change will not alter the manner in which
equipment operation is initiated, nor will the function demands on
credited equipment be changed. No change is being made to the
procedures relied upon to respond to an off-normal event as
described in the UFSAR as a result of this change. As such, no new
failure modes are being introduced. The change does not alter
assumptions made in the safety analysis and licensing basis.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not reduce a margin of safety because
it has no effect on any assumption of the safety analyses. While the
LCO 3.1.4 for Rod Group Alignment Limits is made less restrictive by
eliminating the worth of the GRCAs in MODES 1 and 2 with keff >=1,
no credit is taken in the current design basis for including their
trip reactivity worth. As such, there is no significant reduction in
a margin of safety.
L03 SNC proposes to amend TS 3.1.6 Control Bank Insertion Limits
by changing Note 2. from ``This LCO is not applicable to Gray Rod
Cluster Assembly (GRCA) banks during GRCA bank sequence exchange
with On-Line Power Distribution Monitoring System monitoring
parameters'' to ``This LCO is not applicable to Gray Rod Cluster
Assembly (GRCA) banks for up to one hour during GRCA bank sequence
exchange.''
SNC has evaluated whether or not a significant hazards
consideration is involved
[[Page 6231]]
with the proposed amendment by focusing on the three standards set
forth in 10 CFR 50.92, ``Issuance of amendment,'' as discussed
below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant or a change in the methods governing normal plant
operations. The proposed change to TS 3.1.6 Control Bank Insertion
Limits Note 2 is to not require On Line Power Distribution System
(OPDMS) during GRCA bank sequence exchange and limit the LCO
applicability exception for one hour after the insertion or sequence
or overlap limits are violated due to the short duration of the
sequence exchange. The final mechanical shim (MSHIM) design
established that the GRCA bank sequence exchange will best be
accomplished by moving both banks at the same time. The entire
exchange sequence will only take a few minutes from the time banks
begin moving. During this short duration, OPDMS is not suited for
real time monitoring relative to the time constant for the vanadium
fixed incore detector system. The exchange transient may be
completed before the OPDMS detects a significant change in the core
radial power distribution. In addition, it is unlikely there would
be significant time to take corrective action in response to an
OPDMS alarm if one occurred during the exchange.
The revised LCO note exception is not an initiator of any
accident previously evaluated. As a result, the probability of an
accident previously evaluated is not affected.
The consequences of an accident as a result of the revised LCO
note exception is no different than the consequences of the same
accident during the existing one. As a result, the consequences of
an accident previously evaluated are not affected by this change.
The proposed change does not alter or prevent the ability of
structures, systems, and components from performing their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits. The proposed change does not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated. The proposed change does not
involve a physical alteration of the plant as described in the
UFSAR. No new equipment is being introduced, and equipment is not
being operated in a new or different manner. There are no setpoints,
at which protective or mitigative actions are initiated, affected by
this change.
This change will not alter the manner in which equipment
operation is initiated, nor will the function demands on credited
equipment be changed. No change is being made to the procedures
relied upon to respond to an off-normal event as described in the
UFSAR as a result of this change. As such, no new failure modes are
being introduced.
The change does not alter assumptions made in the safety
analysis and licensing basis. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant as described in the UFSAR. No new equipment is being
introduced, and equipment is not being operated in a new or
different manner. There are no setpoints, at which protective or
mitigative actions are initiated, affected by this change. This
change will not alter the manner in which equipment operation is
initiated, nor will the function demands on credited equipment be
changed. No change is being made to the procedures relied upon to
respond to an off-normal event as described in the UFSAR as a result
of this change. As such, no new failure modes are being introduced.
The change does not alter assumptions made in the safety analysis
and licensing basis. Therefore, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not reduce a margin of safety because
it has no effect on any assumption of the safety analyses. While the
proposed change to TS 3.1.6, Note 2 would not require OPDMS be
functional during GRCA bank sequence exchange for up to one hour,
OPDMS operability is still required by TS 3.2.5 On-Line Power
Distribution Monitoring System (OPDMS)--Monitored Parameters. As
such, there is no significant reduction in a margin of safety.
L04 SNC proposes to amend TS 3.1.7 Rod Position Indication by
deleting Required Action B.2 and renumbering the remaining Condition
B Required Actions.
SNC has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment by focusing on
the three standards set forth in 10 CFR 50.92, ``Issuance of
amendment,'' as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant or a change in the methods governing normal plant
operations. The proposed change is to remove Required Action B.2 for
monitoring and recording Reactor Coolant System (RCS) Tavg (with no
change in required system or device function), such that more
appropriate, albeit less restrictive, actions would be applied.
There are no safety benefits, no acceptance criteria or no actions
associated with any trends for recording Tavg. Monitoring Tavg
provides no power distribution information for unmonitored rods that
isn't already provided by complying with the existing requirements
of Condition A, and average coolant temperature provides no
indication of changes in shutdown margin.
The revised actions are not an initiator of any accident
previously evaluated. As a result, the probability of an accident
previously evaluated is not affected.
The consequences of an accident as a result of the revised LCO
requirements and actions are no different than the consequences of
the same accident during the existing ones. As a result, the
consequences of an accident previously evaluated are not affected by
this change.
The proposed change does not alter or prevent the ability of
structures, systems, and components from performing their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits. The proposed change does not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated. The proposed change does not
involve a physical alteration of the plant as described in the
UFSAR. No new equipment is being introduced, and equipment is not
being operated in a new or different manner. There are no setpoints,
at which protective or mitigative actions are initiated, affected by
this change.
