[Federal Register Volume 83, Number 224 (Tuesday, November 20, 2018)]
[Notices]
[Pages 58607-58626]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-24894]
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NUCLEAR REGULATORY COMMISSION
[NRC-2018-0266]
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 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 October 23, 2018, to November 5, 2018. The
last biweekly notice was published on November 6, 2018.
DATES: Comments must be filed by December 20, 2018. A request for a
hearing must be filed by January 22, 2019.
ADDRESSES: You may submit comments by any of the following methods:
Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0266. Address
questions about Docket IDs in Regulations.gov to Jennifer Borges;
telephone: 301-287-9127; email: [email protected]. For technical
questions, contact the individual listed in the FOR FURTHER INFORMATION
CONTACT section of this document.
Mail comments to: May Ma, Office of Administration, Mail
Stop: TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Janet Burkhardt, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-1384, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2018-0266, 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-2018-0266.
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/
[[Page 58608]]
adams.html. To begin the search, select ``Begin Web-based ADAMS
Search.'' For problems with ADAMS, please contact the NRC's Public
Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or
by email to [email protected]. The ADAMS accession number for each
document referenced (if it is available in ADAMS) is provided the first
time that it is mentioned in this document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2018-0266, facility name, unit
number(s), plant docket number, application date, and subject in your
comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination.
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated; or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment
period such that failure to act in a timely way would result, for
example in derating or shutdown of the facility. If the Commission
takes action prior to the expiration of either the comment period or
the notice period, it will publish in the Federal Register a notice of
issuance. If the Commission makes a final no significant hazards
consideration determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and petition for leave to intervene
(petition) with respect to the action. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309. The NRC's regulations are accessible
electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the 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
[[Page 58609]]
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of the amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission no later
than 60 days from the date of publication of this notice. The petition
must be filed in accordance with the filing instructions in the
``Electronic Submissions (E-Filing)'' section of this document, and
should meet the requirements for petitions set forth in this section,
except that under 10 CFR 2.309(h)(2) a State, local governmental body,
or Federally-recognized Indian Tribe, or agency thereof does not need
to address the standing requirements in 10 CFR 2.309(d) if the facility
is located within its boundaries. Alternatively, a State, local
governmental body, Federally-recognized Indian Tribe, or agency thereof
may participate as a non-party under 10 CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
the discretion of the presiding officer, be permitted to make a limited
appearance pursuant to the provisions of 10 CFR 2.315(a). A person
making a limited appearance may make an oral or written statement of
his or her position on the issues but may not otherwise participate in
the proceeding. A limited appearance may be made at any session of the
hearing or at any prehearing conference, subject to the limits and
conditions as may be imposed by the presiding officer. Details
regarding the opportunity to make a limited appearance will be provided
by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing).
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for leave to intervene (petition), any
motion or other document filed in the proceeding prior to the
submission of a request for hearing or petition to intervene, and
documents filed by interested governmental entities that request to
participate under 10 CFR 2.315(c), must be filed in accordance with the
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR
46562; August 3, 2012). The E-Filing process requires participants to
submit and serve all adjudicatory documents over the internet, or in
some cases to mail copies on electronic storage media. Detailed
guidance on making electronic submissions may be found in the Guidance
for Electronic Submissions to the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit
paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to (1) request a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign submissions and access the E-Filing
system for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition or
other adjudicatory document (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on the NRC's public website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a
digital ID certificate and a docket has been created, the participant
can then submit adjudicatory documents. Submissions must be in Portable
Document Format (PDF). Additional guidance on PDF submissions is
available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the
time the document is submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public website at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC 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
[[Page 58610]]
delivery service upon depositing the document with the provider of the
service. A presiding officer, having granted an exemption request from
using E-Filing, may require a participant or party to use E-Filing if
the presiding officer subsequently determines that the reason for
granting the exemption from use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the
Commission or the presiding officer. If you do not have an NRC-issued
digital ID certificate as described above, click cancel when the link
requests certificates and you will be automatically directed to the
NRC's electronic hearing dockets where you will be able to access any
publicly available documents in a particular hearing docket.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or personal phone
numbers in their filings, unless an NRC regulation or other law
requires submission of such information. For example, in some
instances, individuals provide home addresses in order to demonstrate
proximity to a facility or site. With respect to copyrighted works,
except for limited excerpts that serve the purpose of the adjudicatory
filings and would constitute a Fair Use application, participants are
requested not to include copyrighted materials in their submission.
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2 (Catawba), York County, South Carolina
Date of amendment request: July 19, 2018. A publicly-available
version is in ADAMS under Accession No. ML18200A252.
Description of amendment request: The amendments would modify the
Catawba Updated Final Safety Analysis Report (UFSAR), Section
6.2.4.2.2, ``Containment Valve Injection Water System [CVIWS],'' to
remove the CVIWS supply from specified Safety Injection (NI) and
Containment Spray (NS) Containment Isolation Valves (CIVs), and to
exempt these CIVs from Type C Local Leak Rate Testing (LLRT).
Additionally, the amendments would modify UFSAR, Table 6-77,
``Containment Isolation Valve Data,'' to make corresponding changes.
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 amendment request is to remove select Containment Isolation
Valves from the Local Leak Rate Test (LLRT) program. These valves
were originally included in the LLRT under 10 CFR 50, Appendix J, in
what is now Option A. [Catawba] has been approved for 10 CFR 50,
Appendix J, Option B under License Amendment No. 192/184. Under
Option B, valves may be exempted from LLRT Type C testing if they
are not a potential containment atmosphere leakage path. Based on
the design and operation of the NI and NS Systems, the valves do not
constitute a containment atmospheric leakage path as covered in the
Safety Evaluation. Since the valves are not a leakage path, there is
no impact on the consequence of an accident. Moreover, the valves
are not a part of the Reactor Coolant Pressure Boundary, thus they
do not affect the probability of an accident.
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 systems design and operation are not changing. This test
exemption does not change the way the valves are used as a part of
the NI and NS Systems. A detailed Failure Modes and Effects Analysis
was completed to confirm the system operation would meet the
containment isolation design function.
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
the margin of safety?
Response: No.
The test exemption is within existing regulatory requirements.
The application of a closed loop outside of containment is
appropriate and consistent with regulatory positions. With
containment integrity maintained within the allowable regulatory
framework, there is no reduction in the margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Exelon FitzPatrick, LLC and Exelon Generation Company, LLC, Docket No.
50-333, James A. FitzPatrick Nuclear Power Plant (FitzPatrick), Oswego
County, New York
Date of amendment request: October 2, 2018. A publicly-available
version is in ADAMS under Accession No. ML18275A060.
Description of amendment request: The amendment would modify the
Technical Specifications concerning a change to the method of
calculating core reactivity for the purpose of performing the
reactivity anomaly surveillance at FitzPatrick.
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 Technical Specification change does not affect any
plant systems, structures, or components designed for the prevention
or mitigation of previously evaluated accidents. The amendment would
only change how the reactivity anomaly surveillance is performed.
Verifying that the core reactivity is consistent with predicted
values ensures that accident and transient safety analyses remain
valid. This amendment changes the Technical Specification
requirements such that, rather than performing the surveillance by
comparing predicted to actual control rod density, the surveillance
is performed by a direct comparison of keff. Present day
online core monitoring systems, such as the one in use at the James
A. FitzPatrick Nuclear Power Plant [(JAFNPP)], Unit 1 are capable of
performing the direct measurement of reactivity.
Therefore, since the reactivity anomaly surveillance will
continue to be performed by a viable method, the proposed amendment
does not involve a significant increase in the probability or
consequence of a previously evaluated accident.
[[Page 58611]]
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This Technical Specifications amendment request does not involve
any changes to the operation, testing, or maintenance of any safety-
related, or otherwise important to safety systems. All systems
important to safety will continue to be operated and maintained
within their design bases. The proposed changes to the reactivity
anomaly Technical Specifications will only provide a new, more
efficient method of detecting an unexpected change in core
reactivity.
Since all systems continue to be operated within their design
bases, no new failure modes are introduced and the possibility of a
new or different kind of accident is not created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This proposed Technical Specifications amendment proposes to
change the method for performing the reactivity anomaly surveillance
from a comparison of predicted to actual control rod density to a
comparison of predicted to actual keff. The direct
comparison of keff provides a technically superior method
of calculating any differences in the expected core reactivity. The
reactivity anomaly surveillance will continue to be performed at the
same frequency as is currently required by the Technical
Specifications, only the method of performing the surveillance will
be changed. Consequently, core reactivity assumptions made in safety
analyses will continue to be adequately verified.
