[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