This change will not alter the manner in which equipment
operation is initiated, nor will the function demands on credited
equipment be changed. No change is being made to the procedures
relied upon to respond to an off-normal event as described in the
UFSAR as a result of this change. As such, no new failure modes are
being introduced. The change does not alter assumptions made in the
safety analysis and licensing basis. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant as described in the UFSAR. No new equipment is being
introduced, and equipment is not being operated in a new or
different manner. There are no setpoints, at which protective or
mitigative actions are initiated, affected by this change. This
change will not alter the manner in which equipment operation is
initiated, nor will the function demands on credited equipment be
changed. No change is being made to the procedures relied upon to
respond to an off-normal event as described in the UFSAR as a result
of this change. As such, no new failure modes are being introduced.
The change does not alter assumptions made in the safety analysis
and licensing basis. Therefore, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
[[Page 6232]]
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not reduce a margin of safety because
it has no effect on any assumption of the safety analyses. While the
required actions of LCO 3.1.7 for Rod Position Indication are made
less restrictive by deletion of Action B.2 for monitoring Tavg,
monitoring Tavg provides no power distribution information for
unmonitored rods that aren't already provided by complying with the
existing requirements of Condition A. As such, there is no
significant reduction in a margin of safety.
L06 SNC proposes to amend TS 3.3.1 ``Reactor Trip System
Instrumentation,'' Table 3.3.1-1 FUNCTION 12, (page 2 of 2), Passive
Residual Heat Removal Actuation by deleting SR 3.3.1.9.
SNC has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment by focusing on
the three standards set forth in 10 CFR 50.92, ``Issuance of
amendment,'' as discussed 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 is to delete the Surveillance Requirement
(SR) 3.3.1.9 Channel Calibration for the passive residual heat
removal (PRHR) reactor trip system actuation. The PRHR reactor trip
actuation initiates a reactor trip in the event either of the
parallel PRHR discharge valves is not fully closed. The proper
adjustment of the valve position indication contact inputs to the
breaker position are verified by performance of SR 3.3.1.10 Trip
Actuating Device Operational Test (TADOT). The revised surveillance
requirements are not an initiator to any accident previously
evaluated. The reactor trip from PRHR actuation has not changed, and
the proper adjustment of the valve position indication contact
inputs continues to be addressed by current SR 3.3.1.10. As a
result, the probability of an accident previously evaluated is not
affected.
The consequences of an accident as a result of the revised
surveillance requirements are no different than the consequences of
the same accident during the existing ones. As a result, the
consequences of an accident previously evaluated are not affected by
this change.
The proposed change does not alter or prevent the ability of
structures, systems, and components from performing their intended
function to mitigate the consequences of an initiating event within
the assumed acceptance limits.
The proposed change does not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant as described in the UFSAR. No new equipment is being
introduced, and equipment is not being operated in a new or
different manner. There are no setpoints, at which protective or
mitigative actions are initiated, affected by this change.
This change will not alter the manner in which equipment
operation is initiated, nor will the function demands on credited
equipment be changed. No change is being made to the procedures
relied upon to respond to an off-normal event as described in the
UFSAR as a result of this change. As such, no new failure modes are
being introduced. The change does not alter assumptions made in the
safety analysis and licensing basis. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant as described in the UFSAR. No new equipment is being
introduced, and equipment is not being operated in a new or
different manner. There are no setpoints, at which protective or
mitigative actions are initiated, affected by this change. This
change will not alter the manner in which equipment operation is
initiated, nor will the function demands on credited equipment be
changed. No change is being made to the procedures relied upon to
respond to an off-normal event as described in the UFSAR as a result
of this change. As such, no new failure modes are being introduced.
The change does not alter assumptions made in the safety analysis
and licensing basis. Therefore, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not reduce a margin of safety because
it has no effect on any assumption of the safety analyses. While the
surveillance requirements have been made less restrictive, the
intent of the deleted surveillance requirement remains covered by an
existing surveillance requirement. As such, there is no significant
reduction in a margin of safety.
L07 SNC proposes to amend TS, Section 3.3.5, ``Reactor Trip
System Manual Actuation,'' Table 3.3.5-1 ``Reactor Trip System
Manual Actuation,'' Functions 1. Manual Reactor Trip, 2. Safeguards
Actuation Input from Engineered Safety Feature Actuation System--
Manual and 4. Core Makeup Tank Actuation Input from Engineered
Safety Feature Actuation System--Manual for Required Channels to 2
switches.
SNC has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment by focusing on
the three standards set forth in 10 CFR 50.92, ``Issuance of
amendment,'' as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes define the required channels operable for
manual reactor trip based upon the existing design. Required
channels operable are not an initiator to any accident previously
evaluated. As a result, the probability of an accident previously
evaluated is not affected. The consequences of an accident with
defined number of switches operable for manual reactor trip are no
different than the consequences of the same accident using the
existing required channels operable. As a result, the consequences
of an accident previously evaluated are not affected by this change.
The proposed change does not alter or prevent the ability of
structures, systems, and components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits.
The proposed change does not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
Further, the proposed change does not increase the types or
amounts of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures. The proposed change is consistent with the
safety analysis assumptions and resultant consequences.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant as described in the UFSAR. No new equipment is being
introduced, and equipment is not being operated in a new or
different manner. There are no setpoints, at which protective or
mitigative actions are initiated, affected by this change. This
change will not alter the manner in which equipment operation is
initiated, nor will the function demands on credited equipment be
changed. No change is being made to the procedures relied upon to
respond to an off-normal event as described in the UFSAR as a result
of this change. As such, no new failure modes are being introduced.