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: Donald P. Ferraro, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Suite 305,
Kennett Square, PA 19348.
NRC Branch Chief: James G. Danna.
Exelon Generation Company (EGC), LLC, Docket No. 50-461, Clinton Power
Station (CPS), Unit No. 1, DeWitt County, Illinois
Date of amendment request: September 28, 2018. A publicly-available
version is in ADAMS under Accession No. ML18271A217.
Description of amendment request: The amendment would make
Technical Specification (TS) changes that are consistent with NRC-
approved Industry Technical Specification Task Force (TSTF) Standard
Technical Specification Change Traveler, TSTF-476, Revision 1. The
availability of this TS improvement was announced in the Federal
Register on May 23, 2007 (72 FR 29004).
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 modifies the TS to allow the use of the
improved BPWS [Banked Position Withdrawal Sequence] during shutdowns
if the conditions of NEDO-33091-A, Revision 2, ``Improved BPWS
Control Rod Insertion Process,'' July 2004 [ADAMS Accession No.
ML042230366], have been satisfied. The justifications to support the
specific TS changes are consistent with the approved topical report
and TSTF-476, Revision 1. Since the change only involves changes in
control rod sequencing, the probability of an accident previously
evaluated is not significantly increased, if at all. The
consequences of an accident after adopting TSTF-476 are no different
than the consequences of an accident prior to adopting TSTF-476.
Therefore, the consequences of an accident previously evaluated are
not significantly affected by 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 previously evaluated?
Response: No.
The proposed change will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The control rod drop accident (CRDA)
is the design basis accident for the subject TS changes. This change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change, TSTF-476, Revision 1, incorporates the
improved BPWS, previously approved in NEDO-33091-A, into the CPS TS.
The CRDA is the design basis accident for the subject TS changes. In
order to minimize the impact of a CRDA, the BPWS process was
developed to minimize control rod reactivity worth for boiling water
reactor plants. The proposed improved BPWS further simplifies the
shutdown control rod insertion process, and in order to evaluate it,
the NRC followed the guidelines of Standard Review Plan Section
15.4.9, and referred to General Design Criterion 28 of Appendix A to
10 CFR part 50 as its regulatory requirement. The TSTF stated the
improved BPWS provides the following benefits: (1) Allows the plant
to reach the all-rods-in condition prior to significant reactor cool
down, which reduces the potential for recriticality as the reactor
cools down; (2) reduces the potential for an operator reactivity
control error by reducing the total number of control rod
manipulations; (3) minimizes the need for manual scrams during plant
shutdowns, resulting in less wear on control rod drive (CRD) system
components and CRD mechanisms; and (4) eliminates unnecessary
control rod manipulations at low power, resulting in less wear on
reactor manual control and CRD system components. The addition of
procedural requirements and verifications specified in NEDO-33091-A,
along with the proper use of the BPWS will prevent a CRDA from
occurring while power is below the low power setpoint (LPSP). The
net change to the margin of safety is insignificant. Therefore, this
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: David J. Wrona.
Exelon Generation Company, LLC (Exelon), Docket No. 50-289, Three Mile
Island Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: July 25, 2018. A publicly-available
version is in ADAMS under Accession No. ML18206A545.
Description of amendment request: The amendment would revise the
TMI-1 Renewed Facility Operating License (RFOL) and associated
Technical Specifications (TSs) to the Permanently Defueled Technical
Specifications (PDTSs), consistent with the permanent cessation of
reactor operation and permanent defueling of the reactor. By letter
dated June 20, 2017 (ADAMS Accession No. ML17171A151), Exelon provided
formal notification to the NRC of Exelon's contingent determination to
permanently cease operations at TMI-1 no later than September 30, 2019.
The amendment would eliminate those TSs applicable in operating mode or
modes where fuel is placed in the reactor vessel. The amendment would
change other TS limiting conditions for operation (LCOs), definitions,
surveillance requirements, and administrative controls, as well as
several license conditions. The
[[Page 58612]]
amendment would also modify the licensing basis mitigation strategies
for flood mitigation and aircraft impact protection in the air intake
tunnel.
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 has
certified to the NRC that it has permanently ceased operation and
entered a permanently defueled condition. Because the 10 CFR part 50
license for TMI will no longer authorize operation of the reactor,
or emplacement or retention of fuel into the reactor vessel with the
certifications required by 10 CFR part 50.82(a)(1) submitted, as
specified in 10 CFR part 0.82(a)(2), the occurrence of postulated
accidents associated with reactor operation is no longer credible.
The remaining UFSAR [Updated Final Safety Analysis Report]
Chapter 14 postulated design basis accident (DBA) events that could
potentially occur at a permanently defueled facility would be a Fuel
Handling Accident (FHA) in the Spent Fuel pool (SFP), Waste Gas Tank
Rupture (WGTR), and Fuel Cask Drop Accident (FCDA). The FHA analyses
for TMI shows that, following 60 days of decay time after reactor
shutdown and provided the SFP water level requirements of proposed
TS LCO \3/4\.1.1 are met, the dose consequences are acceptable
without relying on SSCs [structures, systems, and components] to
remain functional for accident mitigation during and following the
event. The one exception to this is the continued function of the
passive SFP structure. The remaining DBAs that support permanently
shutdown and defueled condition do not rely on any active safety
system for mitigation.
The probability of occurrence of previously evaluated accidents
is not increased, since extended operation in a defueled condition
and safe storage and handling of fuel will be the only operations
performed, and therefore, bounded by the existing analyses.
Additionally, the occurrence of postulated accidents associated with
reactor operation will no longer be credible in a permanently
defueled reactor. This significantly reduces the scope of applicable
accidents.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to delete and/or modify certain
[requirements of the] TMI RFOL, TS, or CLB [Current Licensing Basis]
have no impact on facility SSCs affecting the safe storage of spent
irradiated fuel, or on the methods of operation of such SSCs, or on
the handling and storage of spent irradiated fuel itself. The
removal of TS that are related only to the operation of the nuclear
reactor, or only to the prevention, diagnosis, or mitigation of
reactor related transients or accidents, cannot result in different
or more adverse failure modes or accidents than previously evaluated
because the reactor will be permanently shutdown and defueled and
TMI will no longer be authorized to operate the reactor.
The proposed modification or deletion of requirements of the TMI
RFOL, TS, and CLB [does] not affect systems credited in the accident
analysis for the remaining credible DBAs at TMI. The proposed RFOL
and PDTS will continue to require proper control and monitoring of
safety significant parameters and activities. The TS regarding SFP
water level and spent fuel storage is retained to preserve the
current requirements for safe storage of irradiated fuel.
The proposed amendment does not result in any new mechanisms
that could initiate damage to the remaining relevant safety barriers
for defueled plants (fuel cladding, spent fuel racks, SFP integrity,
and SFP water level). Since extended operation in a defueled
condition and safe fuel handling will be the only operation allowed,
and therefore bounded by the existing analyses, such a condition
does not create the possibility of a new or different kind of
accident.
Therefore, the proposed 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 deleting and/or modifying certain
[requirements of the] RFOL, TS, and CLB once the TMI facility has
been permanently shutdown and defueled. Because the 10 CFR part 50
license for TMI [will] no longer [authorize] operation of the
reactor, or emplacement or retention of fuel into the reactor vessel
with the certifications required by 10 CFR part 50.82(a)(1)
submitted, as specified in 10 CFR part 50.82(a)(2), the occurrence
of postulated accidents associated with reactor operation is no
longer credible. The remaining postulated DBA events that could
potentially occur at a permanently defueled facility would be a FHA,
WGTR, and FCDA. The proposed amendment does not adversely affect the
inputs or assumptions of any of the design basis analyses.
The proposed changes are limited to those portions of the RFOL,
TS, and CLB that are not related to the safe storage of irradiated
fuel. The requirements that are proposed to be revised or deleted
from the RFOL, TS, and CLB are not credited in the existing accident
analysis for the remaining applicable postulated accidents; and as
such, do not contribute to the margin of safety associated with the
accident analysis. Postulated design basis accidents involving the
reactor will no longer be possible because the reactor will be
permanently shutdown and defueled and TMI 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.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277
and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and
Lancaster Counties, Pennsylvania
Date of amendment request: September 27, 2018. A publicly-available
version is in ADAMS under Accession No. ML18271A009.
Description of amendment request: The amendment would modify the
applicability for Technical Specification (TS) Section 3.3.6.2,
``Secondary Containment Isolation Instrumentation,'' Functions 3 and 4,
related to reactor building and refueling floor ventilation exhaust,
respectively. This change would be implemented in the fall of 2019.