The change does not alter assumptions made in the safety analysis
and licensing basis. Therefore, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to define the required channels operable
consistent with the plant design does not alter the manner in which
safety limits, limiting safety system settings or limiting
conditions for operation are determined. The safety analysis
acceptance criteria are not affected by this
[[Page 6233]]
change. The proposed change will not result in plant operation in a
configuration outside of the design basis. Therefore, there is no
significant reduction in a margin of safety.
L11 SNC proposes to amend current TS 3.8.3, ``Inverters--
Operating,'' by changing:
1. Action Condition A. from ``One inverter inoperable.'' to
``One or two inverter(s) within one division inoperable.''
2. Second Note in Required Action A.1 from ``Restore inverter to
OPERABLE status.'' to ``Restore inverter(s) to OPERABLE status.''
SNC has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment by focusing on
the three standards set forth in 10 CFR 50.92, ``Issuance of
amendment,'' as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant or a change in the methods governing normal plant
operations. The proposed changes to action conditions to explicitly
define an inverter division that contains two inoperable inverters
is not an accident initiator nor do they impact mitigation of the
consequences of any accident. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed change does not involve a physical alteration of
the plant as described in the UFSAR and does not alter the method of
operation or control of equipment as described in the UFSAR. The
current assumptions in the safety analysis regarding accident
initiators and mitigation of accidents are unaffected by this
change. Plant equipment remains capable of performing mitigative
functions assumed by the accident analysis. No additional failure
modes or mechanisms are being introduced and the likelihood of
previously analyzed failures remains unchanged.
The integrity of fission product barriers, plant configuration,
and operating procedures as described in the UFSAR will not be
affected by this change. Therefore, the consequences of previously
analyzed accidents will not increase because of this change.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to action conditions to explicitly define
an inverter division that contains two inoperable inverters does not
involve a physical alteration of the plant as described in the
UFSAR. No new equipment is being introduced, and equipment is not
being operated in a new or different manner. There are no setpoints,
at which protective or mitigative actions are initiated, that are
affected by this change. This change will not alter the manner in
which equipment operation is initiated, nor will the function
demands on credited equipment be changed. No change is being made to
the procedures relied upon to respond to an off-normal event as
described in the UFSAR as a result of this change. As such, no new
failure modes are being introduced. The change does not alter
assumptions made in the safety analysis and licensing basis.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed change will not reduce a margin of
safety because it has no such effect on any assumption of the safety
analyses.
Operation in accordance with the proposed TS operability ensures
that the plant response to analyzed events continues to provide the
margins of safety assumed by the analysis. Appropriate monitoring
and maintenance, consistent with industry standards, will continue
to be performed. Therefore, there is no significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: 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: November 17, 2017. A publicly-available
version is in ADAMS under Accession No. ML17321B080.
Description of amendment request: The amendment request proposes
changes to combined license (COL) License Condition and changes to the
Updated Final Safety Analysis Report (UFSAR) in the form of departures
from the incorporated plant-specific Design Control Document Tier 2*
and associated Tier 2 information. Specifically, this amendment request
involves a change to COL License Condition requirements regarding the
Natural Circulation (first plant test) using the steam generators and
the Passive Residual Heat Removal Heat Exchanger (first plant test). A
COL License Condition is proposed to be revised to include an exception
that would allow the requirements of a Technical Specification to be
suspended during performance of the Natural Circulation (first plant
test) using the steam generators. In addition, a revised Passive
Residual Heat Removal Heat Exchanger (first plant test) is proposed to
be performed as part of the Power Ascension Testing requirements
instead of as part of the Initial Criticality and Low-Power Testing
requirements as currently specified in a COL License Condition.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not adversely affect the operation of
any systems or equipment that initiate an analyzed accident or alter
any structures, systems, and components (SSC) accident initiator or
initiating sequence of events. The proposed changes do not adversely
affect the ability of the steam generators, applicable reactor trip
functions, and the passive residual heat removal heat exchanger to
perform the required safety function to remove core decay heat
during forced and natural circulation when necessary to prevent
exceeding the reactor core and the reactor coolant system design
limits, and do not adversely affect the probability of inadvertent
operation or failure of the passive residual heat removal heat
exchanger. The proposed changes do not result in any increase in
probability of an analyzed accident occurring, and maintain the
initial conditions and operating limits required by the accident
analysis, and the analyses of normal operation and anticipated
operational occurrences, so that the reactor core and the reactor
coolant system design limits are not exceeded for events requiring
emergency core decay heat removal.
Therefore, the requested amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not affect the operation of any systems
or equipment that may initiate a new or different kind of accident,
or alter any SSC such that a new accident initiator or initiating
sequence of events is created. The proposed changes do not adversely
affect the ability of the steam generators, applicable reactor trip
functions, and the passive residual heat removal heat exchanger to
perform the required safety function to remove core decay heat
during forced and natural circulation when necessary to prevent
exceeding the reactor
[[Page 6234]]
core and the reactor coolant system design limits, and do not
adversely affect the probability of inadvertent operation or failure
of the passive residual heat removal heat exchanger. The proposed
changes do not result in the possibility of an accident occurring,
and maintain the initial conditions and operating limits required by
the accident analysis, and the analyses of normal operation and
anticipated operational occurrences, so that the reactor core and
the reactor coolant system design limits are not exceeded for events
requiring emergency core decay heat removal.
These proposed changes do not adversely affect any other SSC
design functions or methods of operation in a manner that results in
a new failure mode, malfunction, or sequence of events that affect
safety related or nonsafety related equipment. Therefore, this
activity does not allow for a new fission product release path,
result in a new fission product barrier failure mode, or create a
new sequence of events that results in significant fuel cladding
failures.
Therefore, the requested amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes maintain existing safety margins through
continued application of the existing requirements of the UFSAR. The
proposed changes maintain the initial conditions and operating
limits required by the accident analysis, and the analyses of normal
operation and anticipated operational occurrences, so that the
reactor core and the reactor coolant system design limits are not
exceeded for events requiring emergency core decay heat removal.
Therefore, the proposed changes satisfy the same safety functions in
accordance with the same requirements as stated in the UFSAR. These
changes do not adversely affect any design code, function, design
analysis, safety analysis input or result, or design/safety margin.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed changes, and no margin of
safety is reduced. Therefore, the requested amendment does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC 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: December 21, 2017. A publicly-available
version is in ADAMS under Accession No. ML17355A416.
Description of amendment request: The requested amendment proposes
changes to combined license License Condition 2.D by adding a new
condition to address the Tier 2* change process. The proposal also
requests exemptions from the requirements of 10 CFR part 52, Appendix
D, Paragraphs II.F, VIII.B.6.b, and VIII.B.6.c.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes would add a license condition that would
allow use of the Tier 2 departure evaluation process for Tier 2*
departures, where such departures would not have more than a minimal
impact to safety. Changing the criteria by which departures from
Tier 2* information are evaluated to determine if NRC approval is
required does not affect the plant itself. Changing these criteria
does not affect prevention and mitigation of abnormal events, e.g.,
accidents, anticipated operational occurrences, earthquakes, floods
and turbine missiles, or their safety or design analyses. No safety-
related structure, system, component (SSC) or function is adversely
affected. The changes neither involve nor interface with any SSC
accident initiator or initiating sequence of events, and thus, the
probabilities of the accidents evaluated in the Updated Final Safety
Analysis Report (UFSAR) are not affected. Because the changes do not
involve any safety related SSC or function used to mitigate an
accident, 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 would add a license condition that would
allow use of the Tier 2 departure evaluation process for Tier 2*
departures, where such departures would not have more than a minimal
impact to safety. The changes do not affect the safety-related
equipment itself, nor do they affect equipment which, if it failed,
could initiate an accident or a failure of a fission product
barrier. No analysis is adversely affected. No system or design
function or equipment qualification is adversely affected by the
changes. This activity does not allow for a new fission product
release path, result in a new fission product barrier failure mode,
or create a new sequence of events that would result in significant
fuel cladding failures. In addition, the changes do not result in a
new failure mode, malfunction or sequence of events that could
affect safety or safety-related equipment.
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 would add a license condition that would
allow use of the Tier 2 departure evaluation process for Tier 2*
departures, where such departures would not have more than a minimal
impact to safety.
The proposed change is not a modification, addition to, or
removal of any plant SSCs. Furthermore, the proposed amendment is
not a change to procedures or method of control of the nuclear plant
or any plant SSCs. The only impact of this activity is the
application of the current Tier 2 departure evaluation process to
Tier 2* departures.
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.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: December 21, 2017. A publicly-available
version is in ADAMS under Accession No. ML17355A177.
Description of amendment request: The proposed amendment
establishes Conditions, Required Actions, and Completion Times in the
Technical Specification (TS) 3.75 for the Condition where one steam
supply to the turbine driven Auxiliary Feedwater (AFW) pump is
inoperable concurrent with an inoperable motor driven AFW train. In
addition, this amendment establishes changes to the TS, that establish
specific Actions: (1) For when two motor driven AFW trains are
inoperable at the same time and; (2) for when the turbine
[[Page 6235]]
driven AFW train is inoperable either (a) due solely to one inoperable
steam supply, or (b) due to reasons other than one inoperable steam
supply. The licensee stated that the change is consistent with NRC-
approved Technical Specification Task Force (TSTF) Traveler, TSTF-412,
Revision 3, ``Provide Actions for One Steam Supply to Turbine Driven
AFW/EFW Pump Inoperable.'' (ADAMS Accession No. ML070100363).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 10.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, by referencing the environmental evaluation included in
the model safety evaluation published in the Federal Register on July
17, 2007 (72 FR 39089), which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The Auxiliary/Emergency Feedwater (AFW/EFW) System is not an
initiator of any design basis accident or event, and therefore the
proposed changes do not increase the probability of any accident
previously evaluated. The proposed changes to address the condition
of one or two motor driven AFW/EFW trains inoperable and the turbine
driven AFW/EFW train inoperable due to one steam supply inoperable
do not change the response of the plant to any accidents.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not adversely
affect the ability of structures, systems, and components (SSCs) to
perform their intended safety function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of any accident previously evaluated.
Further, the proposed changes do not increase the types and amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures.
Therefore, the changes do not involve a significant increase in
the probability or consequences of any accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not result in a change in the manner in
which the AFW/EFW System provides plant protection. The AFW/EFW
System will continue to supply water to the steam generators to
remove decay heat and other residual heat by delivering at least the
minimum required flow rate to the steam generators. There are no
design changes associated with the proposed changes. The changes to
the Conditions and Required Actions do not change any existing
accident scenarios, nor create any new or different accident
scenarios.
The changes do not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. In addition,
the changes do not impose any new or different requirements or
eliminate any existing requirements. The changes do not alter
assumptions made in the safety analysis. The proposed changes are
consistent with the safety analysis assumptions and current plant
operating practice.
Therefore, the changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by these changes. The proposed changes will not
result in plant operation in a configuration outside the design
basis.
Therefore, it is concluded that 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: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center
Parkway, Birmingham, AL 35242.
NRC Branch Chief: Michael T. Markley.