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 requested changes to TS Section 3.3.6.2 to revise the
applicability of Functions 3 and 4 as proposed does not eliminate
the design function associated with the radiation monitoring
instrumentation. The Secondary Containment Isolation Instrumentation
will continue to automatically initiate closure of appropriate
Secondary Containment Isolation Valves (SCIVs) and start the Standby
Gas Treatment (SGT) system as designed to limit fission product
release during any postulated Design Basis Accidents (DBAs). These
systems are not accident initiators. The proposed changes will
continue to assure that these systems perform their design
functions, which includes mitigating accidents. The proposed changes
do not alter the physical design of any plant Structure, System, or
Components (SSC); therefore, the proposed changes have no adverse
effect on plant operation, or the availability or operation of any
accident mitigation equipment. The plant response to
[[Page 58613]]
DBAs does not change and remains as analyzed in the Updated Final
Safety Analysis Report (UFSAR).
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 requested changes to TS Section 3.3.6.2 to revise the
applicability of Functions 3 and 4 as proposed does not adversely
affect the design function associated with the radiation monitoring
instrumentation. The proposed changes do not change any system
operations or maintenance activities that would create the
possibility of a new or different kind of accident from one
previously evaluated. The Secondary Containment Isolation
Instrumentation and SGT system will continue to function as
designed. The proposed changes will continue to assure that these
systems perform their design functions, which includes mitigating
accidents. The proposed changes do not create new failure modes or
mechanisms and no new accident precursors are created. The proposed
changes do not alter the plant configuration (no new or different
type of equipment is being installed) or require any new or unusual
Operator actions. The proposed changes do not alter the safety
limits or safety analysis assumptions associated with the operation
of the plant. The proposed changes do not introduce any new failure
modes or mechanisms that could result in a new accident. The
proposed changes do not reduce or adversely affect the capabilities
of any plant SSC in the performance of their safety function. Also,
the response of the plant and the Operators following any DBA is
unaffected by the proposed changes.
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 requested changes to TS Section 3.3.6.2 to revise the
applicability of Functions 3 and 4 as proposed does not alter the
design capability associated with the radiation monitoring
instrumentation. The proposed changes have no adverse effect on
plant operation, or the availability or operation of any accident
mitigation equipment. The plant response to DBAs does not change.
The proposed changes do not adversely affect existing plant safety
margins or the reliability of the equipment assumed to operate in
the safety analyses. There is no change being made to safety
analysis assumptions, safety limits or limiting safety system
settings that would adversely affect plant safety as a result of the
proposed changes.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Rd., 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: September 20, 2018. A publicly-available
version is in ADAMS under Accession No. ML18263A199.
Description of amendment request: The amendment would make
administrative changes to Technical Specification 4.4.2.1, ``Inservice
Tendon Surveillance Requirements.'' The amendment would add the words
``except where an alternative, exemption, or relief has been authorized
by the NRC'' to allow NRC-approved exceptions to the 10 CFR 50.55a
requirements. Also, the amendment would add a note to exempt from the
requirements of Surveillance Requirement 4.0.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The addition of the words ``except where an alternative,
exemption, or relief has been authorized by the NRC'' to Technical
Specification (TS) 4.4.2.1 (``lnservice Tendon Surveillance
Requirements'') and the addition of the wording ``The surveillance
interval extension allowed per Surveillance Requirement 4.0.1 is not
permitted'' are administrative changes that have no impact on the
accidents analyzed and are not an accident initiator. Since the
changes do not impact any conditions that would initiate an
accident, the probability or consequences of previously analyzed
events is not increased.
The proposed changes do not involve the modification of any
plant equipment or affect plant operation. The proposed changes will
have no impact on any safety-related structures, systems, or
components.
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.
No safety-related equipment, safety function, or plant operation
will be altered as a result of these proposed administrative
changes. No new operator actions are created as a result of the
proposed changes. These administrative changes have no impact on the
accidents analyzed in the Updated Final Safety Analysis Report
(UFSAR) and are not accident initiators. These proposed changes do
not impact the U.S. Nuclear Regulatory Commission Staff's authority
to review and grant exceptions. The addition of the wording ``The
surveillance interval extension allowed per Surveillance Requirement
4.0.1 is not permitted'' has been added to address the concerns
identified in the U.S. Nuclear Regulatory Commission's Safety
Evaluation Report [(Reference 3 of the licensee's letter dated
September 20, 2018)].
Since these proposed changes do not impact any conditions that
would initiate an accident, there is no possibility of a new or
different kind of accident resulting from these changes. 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 administrative changes do not affect any margins of
safety. The margins of safety presently provided by the Technical
Specifications remain unchanged. The proposed amendment does not
affect the design of the facility or system operating parameters,
does not physically alter safety-related systems, structures, or
components (SSCs) and does not affect the method in which safety-
related systems perform their functions.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1 (FCS), Washington County, Nebraska
Date of amendment request: September 28, 2018. A publicly-available
version is in ADAMS under Accession No. ML18275A323.
Description of amendment request: The proposed amendment would
revise
[[Page 58614]]
the Renewed Facility License and the Permanently Defueled Technical
Specifications (PDTS) for FCS to reflect the requirements after removal
of all remaining spent nuclear fuel from the spent fuel pool (SFP) and
its transfer to dry cask storage within an Independent Spent Fuel
Storage Installation (ISFSI).
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 amendment would modify the FCS renewed facility
operating license and PDTS by deleting the portions of the license
and PDTS that are no longer applicable to a facility with no spent
nuclear fuel stored in the spent fuel pool, while modifying the
remaining portions to correspond to all nuclear fuel stored within
an ISFSI. This amendment becomes effective upon removal of all spent
nuclear fuel from the FCS SFP and its transfer to dry cask storage
within an ISFSI. The definition of safety-related structures,
systems, and components (SSCs) in 10 CFR 50.2 states that safety-
related SSCs are those relied on to remain functional during and
following design basis events to assure:
1. The integrity of the reactor coolant boundary;
2. The capability to shutdown the reactor and maintain it in a
safe shutdown condition; or
3. The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures
comparable to the applicable guideline exposures set forth in 10 CFR
50.34(a)(1) or Sec. 100.11 .
The first two criteria (integrity of the reactor coolant
pressure boundary and safe shutdown of the reactor) are not
applicable to a plant in a permanently defueled condition. The third
criterion is related to preventing or mitigating the consequences of
accidents that could result in potential offsite exposures exceeding
limits. However, after all nuclear spent fuel assemblies have been
transferred to dry cask storage within an ISFSI, none of the SSCs at
FCS are required to be relied on for accident mitigation. Therefore,
none of the SSCs at FCS meet the definition of a safety-related SSCs
stated in 10 CFR 50.2. The proposed deletion of requirements in the
FCS PDTS does not affect systems credited in any accident analysis
at FCS.
Chapter 14 of the FCS Defueled Safety Analysis Report (DSAR)
described the design basis accident related to the SFP. These
postulated accidents are predicated on spent fuel being stored in
the SFP. With the removal of the spent fuel from the SFP, there are
no remaining spent fuel assemblies to be monitored and there are no
credible accidents that require the actions of a Shift Manager,
Certified Fuel Handler, or a Non-certified Operator to prevent
occurrence or mitigate the consequences of an accident associated
with nuclear fuel. The proposed changes do not have an adverse
impact on the remaining decommissioning activities or any of their
postulated consequences. The proposed changes related to the
relocation of certain administrative requirements do not affect
operating procedures or administrative controls that have the
function of preventing or mitigating any accidents applicable to the
safe management of irradiated fuel or decommissioning of the
facility. Therefore, the proposed amendment 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 eliminate the operational requirements and
certain design requirements associated with the storage of the spent
fuel in the SFP, and relocate certain administrative controls to the
Quality Assurance Topical Report which is a licensee-controlled
document. After the removal of the spent fuel from the SFP and
transfer to the ISFSI, there are no spent fuel assemblies that
remain in the SFP. Coupled with a prohibition against storage of
fuel in the SFP, the potential for fuel related accidents is
removed. The proposed changes do not introduce any new failure
modes.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The removal of all spent nuclear fuel from the SFP into storage
in casks within an ISFSI, coupled with a prohibition against future
storage of fuel within the SFP, removes the potential for fuel
related accidents.
The design basis and accident assumptions within the FCS DSAR
and the PDTS relating to safe management and safety of spent fuel in
the SFP are no longer applicable. The proposed changes do not affect
remaining plant operations, systems, or components supporting
decommissioning activities.
The requirements for SSCs that have been deleted from the FCS
PDTS are not credited in the existing accident analysis for any
applicable postulated accident; and as such, do not contribute to
the margin of safety associated with the accident analysis.
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: Stephen M. Bruckner, Attorney, Fraser
Stryker PC LLO, 500 Energy Plaza, 409 South 17th Street, Omaha, NE
68102.
NRC Branch Chief: Bruce A. Watson.
South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station
(VCSNS), Unit No. 1, Fairfield County, South Carolina
Date of amendment request: September 27, 2018. A publicly-available
version is in ADAMS under Accession No. ML18270A360.
Description of amendment request: The proposed amendment would
correct a non-conservative Technical Specification (TS) 3/4.8.2, ``DC
[Direct Current] Sources -Operating,'' by revising the inter-cell
resistance value listed in Surveillance Requirements (SRs) 4.8.2.1.b.2
and 4.8.2.1.c.3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. [Do] the proposed change[s] involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Performing the proposed changes in battery parameter
surveillance testing and verification is not a precursor of any
accident previously evaluated. Furthermore, these changes will help
to ensure that the voltage and capacity of the batteries is such
that they will provide the power assumed in calculations of design
basis accident mitigation. Therefore, SCE&G concludes that the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the VCSNS TS SR do not involve any
physical modification of the plant or how the plant is operated. No
new or different type of equipment will be installed. The proposed
changes involve surveillance testing and verification activities. No
new failure modes/effects which could lead to an accident whose
consequences exceed the consequences of accidents previously
analyzed will be introduced by the changes to the TS SR.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission
[[Page 58615]]
product barriers to perform their design functions during and
following an accident situation. These barriers include the fuel
cladding, the reactor coolant system, and the containment system.
The performance of the fuel cladding, reactor coolant, and
containment systems will not be impacted by the proposed changes.
The proposed VCSNS revisions of the SRs ensure the continued
availability and operability of the batteries. As such, sufficient
DC capacity to support operation of mitigation equipment remains
within the design basis. Therefore, SCE&G concludes that 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: Kathryn M. Sutton, Morgan, Lewis & Bockius
LLP, 1111 Pennsylvania Avenue NW, Washington, DC 20004.
NRC Branch Chief: Michael T. Markley.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station (VCSNS),
Unit No. 1, Fairfield County, South Carolina
Date of amendment request: October 8, 2018. A publicly-available
version is in ADAMS under Accession No. ML18281A014.
Description of amendment request: The proposed amendment would
revise the Surveillance Requirement (SR) of Technical Specification
(TS) 4.4.6.2.2 (a) to allow the reactor coolant system (RCS) pressure
isolation valve (PIV) leakage test to be extended to a performance-
based frequency not to exceed 3 refueling outages (RFOs) or 60 months
following two consecutive satisfactory tests.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves revising the VCSNS Unit 1, TS
wording to reflect a performance-based surveillance testing interval
for leakage testing of the RCS PIVs. Specifically, the proposed
change revises TS surveillance requirement (SR) 4.4.6.2.2.a to test
the RCS PIVs at a frequency from each RFO to a maximum of every
third RFO or 60 months by verifying that each of the PIVs tested in
the associated RFO based on performance are within the TS allowable
leakage limits. The RCS PIVs are defined as two normally closed
valves in series with the reactor coolant pressure boundary (RCPB),
which separate the high-pressure RCS from an attached lower pressure
system. Excessive PIV leakage could lead to overpressure of the low-
pressure piping or components, potentially resulting in a LOCA
[loss-of-coolant accident] outside of containment.
TS SR 4.4.6.2.2.a for RCS PIVs provides added assurance of valve
integrity thereby reducing the probability of gross valve failure
and consequent ISLOCA [intersystem loss-of-coolant accident]. The
RCS PIV allowable leakage limit applies to each individual valve.
This proposed change does not revise any of the TS RCS PIV allowable
leakage limits. In addition, the RCS PIVs will continue to be tested
per the VCSNS Inservice Testing Program in accordance with Title 10,
Code of Federal Regulations (CFR), Section 50.55a, ``Codes and
standards.'' The activity does not involve a physical change to the
plant or a change in the manner in which the plant is operated or
controlled. By transitioning to a performance-based leakage testing
interval, these valves will continue to be demonstrated
operationally ready and reliable. In the event of a PIV leakage test
failure, PIV testing would require the component to return to the
initial interval of every RFO until good performance is re-
established. Therefore, there is no impact on the assurance that the
RCS PIVs will be able to perform their safety function(s).
Therefore, the proposed TS 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 VCSNS TS wording to
reflect a performance-based surveillance testing interval for
leakage testing of the RCS PIVs from each RFO to a maximum of every
third RFO or 60 months based on valve performance. The technical
testing methodology and associated acceptance criteria remain
unchanged. The change in the testing frequency is a performance-
based approach, which has been demonstrated acceptable in numerous
applications across the industry (RCS PIV testing, 10 CFR 50,
Appendix J, Option B).
The testing requirements involved to periodically demonstrate
the integrity of the RCS PIVs exist to ensure the plant's ability to
mitigate the consequences of an accident. There are not any accident
initiators or precursors affected by this change. The proposed TS
change does not involve a physical change to the plant or the manner
in which the plant is operated or controlled.
Therefore, the proposed TS 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 involves revising the TS SR 4.4.6.2.2.a and
associated TS Bases to reflect a performance-based surveillance
testing frequency of the RCS PIVs from each RFO to a maximum of
every third RFO or 60 months. The technical testing methodology and
associated TS allowable leakage limits/acceptance criteria remain
unchanged. The testing frequency uses a performance based approach,
which has been demonstrated acceptable in numerous applications
across the industry (RCS PIV testing, 10 CFR 50, Appendix J, Option
B). Thus, this amendment request does not alter the manner in which
safety limits, limiting safety system set points, or limiting
conditions for operation are determined. The RCS PIVs will continue
to be tested per the VCSNS Inservice Testing Program in accordance
with 10 CFR 50.55a.
The primary reason for performance-based PIV test intervals is
to eliminate unnecessary thermal cycles. The VCSNS program for
monitoring fatigue due to operational cycles and transients consists
of review, evaluation, and documentation of RCS operational
transients/cycles based on recorded plant operating parameters
(i.e., temperature, pressure, flow) for compliance with Technical
Specification Sections 3.5.2, 3.5.3, and 5.7.1.
An additional reason for requesting performance-based PIV test
intervals is dose reduction to conform with NRC and industry As Low
As Reasonably Achievable (ALARA) radiation dose principles. The
nominal fuel cycle lengths at VCSNS, Unit 1, are 18 months. However,
since RFOs may be scheduled slightly beyond 18 months, a 60-month
period is used to provide a bounding timeframe to encompass three
RFOs. The review of recent historical data identified that PIV
testing each RFO results in a total personnel dose of approximately
300 millirem (milli-Roentgen Equivalent Man, or mrem). Assuming all
of the PIVs remain classified as good performers, the proposed
extended test intervals would provide for a savings of approximately
600 mrem over an approximate 60-month period (three RFOs).
The proposed surveillance interval extension for the RCS PIVs is
based on the performance of the PIVs. The proposed TS change does
not involve a physical change to the plant or a change in the manner
in which the plant is operated or controlled. The design, operation,
testing methods, and acceptance criteria for the RCS PIV testing
specified in applicable codes and standards will continue to be met.
Therefore, the proposed TS 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 M. Sutton, Morgan, Lewis & Bockius
LLP, 1111 Pennsylvania Avenue NW, Washington, DC 20004.
[[Page 58616]]
NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Inc., Docket No. 52-025, Vogtle
Electric Generating Plant (VEGP), Unit 3, Burke County, Georgia
Date of amendment request: October 19, 2018. A publicly-available
version is in ADAMS under Accession No. ML18292A660.
Description of amendment request: The requested amendment proposes
to depart from certified AP1000 Design Control Document (DCD) Tier 2*
material that has been incorporated into the Updated Final Safety
Analysis Report (UFSAR). Specifically, the proposed departure consists
of changes to Tier 2* information in the UFSAR (which includes the
plant-specific DCD information) to change the vertical reinforcement
information provided in the VEGP Unit 3 column line 1 wall from
elevation 135'-3'' to 137'-0''.
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.
As described in UFSAR Subsection 3H.5.1.1, the exterior wall at
column line 1 (Wall 1) is located at the south end of the auxiliary
building. It is a reinforced concrete wall extending from the
basemat at elevation 66'-6'' to the roof at elevation 180'-0''.