U.S. Department of Transportation, Maritime Administration, Docket No.
50-238, Nuclear Ship Savannah, Baltimore, Maryland
Date of amendment request: October 31, 2017. A publicly-available
version is in ADAMS under Accession No. ML17307A036.
Description of amendment request: The amendment would revise the
license to remove a condition that prevents dismantling and disposing
of the facility without prior approval of the Commission.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes are administrative and do not involve
modification of any plant equipment or affect basic plant operation.
The NSS's reactor is not operational and the level of
radioactivity in the NSS has significantly decreased from the levels
that existed when the 1976 Possession-only License was issued. No
aspect of any of the proposed changes is an initiator of any
accident previously evaluated. Consequently, the probability of an
accident previously evaluated is not significantly increased.
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.
Both of the proposed changes are administrative and do not
involve physical alteration of plant equipment that was not
previously allowed by Technical Specifications. These proposed
changes do not change the method by which any safety-related system
performs its function. As such, no new or different types of
equipment will be installed, and the basic operation of installed
equipment is unchanged. The methods governing plant operation and
testing remain consistent with current safety analysis assumptions.
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.
Both of the proposed changes are administrative in nature. No
margins of safety exist that are relevant to the ship's defueled and
partially dismantled reactor. As such, there are no changes being
made to safety analysis assumptions, safety limits or safety system
settings that would adversely affect plant safety as a result of the
proposed changes. The proposed changes involve revising the language
of the license to clearly state previously approved changes, and to
delete archaic requirements.
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.
[[Page 6236]]
Advisor for licensee: Erhard W. Koehler, U.S. Department of
Transportation, Maritime Administration, 1200 New Jersey Ave. SE,
Washington, DC 20590.
NRC Branch Chief: Bruce A. Watson, CHP.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: November 7, 2017. A publicly-available
version is in ADAMS under Accession No. ML17317A464.
Description of amendment request: The amendments would revise the
Surry Power Station (Surry), Units 1 and 2, Facility Operating License
Numbers DPR-32 and DPR-37, respectively, in the form of new License
Conditions, and Technical Specification (TS) 3.16, ``Emergency Power
System,'' to allow a one-time extension of the Allowed Outage Time
(AOT) in TS 3.16 Action B.2 from 7 days to 21 days. The requested
temporary 21-day AOT is needed to replace Reserve Station Service
Transformer C (RSST-C) and associated cabling during the Surry Unit 2
fall 2018 refueling outage. The existing RSST-C is original plant
equipment and is reaching the end of its dependable service life.
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 license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No.
The proposed change adds a footnote to TS 3.16, ``Emergency
Power System,'' to allow a one-time extension of the AOT in TS 3.16
Action B.2 from 7 days to 21 days to facilitate the replacement of
RSST-C and associated cabling.
During the temporary 21-day AOT, the station emergency buses
will continue to be fed from redundant, separate, reliable offsite
sources that are capable of supporting the emergency loads under
worst-case conditions considering a single failure.
There are two (2) emergency buses for each unit: Buses 1H and 1J
(Unit 1), and Buses 2H and 2J (Unit 2). While RSST-C is being
replaced during the temporary 21-day AOT, Buses 1J and 2H will
continue to be energized from a designated primary offsite source,
System (Switchyard) Reserve Transformer (SRT) 4. Buses 1H and 2J
will be energized from Main Step-up Transformer 2, which is the Unit
2 designated dependable alternate source.
In both configurations Transfer Bus F is fed through two, in
series, transformers.
The normal configuration feeds Transfer Bus F from the
230 kV switchyard via two (2) transformers (SRT-2 and RSST-C) and
two (2) breakers. The 230 kV switchyard is connected to ten (10)
offsite circuits.
The temporary 21-day AOT configuration feeds Transfer
Bus F from the 500 kV switch yard via two (2) transformers (Main
Step-up Transformer 2 and Station Service Transformer 2C) and three
(3) breakers. The 500 kV switchyard is connected to 3 offsite
circuits.
A risk assessment has been performed for the temporary 21-day
AOT configuration. The assessment concluded that the probability of
a loss of offsite power for the proposed configuration is very low.
Thus, the proposed change does not significantly increase the
probability of an accident previously evaluated because: (a) The
emergency buses continue to be feed from redundant, separate,
reliable offsite sources and (b) the effect of the proposed
configuration on the probability of a loss of offsite power is very
low.
There is no increase in the consequences of an accident because
the emergency buses continue to be fed from redundant, separate,
reliable offsite circuits and the onsite power sources (i.e., the
Emergency Diesel Generators) are unaffected.
The consequences of both a Loss of Offsite Power (LOOP) and a
Station Blackout (SBO) have been evaluated in the UFSAR. There is no
change in the station responses to a LOOP or an SBO as a result of
the extended AOT because RSST-C is not included in designated
equipment used in the LOOP and SBO coping strategies.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed configuration does not result in a change in the
manner in which the electrical distribution subsystems downstream of
RSST-C provide plant protection. During the temporary AOT (21 days
total), the only change is to substitute the reliable Unit 2
designated dependable alternate source for a primary offsite power
source for Emergency Buses 1H and 2J. Other sources of offsite and
onsite power are unaffected, and other aspects of the offsite and
onsite power supplies are unchanged and unaffected.
There are no changes to the other RSSTs or to the supporting
systems operating characteristics or conditions.
There is no change in the station responses to a LOOP or an SBO
because RSST-C is not included in the designated equipment used in
the LOOP and SSO coping strategies.
Therefore, the proposed change does 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 TS change does not affect the acceptance criteria
for any analyzed event, nor is there a change to any safety limit.