Deviations were identified in the constructed wall from the design
requirements. The proposed change modifies the vertical
reinforcement information provided in the VEGP Unit 3 Wall 1 from
elevation 135'-3'' to 137'- 0''. This change maintains conformance
to the [American Concrete Institute (ACI)] 318-11 and ACI 349-01
codes and has no adverse impact on the seismic response of Wall 1.
Wall 1 continues to withstand the design basis loads without loss of
structural integrity or the safety-related functions. The proposed
change does not affect the operation of any system or equipment that
initiates an analyzed accident or alter any SSC [structures,
systems, and components] accident initiator or initiating sequence
of events.
This change does not adversely affect the design function of the
VEGP Unit 3 Wall 1 or the SSCs contained within the auxiliary
building. This change does not involve any accident initiating
components or events, thus leaving the probabilities of an accident
unaltered.
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 change modifies the vertical reinforcement
information provided in the VEGP Unit 3 Wall 1 from elevation 135'-
3'' to 137'-0''. As demonstrated by the continued conformance to the
applicable codes and standards governing the design of the
structures, the wall withstands the same effects as previously
evaluated. The proposed change does 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 change does
not adversely affect the design function of the auxiliary building
Wall 1 or 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 non-safety-related
equipment. This change does not allow for a new fission product
release path, result in a new fission product barrier failure mode,
or create a new sequence of events that result in significant fuel
cladding failures.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change modifies the vertical reinforcement
information provided in the VEGP Unit 3 Wall 1 from elevation 135'-
3'' to 137'-0''. This change maintains conformance to the ACI 318-11
and ACI 349-01 codes. The change to the vertical reinforcement
elevation 135'-3'' to 137'-0'' does not change the performance of
the affected portion of the auxiliary building for postulated loads.
The criteria and requirements of ACI 349-01 provide a margin of
safety to structural failure. The design of the auxiliary building
structure conforms to criteria and requirements in ACI 349-01 and
therefore, maintains the margin of safety. The change does not alter
any design function, design analysis, or safety analysis input or
result, and sufficient margin exists to justify departure from the
Tier 2* requirements for the wall. As such, because the system
continues to respond to design basis accidents in the same manner as
before without any changes to the expected response of the
structure, no safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed changes.
Accordingly, no significant safety margin is reduced by the change.
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: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer L. Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke
County, Georgia
Date of amendment request: October 11, 2018. A publicly-available
version is in ADAMS under Accession Nos. ML18284A447.
Description of amendment request: The requested amendment proposes
changes to plant-specific Design Control Document (DCD) Tier 2
information in the Updated Final Safety Analysis Report (UFSAR) that
involve changes to combined license (COL) Appendix C, and corresponding
changes to plant-specific Tier 1 information. The changes would revise
the COL to relocate the power operated relief valves in the COL
Appendix C, Inspections, Tests, Analyses, and Acceptance Criteria and
in the UFSAR. An initial Federal Register notice was published on
September 19, 2018 (83 FR 47375), providing an opportunity to comment,
request a hearing, and petition for leave to intervene for a License
Amendment Request (LAR) for the VEGP COLs. The licensee has submitted a
revision, dated October 11, 2018, to the original LAR that was dated
August 10, 2018. This revision increases the scope of the original LAR.
Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from
elements of the design as certified in the 10 CFR part 52, Appendix D,
design certification rule is also requested for the plant-specific DCD
Tier 1 departures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not affect the operation or reliability
of any system, structure or component (SSC) required to maintain a
normal power operating condition or to mitigate anticipated
transients without safety-related systems. With the proposed
changes, the PORV [Power Operated Relief Valve] block valves are
still able to perform the safety-related functions of containment
isolation, steam generator isolation, and steam generator relief
isolation. There is no
[[Page 58617]]
change to the PORV block valves safety class or safety-related
functions.
The relocation of the branch line in which the PORV block valves
are installed in allows the PORV block valves to be closer to the
containment penetration and maintain compliance with General Design
Criterion (GDC) 57 for locating containment isolation valves as
close to the containment as practical.
There is no impact to Chapter 15 evaluations. Changes to the
PORV block valve and line size do not impact the mass releases to
the atmosphere during a Steam Generator Tube Rupture accident. The
mass release is limited by the PORV which is more restrictive than
the PORV block valve and line size.
There is no impact to any assumed leakage through the PORV line.
The existing 12-inch PORV has a design function to limit leakage
through the PORV line. Increasing the PORV block valve to 12 inches
will increase the leakage through the PORV block valve however it
will be that same leakage rate as the 12-inch PORV. Therefore, the
leakage rate through the PORV line does not increase and there is no
impact to radiation doses.
There is no impact to the assumptions or analysis in the
completed safety analysis for radiation doses as a result of the
change.
There is no impact to the conclusions of the Pipe Rupture Hazard
Analysis (PRHA) because the PORV line is Break Exclusion Zone (BEZ)
piping. The proposed changes do not result in any new postulated
break locations. Updated analyses confirm that the integrity of the
wall adjacent to the MCR [main control room] is unaffected by a
postulated main steam line break that causes the PORV line to impact
the wall.
There is no change to the valve motor operator. The current
motor operator is sufficient to operate the new 12-inch globe valve.
Therefore, there is no impact to the Class 1E dc [direct current]
and UPS [uninterruptable power supply] System (IDS) battery sizing.
There is no change to the valve stroke time, therefore there is no
impact to valve open/closure times.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not affect the operation of systems or
equipment that could 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. With the proposed changes, the PORV
block valves are still able to perform the safety related functions
of containment isolation, steam generator isolation, and steam
generator relief isolation. There is no change to the PORV block
valves safety class or safety-related functions.
The relocation of the branch line in which the PORV block valves
are installed in allows the PORV block valves to be closer to the
containment penetration and maintain compliance with General Design
Criterion (GDC) 57 for locating containment isolation valves as
close to the containment as practical.
There is no impact to Chapter 15 evaluations. Changes to the
PORV block valve and line size do not impact the mass releases to
the atmosphere during a Steam Generator Tube Rupture accident. The
mass release is limited by the PORV which is more restrictive than
the PORV block valve and line size.
There is no impact to any assumed leakage through the PORV line.
The existing 12-inch PORV has a design function to limit leakage
through the PORV line. Increasing the PORV block valve to 12 inches
will increase the leakage through the PORV block valve however it
will be that same leakage rate as the 12-inch PORV. Therefore, the
leakage rate through the PORV line does not increase and there is no
impact to radiation doses.
There is no impact to the assumptions or analysis in the
completed safety analysis for radiation doses as a result of the
change.
There is no impact to the conclusions of the Pipe Rupture Hazard
Analysis (PRHA) because the PORV line is Break Exclusion Zone (BEZ)
piping. The proposed changes do not result in any new postulated
break locations. Updated analyses confirm that the integrity of the
wall adjacent to the MCR is unaffected by a postulated main steam
line break that causes the PORV line to impact the wall.
There is no change to the valve motor operator. The current
motor operator is sufficient to operate the new 12-inch globe valve.
Therefore, there is no impact to the Class 1E dc and UPS System
(IDS) battery sizing. There is no change to the valve stroke time,
therefore there is no impact to valve open/closure times.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not affect existing safety margins. With
the proposed changes, the PORV block valves are still able to
perform the safety-related functions of containment isolation, steam
generator isolation, and steam generator relief isolation. There is
no change to the PORV block valves safety class or safety-related
functions.
The relocation of the branch line in which the PORV block valves
are installed in allows the PORV block valves to be closer to the
containment penetration and maintain compliance with General Design
Criterion (GDC) 57 for locating containment isolation valves as
close to the containment as practical.
There is no impact to Chapter 15 evaluations. Changes to the
PORV block valve and line size do not impact the mass releases to
the atmosphere during a Steam Generator Tube Rupture accident. The
mass release is limited by the PORV which is more restrictive than
the PORV block valve and line size.
There is no impact to any assumed leakage through the PORV line.
The existing 12-inch PORV has a design function to limit leakage
through the PORV line. Increasing the PORV block valve to 12 inches
will increase the leakage through the PORV block valve however it
will be that same leakage rate as the 12-inch PORV. Therefore, the
leakage rate through the PORV line does not increase and there is no
impact to radiation doses.
There is no impact to the assumptions or analysis in the
completed safety analysis for radiation doses as a result of the
change.