The proposed TS change does not affect any structures, systems or
components or their capability to perform their intended functions.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. Neither the safety analyses nor the safety
analysis acceptance criteria are affected by this change. The
proposed change will not result in plant operation in a
configuration outside the current design basis as the design basis
includes use of the Unit 2 dependable alternate source. The proposed
TS change allows use of the Unit 2 dependable alternate power source
as the primary source for buses 1H and 2J for a period of up to 21
days. The margin of safety is maintained by maintaining the
capability to supply Emergency Buses 1H and 2J with a redundant,
separate, reliable offsite power source, and maintaining the onsite
power sources in their design basis configuration. Therefore, the
proposed change does not involve a significant reduction in margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
NRC Branch Chief: Michael T. Markley.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these
[[Page 6237]]
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.
Arizona Public Service Company, et al. (APS), Docket Nos. STN 50-528,
STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station
(PVNGS), Units 1, 2, and 3, Maricopa County, Arizona
Date of amendment: July 1, 2016, as supplemented by letters dated
June 2 and December 15, 2017.
Description of amendment request: The amendments revised the
Technical Specifications for PVNGS, Units 1, 2, and 3, to support the
implementation of next generation fuel (NGF). In addition to the
license amendment request, APS requested an exemption from certain
requirements of 10 CFR 50.46, ``Acceptance criteria for emergency core
cooling systems [ECCS] for light-water nuclear power reactors,'' and 10
CFR part 50, Appendix K, ``ECCS Evaluation Models,'' to allow the use
of Optimized ZIRLOTM as a fuel rod cladding material.
The proposed change would allow for the implementation of NGF
including the use of Optimized ZIRLOTM fuel rod cladding
material. The NGF assemblies contain advanced features to enhance fuel
reliability, thermal performance, and fuel cycle economics.
Date of issuance: January 23, 2018.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 205 (Unit 1), 205 (Unit 2), and 205 (Unit 3). A
publicly-available version is in ADAMS under Accession No. ML17319A107;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74:
The amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: October 4, 2016 (81 FR
68469). The supplemental letters dated June 2 and December 15, 2017,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 23, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station (CPS), Unit No. 1, DeWitt County, Illinois
Date of amendment request: May 4, 2017.
Brief description of amendment: The amendment deletes a
Surveillance Requirement Note associated with TS 3.5.1, ``ECCS
[Emergency Core Cooling System]--Operating,'' TS 3.5.2, ``ECCS--
Shutdown,'' and TS 3.6.1.7, ``Residual Heat Removal (RHR) Containment
Spray System,'' to more appropriately reflect the RHR system design,
and ensure the RHR system operation is consistent with the technical
specification (TS) Limiting Condition for Operation (LCO) requirements.
The amendment also adds a Note in the LCO for TS 3.5.1, TS 3.5.2, TS
3.6.1.7, TS 3.6.1.9, ``Feedwater Leakage Control System,'' and TS
3.6.2.3, ``Residual Heat Removal (RHR) Suppression Pool Cooling,'' to
clarify that one of the required subsystems in each of the affected TS
sections listed above may be inoperable during alignment and operation
of the RHR system for Shutdown Cooling (i.e., decay heat removal) with
the reactor steam dome pressure less than the RHR cut in permissive
value.
Date of issuance: January 22, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No(s): 215. A publicly-available version is in ADAMS
under Accession No. ML17324A354; documents related to this amendment
are listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-62: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 5, 2017 (82 FR
31095).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 22, 2018.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of amendment request: December 23, 2013, as supplemented by
letters dated February 14, 2017; April 27, May 27, June 26, November 6,
and December 21, 2015; February 24 and May 12, 2016; and January 30,
April 21, June 23, August 22, October 25, and November 29, 2017.
Brief description of amendments: The amendments revised the Beaver
Valley, Unit Nos. 1 and 2, Renewed Facility Operating Licenses (RFOLs)
to establish and maintain a risk-informed, performance-based fire
protection program in accordance with the requirements of 10 CFR
50.48(c).
Date of issuance: January 22, 2018.
Effective date: As of the date of issuance and shall be implemented
consistent with paragraph 2.C.(5) for Unit No. 1, and paragraph 2.F for
Unit No. 2, of the RFOLs.
Amendment Nos.: 301 (Unit No. 1) and 190 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML17291A081;
documents related to these amendments are listed in the safety
evaluation enclosed with the amendments.
RFOL Nos. DPR-66 and NPF-73: Amendments revised the RFOLs.
Date of initial notice in Federal Register: September 9, 2014 (79
FR 53458). The supplemental letters dated April 27, May 27, June 26,
November 6, and December 21, 2015; February 24 and May 12, 2016; and
January 30, April 21, June 23, August 22, October 25, and November 29,
2017, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated January 22, 2018.
No significant hazards consideration comments received: No.
[[Page 6238]]
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: June 20, 2017.
Brief description of amendment: The amendment revised technical
specifications (TSs) to delete the list of diesel generator critical
trips from TS Surveillance Requirement (SR) 3.8.1.13 and clarify that
the purpose of the SR is to verify that the non-critical automatic
trips are bypassed.
Date of issuance: January 18, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 179. A publicly-available version is in ADAMS under
Accession No. ML17325B690; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-58: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 15, 2017 (82 FR
38718).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 18, 2018.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit 1 (FCS), Washington County, Nebraska
Date of amendment request: June 9, 2017, as supplemented by letter
dated September 21, 2017.