The piping analysis for the affected piping has been revised in
accordance with the requirements of the UFSAR. All stresses and
interface loads remain acceptable and within the limits described in
the UFSAR. The piping support calculations have been revised using
the load combinations prescribed in the UFSAR, and the critical
interaction ratio for each support is less than 1.0; therefore, a
positive design margin exists. The proposed changes did not affect
any of the piping packages chosen (as listed in the UFSAR) to
demonstrate piping design for piping design acceptance criteria
closure. There is no impact to the conclusions of the Pipe Rupture
Hazard Analysis (PRHA) because the PORV line is Break Exclusion Zone
(BEZ) piping. The proposed changes do not result in any new
postulated break locations. Updated analyses confirm that the
integrity of the wall adjacent to the MCR is unaffected by a
postulated main steam line break that causes the PORV line to impact
the wall. The piping and components downstream of the PORV are
nonsafety-related and are not affected by this activity.
The structural concrete floors and walls which make up the
bounds of the affected rooms were analyzed for the downstream
impacts due to the proposed changes. The results conclude that the
applicable acceptance criteria of the UFSAR are met. All applicable
load combinations shown in the UFSAR were considered. Critical
sections defined in the UFSAR within the scope of analysis remain
unchanged along with the typical reinforcement configuration
presented in the UFSAR. Therefore, all structural evaluations are
within the bounds of the acceptance criteria and meet the licensing
requirements imposed in the UFSAR.
There is no change to the valve motor operator. The current
motor operator is sufficient to operate the new 12-inch globe valve.
Therefore, there is no impact to the Class 1E dc and UPS System
(IDS) battery sizing. There is no change to the valve stroke time,
therefore there is no impact to valve open/closure times.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
[[Page 58618]]
NRC Branch Chief: Jennifer L. Dixon-Herrity.
Tennessee Valley Authority (TVA), Docket No. 50-391, Watts Bar Nuclear
Plant (WBN), Unit 2, Rhea County, Tennessee
Date of amendment request: May 14, 2018. A publicly available
version is in ADAMS under Accession No. ML18138A232.
Description of amendment request: The proposed amendment would
modify the WBN, Unit 2, Technical Specification (TS) 5.7.2.12, ``Steam
Generator (SG) Program,'' and TS 5.9.9, ``Steam Generator Tube
Inspection Report,'' to use the voltage-based alternate repair criteria
(ARC) specified in the guidelines contained in Generic Letter (GL) 95-
05, ``Voltage-Based Repair Criteria for Westinghouse Steam Generator
Tubes Affected by Outside Diameter Stress Corrosion Cracking.''
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.
Allowing the use of alternate repair criteria as proposed in
this amendment request does not involve a significant increase in
the probability or consequence of an accident previously evaluated.
Tube burst criteria are inherently satisfied during normal
operating conditions due to the proximity of the TSP [tube support
plates]. Test data indicates that tube burst cannot occur within the
TSP, even for tubes, which have 100% through-wall electric discharge
machining (EDM) notches, 0.75 inches long, provided that the TSP is
adjacent to the notched area. Because tube-to-tube support plate
proximity precludes tube burst during normal operating conditions,
use of the criteria must retain tube integrity characteristics,
which maintain a margin of safety of 1.4 times the bounding faulted
condition [i.e., main steam line break (MSLB)] differential pressure
of 2405 psig. GL 95-05 recommends that maintenance of a safety
factor of 1.4 times the MSLB pressure differential, consistent with
the structural limits in Regulatory Guide (RG) 1.121, on tube burst
is satisfied by 3/4-inch diameter tubing with bobbin coil
indications with signal amplitudes less than the tube structural
limit (VSL) of 6.03 volts, regardless of the indicated
depth measurement. At the FDB [flow distribution baffles], a safety
factor of three against the normal operating condition [Delta]P is
applied. A voltage of VSL = 3.81 volts satisfies the
burst capability recommendation at the FDB.
The upper voltage repair limit (VURL) will be
determined prior to each outage using the most recently approved NRC
database to determine the VSL. The structural limit is
reduced by allowances for nondestructive examination (NDE)
uncertainty (VNDE) and growth (VG) to
establish VURL.
Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated MSLB
outside of containment but upstream of the main steam isolation
valves (MSIVs) represents the most limiting radiological condition
relative to the alternate voltage-based repair criteria. In support
of implementation of the revised repair limit, TVA will determine
whether the distribution of cracking indications at the tube support
plate intersections during future cycles are projected to be such
that primary to secondary leakage would result in site boundary
doses within a fraction of the 10 CFR 100 guidelines or control room
doses within the 10 CFR 50, Appendix A, General Design Criterion
(GDC) 19 limit. A separate calculation has determined this allowable
MSLB leakage limit to be four gallons per minute (gpm) in the
faulted loop.
The methods for calculating the radiological dose consequences
for this postulated MSLB are consistent with the WBN dual-unit
Updated Final Safety Analysis Report (UFSAR) Chapter 15.
In summary, the calculated radiological consequences in the
control room and at the exclusion area boundary and the low
population zone are in compliance with the guidelines in the
Standard Review Plan, Chapter 15, and the regulations in 10 CFR 50,
Appendix A, GDC 19, and 10 CFR 100 reported for the postulated
steamline break event. Therefore, it is concluded that the proposed
changes do not result in a significant increase in the radiological
consequences of an accident previously analyzed.
Consistent with the guidance of GL 95-05, Section 2.c, the WBN
Unit 2 MSLB leak rate analysis would be performed, prior to
returning the SGs to service, based on either the projected next
end-of-cycle (EOC) voltage distribution or the actual measured
bobbin voltage distribution. The method to be used for the first
outage when ODSCC [outside diameter stress corrosion cracking]
indication growth rates are available will be based on the
indications found during that outage. As noted in GL 95-05, it may
not always be practical to complete EOC calculations prior to
returning the SGs to service. Under these circumstances, it is
acceptable to use the actual measured bobbin voltage distribution
instead of the projected EOC voltage distribution to determine
whether the reporting criteria are being satisfied.
Therefore, the voltage-based ARC at WBN Unit 2 does not
adversely affect SG tube integrity and implementation is shown to
result in acceptable radiological dose consequences. Therefore, the
proposed TS change does not result in a significant increase in the
probability or consequences of an accident previously evaluated
within the WBN Unit 2 UFSAR.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Implementation of the proposed SG tube voltage-based ARC does
not introduce any changes to the plant design basis. Neither a
single nor multiple tube rupture event would be expected in an SG in
which the repair limit has been applied (during all plant
conditions).
The bobbin probe voltage-based tube repair criteria of 1.0 volt
is supplemented by: enhanced eddy current inspection guidelines to
provide consistency in voltage normalization, a 100 percent eddy
current inspection sample size at the tube support plate elevations,
and rotating probe coil (RPC) or equivalent inspection requirements
for the larger indications left in service to characterize the
principal degradation as ODSCC.
As SG tube integrity upon implementation of the 1.0 volt repair
limit continues to be maintained through in-service inspection and
primary to secondary leakage monitoring, the possibility of a new or
different kind of accident from any accident previously evaluated is
not created.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The use of the voltage-based bobbin probe tube support plate
elevation repair criteria at WBN Unit 2 maintains SG tube integrity
commensurate with the guidance of RG 1.121. RG 1.121 describes a
method acceptable to the NRC for meeting GDCs 14, 15, and 32 by
reducing the probability or the consequences of SG tube rupture.
This reduction is accomplished by determining the limiting
conditions of degradation of steam generator tubing, as established
by in-service inspection, for which tubes with unacceptable cracking
should be removed from service. Upon implementation of the proposed
criteria, even under the worst-case conditions, the occurrence of
ODSCC at the TSP elevations is not expected to lead to an SG tube
rupture event during normal or faulted plant conditions. The EOC
distribution of crack indications at the tube support plate
elevations is confirmed to result in acceptable primary to secondary
leakage during all plant conditions and that radiological
consequences are not adversely impacted.
Implementation of the TSP intersection voltage-based repair
criteria will decrease the number of tubes that must be plugged. The
installation of SG tube plugs reduces the reactor coolant system
flow margin. Thus, implementation of the 1.0 volt repair limit will
maintain the margin of flow that would otherwise be reduced in the
event of increased tube plugging.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
[[Page 58619]]
NRC Branch Chief: Undine S. Shoop.
Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar
Nuclear Plant (WBN), Units 1 and 2, Rhea County, Tennessee
Date of amendment request: February 28, 2018. A publicly-available
version is in ADAMS under Accession No. ML18060A337.
Description of amendment request: The proposed amendments would
modify the WBN, Units 1 and 2, Technical Specification (TS) 3.8.9, to
add a new Condition C with an 8-hour completion for performing
maintenance on the opposite unit's vital bus when the opposite unit is
in Mode 5, Mode 6, or defueled. The proposed change would allow greater
operational flexibility for two-unit operation at WBN.