Brief description of amendment: The amendment deleted Technical
Specification (TS) 2.8.3(6), ``Spent Fuel Cask Loading,'' and
associated Figure 2-11, ``Limiting Burnup Criteria for Acceptable
Storage in Spent Fuel Cask''; TS 3.2, Table 3-5, item 24, ``Spent Fuel
Cask Loading''; TS 4.3.1.3, Design Features associated with spent fuel
casks; and portions of TS 3.2, Table 3-4, item 5, footnote (4) on boron
concentration associated with cask loading.
Date of issuance: January 19, 2018.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 296. A publicly-available version is in ADAMS under
Accession No. ML17338A172; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-40: The amendment
revised the renewed facility operating license and TSs.
Date of initial notice in Federal Register: August 15, 2017 (82 FR
38718).
The supplemental letter dated September 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 NRC staff 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 January 19, 2018.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC and Exelon Generation Company, LLC, Docket Nos. 50-272
and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: March 6, 2017, as supplemented by
letters dated May 4, 2017, and September 14, 2017.
Brief description of amendments: The amendments revised Technical
Specification 3.6.2.3, ``Containment Cooling System,'' to extend the
containment fan coil unit allowed outage time from 7 days to 14 days
for one or two inoperable containment fan coil units.
Date of issuance: January 18, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 321 (Unit 1) and 302 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17349A108; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-70 and DPR-75: The
amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: June 6, 2017 (82 FR
26136). The supplemental letter dated September 14, 2017, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 18, 2018.
No significant hazards consideration comment received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia
Date of amendment request: April 7, 2017.
Brief description of amendments: The amendment revises the
requirements of Technical Specification (TS) 3.6.4.1, ``Secondary
Containment,'' associated with Surveillance Requirement (SR) 3.6.4.1.2.
Specifically, SR 3.6.4.1.2 verifies that one secondary containment
access door in each access opening is closed. The amendments would
allow for brief, inadvertent, simultaneous opening of redundant
secondary containment access doors during normal entry and exit
conditions.
Date of issuance: January 22, 2018.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1-289, Unit 2-234. A publicly-available
version is in ADAMS under Accession No. ML17355A440; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: August 29, 2017 (82 FR
41070).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 22, 2018.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: May 31, 2017, and supplemented by letter
dated November 16, 2017.
Description of amendment: The amendment authorizes changes to the
VEGP Units 3 and 4 Updated Final Safety Analysis Report in the form of
departures from the plant-specific Design Control Document Tier 2
information and involves changes to the administrative controls for
unborated water flow paths to the reactor coolant
[[Page 6239]]
system to support chemical additions during periods when the reactor
coolant pumps are not in operation. These proposed changes are
reflected in Appendix A, Technical Specifications.
Date of issuance: January 9, 2018.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 105 (Unit 3) and 104 (Unit 4). A publicly-available
version is in ADAMS under Accession No. ML17297A349; documents related
to this amendment are listed in the Safety Evaluation enclosed with the
amendment.
Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment
revised the Facility Combined License.
Date of initial notice in Federal Register: September 12, 2017 (82
FR 42853). The supplemental letter dated November 16, 2017, provided
additional information that clarified the application, did not expand
the scope of the application request 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 the Safety Evaluation dated January 9, 2018.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-206, 50-361,
and 50-362, San Onofre Nuclear Generating Station (SONGS), Units 1, 2,
and 3, San Diego County, California
Date of amendment request: December 15, 2016.
Brief description of amendments: The amendments replace the SONGS,
Units 1, 2, and 3 Permanently Defueled Technical Specifications (TS)
with Independent Spent Fuel Storage Installation (ISFSI) Only TS. These
changes reflect the removal of all spent nuclear fuel from the SONGS,
Units 2 and 3, spent fuel pools and its transfer to dry cask storage
within the onsite ISFSI. The changes also make conforming revisions to
the SONGS, Unit 1, TS and combine them with the SONGS, Units 2 and 3,
TS. These changes will more fully reflect the permanently shutdown
status of the decommissioning facility, as well as the reduced scope of
structures, systems, and components necessary to ensure plant safety
once all spent fuel has been permanently moved to the SONGS ISFSI, an
activity which is currently scheduled for completion in 2019.
Date of issuance: January 9, 2017.
Effective date: As of the date Southern California Edison submits a
written notification to the NRC that all spent nuclear fuel assemblies
have been transferred out of the SONGS spent fuel pools and placed in
storage within the onsite independent spent fuel storage installation,
and shall be implemented within 60 days.
Amendment Nos.: Unit 1-169, Unit 2-237, and Unit 3-230: A publicly-
available version is in ADAMS under Accession No. ML17345A657;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Facility Operating License Nos. DPR-13, NPF-10, and NPF-15: The
amendments revise the Facility Operating Licenses.
Date of initial notice in Federal Register: February 14, 2017 (82
FR 10600).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 9, 2017.
No significant hazards consideration comments received: No.
Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna
Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: January 25, 2017, as supplemented by
letters dated March 21, 2017; August 4, 2017; and December 4, 2017.
Brief description of amendments: The amendments revised certain
surveillance requirements in Technical Specification 3.8.1, ``AC
[Alternating Current] Sources--Operating.'' The changes are in the use
of steady-state voltage and frequency acceptance criteria for onsite
standby power source of the diesel generators, allowing for the use of
new and more conservative design analysis.
Date of issuance: January 22, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 269 (Unit 1) and 251 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17352A711; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: June 6, 2017 (82 FR
26139). The supplemental letters dated August 4, 2017, and December 4,
2017, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 22, 2018.
No significant hazards consideration comments received: No.