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 modifies the Required Actions for the
opposite unit's 120-volt (V) alternating current (AC) vital bus
system. This change will not affect the probability of an accident,
because the distribution system is not an initiator of any accident
sequence analyzed in the UFSAR [updated final safety analysis
report]. Rather, the opposite unit's distribution system support
equipment is used to mitigate accidents. The consequences of an
analyzed accident will not be significantly increased because the
minimum requirements for distribution systems will be maintained to
ensure the availability of the required power to mitigate accidents
assumed in the UFSAR. Operation in accordance with the proposed TS
will ensure that sufficient onsite electrical distribution systems
are operable as required to support the unit's required features.
Therefore, the mitigating functions supported by the onsite
electrical distribution systems will continue to provide the
protection assumed by the accident analysis. The integrity of
fission product barriers, plant configuration, and operating
procedures as described in the UFSAR will not be affected by the
proposed changes. Thus, the consequences of previously analyzed
accidents will not increase by implementing these changes.
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 modifies the Required Actions for the
opposite unit's 120V AC vital bus system. This change will not
physically alter the plant (no new or different type of equipment
will be installed). The proposed change will maintain the minimum
requirements for onsite electrical distribution systems to ensure
the availability of the equipment required to mitigate accidents
assumed in the UFSAR.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change modifies the Required Actions for the
opposite unit's 120V AC vital bus system. The margin of safety is
not affected by this change because the minimum requirements for
onsite electrical distribution systems will be maintained to ensure
the availability of the required power to shutdown the reactor and
maintain it in a safe shutdown condition after an AOO [anticipated
operational occurrence] or a postulated DBA [design-basis accident].
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Undine S. Shoop.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, Wake and Chatham Counties, North Carolina
Date of amendment request: June 28, 2017, as supplemented by
letters dated July 20 and September 14, 2017; and January 18, February
16, and April 13, 2018.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) for fuel storage criticality to account for the
use of neutron absorbing spent fuel pool rack inserts and soluble boron
for the purpose of criticality control in the boiling-water reactor
storage racks that currently credit Boraflex.
Date of issuance: October 22, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 167. A publicly-available version is in ADAMS under
Accession No. ML18204A286; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-63: The amendment
revised the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: December 5, 2017 (82 FR
57481). The supplemental letters dated July 20 and September 14, 2017;
and January 18, February 16, and April 13, 2018, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no
[[Page 58620]]
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 October 22, 2018.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: October 23, 2017, as supplemented by
letters dated November 15, 2017, and June 27, 2018.
Brief description of amendment: The amendment replaced the existing
Technical Specification (TS) requirements related to ``operations with
a potential for draining the reactor vessel'' (OPDRVs) with new
requirements on reactor pressure vessel (RPV) water inventory control
to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3 requires RPV
water level to be greater than the top of active irradiated fuel. The
changes are based on NRC-approved Technical Specifications Task Force
(TSTF) Traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water
Inventory Control.''
Date of issuance: October 30, 2018.
Effective date: As of its date of issuance and shall be implemented
at the beginning of the next refueling outage scheduled for May 2019.
Amendment No.: 251. A publicly-available version is in ADAMS under
Accession No. ML18255A350; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-21: The amendment
revised the Renewed Facility Operating License and TS.
Date of initial notice in Federal Register: January 16, 2018 (83 FR
2227). The supplemental letter dated June 27, 2018, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 30, 2018.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
1 (ANO-1), Pope County, Arkansas
Date of amendment request: October 2, 2017, as supplemented by
letters dated April 26 and August 10, 2018.
Brief description of amendment: The amendment revised the ANO-1
Technical Specification (TS) Bases for TS 3.7.5, ``Emergency Feedwater
(EFW) System,'' to identify the conditions in which TS 3.7.5, Condition
A, 7-day Completion Time (CT) and Condition C, 24-hour CT should apply
to the ANO-1 turbine-driven EFW pump steam supply valves.
Date of issuance: October 24, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 261. A publicly-available version is in ADAMS under
Accession No. ML18260A339; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-51: The amendment
revised the TS Bases.
Date of initial notice in Federal Register: December 5, 2017 (82 FR
57473). The supplemental letters dated April 26 and August 10, 2018,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 24, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station (Oyster Creek), Ocean County, New Jersey
Date of amendment request: November 16, 2017, as supplemented by
letter dated March 29, 2018.
Brief description of amendment: The amendment revised the Oyster
Creek Renewed Facility Operating License and the associated Technical
Specifications (TS) to Permanently Defueled Technical Specifications
consistent with the permanent cessation of operations and permanent
removal of fuel from the reactor vessel.
Date of issuance: October 26, 2018.
Effective date: The license amendment is effective on November 16,
2018, and shall be implemented in 60 days from the effective date.
Amendment No.: 295. A publicly-available version is in ADAMS under
Accession No. ML18227A338; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-16: The amendment
revised the Renewed Facility Operating License and TS.
Date of initial notice in Federal Register: January 16, 2018 (83 FR
2229). The supplemental letter dated March 29, 2018, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 26, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-244, R. E. Ginna Nuclear
Power Plant, Wayne County, New York
Date of amendment request: June 25, 2018, as supplemented by letter
dated August 29, 2018.
Brief description of amendment: The amendment revised the R. E.
Ginna Nuclear Power Plant's Technical Specification (TS) 3.1.4, ``Rod
Group Alignment Limits''; TS 3.1.5, ``Shutdown Bank Insertion Limit'';
TS 3.1.6, ``Control Bank Insertion Limits''; and TS 3.1.7, ``Rod
Position Indication,'' consistent with NRC-approved Technical
Specifications Task Force (TSTF) Traveler TSTF-547, Revision 1,
``Clarification of Rod Position Requirements,'' dated March 4, 2016.
Date of issuance: October 31, 2018.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 131. A publicly-available version is in ADAMS under
Accession No. ML18295A630; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-18: The amendment
revised the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: July 31, 2018 (83 FR
36976). The supplemental letter dated August 29, 2018, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's
[[Page 58621]]
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 October 31, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No. 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: April 25, 2018.
Brief description of amendments: The amendments revised the
Technical Specification (TS) requirements for inoperable snubbers for
each facility. The amendments also made other administrative changes to
the TS.
Date of issuance: October 29, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Clinton--220 (Unit 1); Dresden--259 (Unit 2), 252
(Unit 3); LaSalle--231 (Unit 1), 217 (Unit 2); and Quad Cities--271
(Unit 1), 266 (Unit 2). A publicly-available version is in ADAMS under
Accession No. ML18254A367. Documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Operating License Nos. NPF-62, DPR-19, DPR-25, NPF-11,
NPF-18, DPR-29, and DPR-30: The amendments revised the Facility
Operating Licenses and TS.
Date of initial notice in Federal Register: June 19, 2018 (83 FR
28460).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 29, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-277 and 50-278, Peach
Bottom Atomic Power Station, Units 2 and 3, York County, Pennsylvania
Date of amendment request: August 30, 2017, as supplemented by
letters dated October 24, 2017; and May 7, June 6, August 10, and
August 22, 2018.
Brief description of amendments: The amendments added a new license
condition to the Renewed Facility Operating Licenses to allow the
implementation of risk-informed categorization and treatment of
structures, systems, and components for nuclear power reactors in
accordance with 10 CFR 50.69.
Date of issuance: October 25, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 321 (Unit 2) and 324 (Unit 3). A publicly-available
version is in ADAMS under Accession No. ML18263A232; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the Renewed Facility Operating Licenses.
Date of initial notice in Federal Register: November 21, 2017 (82
FR 55404). The supplemental letters dated May 7, June 6, August 10, and
August 22, 2018, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the NRC staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 25, 2018.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Units 1 and 2 (Calvert Cliffs), Calvert
County, Maryland
Date of amendment request: February 25, 2016, as supplemented by
letters dated April 3, 2017, and January 11, January 18, June 21, and
August 27, 2018.
Brief description of amendments: The amendments revised the Calvert
Cliffs Technical Specifications (TS) related to completion times for
required actions to provide the option to calculate longer risk-
informed completion times. The amendments also added a new program, the
``Risk Informed Completion Time Program,'' to TS Section 5.5,
``Programs and Manuals.''
Date of issuance: October 30, 2018.
Effective date: As of the date of its issuance and shall be
implemented within 180 days.
Amendment Nos.: 326 (Unit 1) and 304 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML18270A130; documents related
to these amendments are listed in the safety evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-53 and DPR-69: The
amendments revised the Renewed Facility Operating Licenses and TS.
Date of initial notice in Federal Register: September 4, 2018 (83
FR 44920). The supplemental letter dated August 27, 2018, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a safety evaluation dated October 30, 2018.