IV. Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses and Final Determination of No Significant Hazards
Consideration and Opportunity for a Hearing (Exigent Public
Announcement or Emergency Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual notice of
consideration of issuance of amendment, proposed no significant hazards
consideration determination, and opportunity for a hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the
[[Page 6240]]
plant's licensed power level, the Commission may not have had an
opportunity to provide for public comment on its no significant hazards
consideration determination. In such case, the license amendment has
been issued without opportunity for comment. If there has been some
time for public comment but less than 30 days, the Commission may
provide an opportunity for public comment. If comments have been
requested, it is so stated. In either event, the State has been
consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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 51.12(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
application for amendment, (2) the amendment to Facility Operating
License or Combined License, as applicable, 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.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, any persons (petitioner) whose interest
may be affected by this action may file a request for a hearing and
petition for leave to intervene (petition) with respect to the action.
Petitions shall be filed in accordance with the Commission's ``Agency
Rules of Practice and Procedure'' in 10 CFR part 2. Interested persons
should consult a current copy of 10 CFR 2.309. The NRC's regulations
are accessible electronically from the NRC Library on the NRC's website
at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a
copy of the regulations is available at the NRC's Public Document Room,
located at One White Flint North, Room O1-F21, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of the amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission no later
than 60 days from the date of publication of this notice. The petition
must be filed in accordance with the filing instructions in the
``Electronic Submissions (E-Filing)'' section of this document, and
should meet the requirements for petitions set forth in this section,
except that under 10 CFR 2.309(h)(2) a State, local governmental body,
or federally
[[Page 6241]]
recognized Indian Tribe, or agency thereof does not need to address the
standing requirements in 10 CFR 2.309(d) if the facility is located
within its boundaries. Alternatively, a State, local governmental body,
Federally-recognized Indian Tribe, or agency thereof may participate as
a non-party under 10 CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
the discretion of the presiding officer, be permitted to make a limited
appearance pursuant to the provisions of 10 CFR 2.315(a). A person
making a limited appearance may make an oral or written statement of
his or her position on the issues but may not otherwise participate in
the proceeding. A limited appearance may be made at any session of the
hearing or at any prehearing conference, subject to the limits and
conditions as may be imposed by the presiding officer. Details
regarding the opportunity to make a limited appearance will be provided
by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for leave to intervene (petition), any
motion or other document filed in the proceeding prior to the
submission of a request for hearing or petition to intervene, and
documents filed by interested governmental entities that request to
participate under 10 CFR 2.315(c), must be filed in accordance with the
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR
46562, August 3, 2012). The E-Filing process requires participants to
submit and serve all adjudicatory documents over the internet, or in
some cases to mail copies on electronic storage media. Detailed
guidance on making electronic submissions may be found in the Guidance
for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit
paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to (1) request a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign submissions and access the E-Filing
system for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition or
other adjudicatory document (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on the NRC's public website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a
digital ID certificate and a docket has been created, the participant
can then submit adjudicatory documents. Submissions must be in Portable
Document Format (PDF). Additional guidance on PDF submissions is
available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the
time the document is submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public website at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC 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.
[[Page 6242]]
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2, Pope County, Arkansas
Date of amendment request: December 28, 2017.
Description of amendment: The amendment revised a note to Technical
Specification Surveillance Requirement (SR) 4.1.3.1.2, such that
Control Element Assembly (CEA) 4 may be excluded from the remaining
quarterly performances of the SR in Cycle 26. The amendment allows the
licensee to delay exercising CEA 4 until after repairs can be made
during the next outage.
Date of issuance: January 18, 2018.
Effective date: As of the date of issuance and shall be implemented
as soon as practicable and prior to the time in which SR 4.1.3.1.2 must
be completed.
Amendment No.: 308. A publicly-available version is in ADAMS under
Accession No. ML18011A064; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Renewed Facility Operating License and Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. Public notice of the proposed amendment was
published in the Arkansas Democrat-Gazette, located in Little Rock,
Arkansas, from January 6 through January 7, 2018. The notice provided
an opportunity to submit comments on the Commission's proposed NSHC
determination. No comments were received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated January 18, 2018.
Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel,
Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East,
Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant
(WBN), Unit 2, Rhea County, Tennessee
Date of amendment request: January 10, 2018, as supplemented by
letter dated January 17, 2018.
Description of amendment: The amendment revised Technical
Specification (TS) 3.3.4, ``Remote Shutdown Instrumentation,'' to make
a one-time change to TS Table 3.3.4-1, Function 4a, ``RCS Hot Leg
Temperature Indication,'' to permit the temperature indicator for the
Reactor Coolant System Loop 3 hot leg to be inoperable for the
remainder of WBN Unit 2 Operating Cycle 2, the refueling outage for
which is scheduled to start in spring 2019. The amendment also added a
condition to the operating license to require implementation of
compensatory measures described in the application that will remain in
effect until the temperature indicator is returned to an operable
condition.
Date of issuance: January 25, 2018.
Effective date: As of date of issuance.
Amendment No.: 19. A publicly-available version is in ADAMS under
Accession No. ML18022B106; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-96: Amendment revised the
technical specifications and operating license.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. The Rhea County Herald-News and The Advocate
& Democrat on January 21, 2018, and The Daily Post-Athenian on January
22 and January 23, 2018. The notice provided an opportunity to submit
comments on the Commission's proposed NSHC determination. The
supplemental letter dated January 17, 2018, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the notice.
No comments have been received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a Safety Evaluation dated January 25, 2018.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Undine Shoop.
Dated at Rockville, Maryland, this 6th day of February 2018.
For the Nuclear Regulatory Commission.
Greg A. Casto,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2018-02636 Filed 2-12-18; 8:45 am]
BILLING CODE 7590-01-P