No significant hazards consideration comments received: No.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: August 2, 2018.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) by removing Figure 5.1-1, ``Site Area
Map''; removing Technical Specification references to Figure 5.1-1; and
adding a site description.
Date of issuance: November 2, 2018.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 246 (Unit No. 1) and 197 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML18274A224;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-67 and NPF-16: The
amendments revised the Renewed Facility Operating Licenses and TS.
Date of initial notice in Federal Register: August 28, 2018 (83 FR
43905).
The Commission's related evaluation of the amendments is contained
in a
[[Page 58622]]
Safety Evaluation dated November 2, 2018.
No significant hazards consideration comments received: No.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center (DAEC), Linn County, Iowa
Date of amendment request: November 10, 2017.
Brief description of amendment: The amendment revised the Technical
Specifications (TS) for DAEC to adopt Technical Specifications Task
Force (TSTF) Traveler TSTF-551, Revision 3, ``Revise Secondary
Containment Surveillance Requirements,'' dated November 10, 2017 (ADAMS
Accession No. ML17318A240).
Date of issuance: October 31, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 307. A publicly-available version is in ADAMS under
Accession No. ML18241A383; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-49: The amendment
revised the Renewed Facility Operating License and TS.
Date of initial notice in Federal Register: February 27, 2018 (83
FR 8517).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 2018.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello
Nuclear Generating Plant (Monticello), Wright County, Minnesota
Date of amendment request: October 20, 2017, as supplemented by
letters dated June 1 and September 11, 2018.
Brief description of amendment: The amendment revised the
Monticello Technical Specification (TS) to adopt Technical
Specification Task Force (TSTF) Traveler TSTF-542, ``Reactor Pressure
Vessel Water Inventory Control.''
Date of issuance: October 29, 2018.
Effective date: As of the date of issuance and shall be implemented
prior to the next refueling outage.
Amendment No.: 198. A publicly-available version is in ADAMS under
Accession No. ML18250A075; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-22. The amendment
revised the Renewed Facility Operating License and TS.
Date of initial notice in Federal Register: December 19, 2017 (82
FR 60228). The supplemental letters dated June 1 and September 11,
2018, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the 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 October 29, 2018.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station
(Hope Creek), Salem County, New Jersey
Date of amendment request: September 21, 2017, as supplemented by
letters dated June 27, July 19, and September 6, 2018.
Brief description of amendment: The amendment revised the Hope
Creek Technical Specifications (TS) by replacing the existing
specifications related to ``operation with a potential for draining the
reactor vessel'' with revised requirements for reactor pressure vessel
water inventory control to protect Safety Limit 2.1.4. Safety Limit
2.1.4 requires reactor vessel water level to be greater than the top of
active irradiated fuel. The amendment adopted changes with variations,
as noted in the license amendment request, and is based on the NRC-
approved safety evaluation for Technical Specifications Task Force
(TSTF) Traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water
Inventory Control,'' dated December 20, 2016.
Date of issuance: October 30, 2018.
Effective date: As of the date of issuance and shall be implemented
prior to entering Operating Condition 4 for the next Hope Creek
refueling outage schedule for fall 2019 (H1R22).
Amendment No.: 213. A publicly-available version is in ADAMS under
Accession No. ML18260A203; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendments.
Renewed Facility Operating License No. NPF-57: The amendment
revised the Renewed Facility Operating License and TS.
Date of initial notice in Federal Register: January 30, 2018 (83 FR
4294). The supplemental letters dated June 27, July 19, and September
6, 2018, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the NRC staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 30, 2018.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2 (Vogtle), Burke
County, Georgia
Date of amendment request: September 12, 2017, as supplemented by
letter dated April 5, 2018.
Brief description of amendments: The amendments revised Technical
Specification (TS) 5.5.17, ``Containment Leakage Rate Testing
Program,'' for Vogtle to (1) increase the existing Type A integrated
leakage rate test interval from 10 to 15 years; (2) extend the Type C
containment isolation valve leaking testing to a 75-month frequency;
(3) adopt the use of American National Standards Institute/American
Nuclear Society 56.8-2002, ``Containment System Leakage Testing
Requirements''; and (4) adopt a more conservative grace interval for
Type A, B, and C tests.
Date of issuance: October 29, 2018.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 197 (Unit 1) and 180 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML18263A039; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-68 and NPF-81: The
amendments revised the Renewed Facility Operating Licenses and TS.
Date of initial notice in Federal Register: December 5, 2017 (82 FR
57474). The supplemental letter dated April 5, 2018, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 29, 2018.
No significant hazards consideration comments received: No.
[[Page 58623]]
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: April 13, 2018, as supplemented by
letter dated August 10, 2018.
Description of amendment: The amendment authorized changes to the
VEGP Units 3 and 4 Combined Operating License (COL) Appendix A,
Technical Specifications (TS). The amendment authorized departures from
associated Updated Final Safety Analysis Report information (which
includes the plant specific design control document Tier 2 information)
with changes which conform with the authorized TS changes.
Date of issuance: October 11, 2018.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 146 (Unit 3) and 145 (Unit 4). A publicly-available
version is in ADAMS under Accession No. ML18248A137; documents related
to this amendment are listed in the Safety Evaluation enclosed with the
amendment.
Facility Combined Licenses Nos. NPF-91 and NPF-92: The amendment
revised the Facility Combined Licenses and TS.
Date of initial notice in Federal Register: June 27, 2018 (83 FR
30199). The supplemental letter dated August 10, 2018, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in the Safety Evaluation dated October 11, 2018.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant,
Unit 2, Rhea County, Tennessee
Date of amendment request: October 11, 2017.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.3.1, Table 3.3.1-1, ``Reactor Trip System (RPS)
Instrumentation,'' to increase the values for the nominal trip setpoint
and the allowable value for Function 14.a, ``Turbine Trip - Low Fluid
Oil Pressure.'' The changes are due to the planned replacement and
relocation of the pressure switches from the low pressure auto-stop
trip fluid oil header to the high pressure turbine electrohydraulic
control (EHC) oil header. The changes are needed due to the higher EHC
system operating pressure.
Date of issuance: October 30, 2018.
Effective date: As of the date of issuance and shall be implemented
no later than startup from the Unit 2 refueling outage scheduled for
spring 2019.
Amendment No.: 22. A publicly-available version is in ADAMS under
Accession No. ML18255A156; documents related to the amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-96: The amendment revised the
Facility Operating License and TS.
Date of initial notice in Federal Register: March 13, 2018 (83 FR
10924).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 30, 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 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.
[[Page 58624]]
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-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
[[Page 58625]]
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.
Vistra Operations Company LLC, Docket Nos. 50-445 and 50-446, Comanche
Peak Nuclear Power Plant (CPNPP), Unit Nos. 1 and 2, Somervell County,
Texas
Date of amendment request: September 5, 2018, as supplemented by
letters dated September 20 and October 3, 2018.
Description of amendment: The amendments revised the CPNPP
Technical Specification (TS) 3.8.4, ``DC [Direct Current] Sources--
Operating,'' by adding a new REQUIRED ACTION to CONDITION B and an
extended COMPLETION TIME on a one-time basis to repair two affected
battery cells on the CPNPP Unit 1, Train B safety-related batteries.
Date of issuance: October 25, 2018.
Effective date: As of the date of issuance and shall be implemented
immediately as of its date of issuance.
Amendment Nos.: Unit 1--170; Unit 2--170. A publicly-available
version is in ADAMS under Accession No. ML18267A384; documents related
to the amendments are listed in the Safety Evaluation enclosed with the
amendments.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and TS.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes.
The license amendment request was originally noticed in the Federal
Register on September 18, 2018 (83 FR 47203). Subsequently, by letters
dated September 20 and October 3, 2018, the licensee provided
additional information that expanded the scope of the amendment request
as originally noticed in the Federal Register. Accordingly, on October
10, 2018 (83 FR 50971), the NRC published a second proposed NSHC
determination, which superseded the original notice in its
[[Page 58626]]
entirety. This included an individual 14-day notice for comments and
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided an opportunity to request a hearing by December 10, 2018, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendments.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a Safety Evaluation dated October 25, 2018.
Attorney for licensee: Timothy P. Matthews, Esq., Morgan, Lewis and
Bockius, 1111 Pennsylvania Avenue NW, Washington, DC 20004.
NRC Branch Chief: Robert J. Pascarelli.
Dated at Rockville, Maryland, this 8th day of November 2018.
For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2018-24894 Filed 11-19-18; 8:45 am]
BILLING CODE 7590-01-P