[Federal Register Volume 82, Number 175 (Tuesday, September 12, 2017)]
[Notices]
[Pages 42844-42857]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-19214]



[[Page 42844]]

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NUCLEAR REGULATORY COMMISSION

[NRC-2017-0189]


Biweekly Notice: Applications and Amendments to Facility 
Operating Licenses and Combined Licenses Involving No Significant 
Hazards Considerations

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, 
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from August 15, 2017 to August 28, 2017. The 
last biweekly notice was published on August 29, 2017.

DATES: Comments must be filed by October 12, 2017. A request for a 
hearing must be filed by November 13, 2017.

ADDRESSES: You may submit comments by any of the following methods:
    <bullet> Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0189. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: <a href="/cdn-cgi/l/email-protection#e3a082918c8fcda4828f8f82848b8691a38d9180cd848c95"><span class="__cf_email__" data-cfemail="07446675686b2940666b6b66606f62754769756429606871">[email protected]</span></a>. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
    <bullet> Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: TWFN-8-D36M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Paula Blechman, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-2242; email: <a href="/cdn-cgi/l/email-protection#b6e6d7c3dad798f4dad3d5dedbd7d8f6d8c4d598d1d9c0"><span class="__cf_email__" data-cfemail="2979485c4548076b454c4a4144484769475b4a074e465f">[email protected]</span></a>.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2017-0189, facility name, unit 
numbers, plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
    <bullet> Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0189.
    <bullet> NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to <a href="/cdn-cgi/l/email-protection#552531277b2730263a20273630153b27367b323a23"><span class="__cf_email__" data-cfemail="1f6f7b6d316d7a6c706a6d7c7a5f716d7c31787069">[email protected]</span></a>. 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.
    <bullet> NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2017-0189, facility name, unit 
numbers, plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

II. Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses and Combined Licenses and Proposed No Significant 
Hazards Consideration Determination

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated, or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment 
period such that failure to act in a timely way would result, for 
example in derating or shutdown of the facility. If the Commission 
takes action prior to the expiration of either the comment period or 
the notice period, it will publish in the Federal Register a notice of 
issuance. If the Commission makes a final no significant hazards 
consideration determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

[[Page 42845]]

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if 
appropriate, a notice of a hearing will be issued.
    As required by 10 CFR 2.309(d) the petition should specifically 
explain the reasons why intervention should be permitted with 
particular reference to the following general requirements for 
standing: (1) The name, address, and telephone number of the 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
petitioner's property, financial, or other interest in the proceeding; 
and (4) the possible effect of any decision or order which may be 
entered in the proceeding on the petitioner's interest.
    In accordance with 10 CFR 2.309(f), the petition must also set 
forth the specific contentions which the petitioner seeks to have 
litigated in the proceeding. Each contention must consist of a specific 
statement of the issue of law or fact to be raised or controverted. In 
addition, the petitioner must provide a brief explanation of the bases 
for the contention and a concise statement of the alleged facts or 
expert opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to the specific sources and 
documents on which the petitioner intends to rely to support its 
position on the issue. The petition must include sufficient information 
to show that a genuine dispute exists with the applicant or licensee on 
a material issue of law or fact. Contentions must be limited to matters 
within the scope of the proceeding. The contention must be one which, 
if proven, would entitle the petitioner to relief. A petitioner who 
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene. 
Parties have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that party's admitted 
contentions, including the opportunity to present evidence, consistent 
with the NRC's regulations, policies, and procedures.
    Petitions must be filed no later than 60 days from the date of 
publication of this notice. Petitions and motions for leave to file new 
or amended contentions that are filed after the deadline will not be 
entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in 
accordance with the filing instructions in the ``Electronic Submissions 
(E-Filing)'' section of this document.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to establish when the hearing is held. If the final determination is 
that the amendment request involves no significant hazards 
consideration, the Commission may issue the amendment and make it 
immediately effective, notwithstanding the request for a hearing. Any 
hearing would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of the amendment unless the Commission finds an imminent 
danger to the health or safety of the public, in which case it will 
issue an appropriate order or rule under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission no later 
than 60 days from the date of publication of this notice. The petition 
must be filed in accordance with the filing instructions in the 
``Electronic Submissions (E-Filing)'' section of this document, and 
should meet the requirements for petitions set forth in this section, 
except that under 10 CFR 2.309(h)(2) a State, local governmental body, 
or Federally-recognized Indian Tribe, or agency thereof does not need 
to address the standing requirements in 10 CFR 2.309(d) if the facility 
is located within its boundaries. Alternatively, a State, local 
governmental body, Federally-recognized Indian Tribe, or agency thereof 
may participate as a non-party under 10 CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
the discretion of the presiding officer, be permitted to make a limited 
appearance pursuant to the provisions of 10 CFR 2.315(a). A person 
making a limited appearance may make an oral or written statement of 
his or her position on the issues but may not otherwise participate in 
the proceeding. A limited appearance may be made at any session of the 
hearing or at any prehearing conference, subject to the limits and 
conditions as may be imposed by the presiding officer. Details 
regarding the opportunity to make a limited appearance will be provided 
by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate under 10 CFR 2.315(c), must be filed in accordance with the 
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 
46562, August 3, 2012). The E-Filing process requires participants to 
submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC's Web site at 
http://www.nrc.gov/site-help/e-submittals.html. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at

[[Page 42846]]

<a href="/cdn-cgi/l/email-protection#88e0ede9fae1e6efa6ece7ebe3edfcc8e6faeba6efe7fe"><span class="__cf_email__" data-cfemail="4b232e2a3922252c652f2428202e3f0b253928652c243d">[email protected]</span></a>, or by telephone at 301-415-1677, to (1) request 
a digital identification (ID) certificate, which allows the participant 
(or its counsel or representative) to digitally sign submissions and 
access the E-Filing system for any proceeding in which it is 
participating; and (2) advise the Secretary that the participant will 
be submitting a petition or other adjudicatory document (even in 
instances in which the participant, or its counsel or representative, 
already holds an NRC-issued digital ID certificate). Based upon this 
information, the Secretary will establish an electronic docket for the 
hearing in this proceeding if the Secretary has not already established 
an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a 
digital ID certificate and a docket has been created, the participant 
can then submit adjudicatory documents. Submissions must be in Portable 
Document Format (PDF). Additional guidance on PDF submissions is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the 
time the document is submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
document on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before adjudicatory documents are 
filed so that they can obtain access to the documents via the E-Filing 
system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public Web site at http://www.nrc.gov/site-help/e-submittals.html, by 
email to <a href="/cdn-cgi/l/email-protection#e8a5bba0acc6ba8d9b879d9a8b8da8869a8bc68f879e"><span class="__cf_email__" data-cfemail="eaa7b9a2aec4b88f99859f98898faa849889c48d859c">[email protected]</span></a>, or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 6 p.m., Eastern Time, Monday through Friday, excluding government 
holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
stating why there is good cause for not filing electronically and 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff. Participants filing adjudicatory documents in this 
manner are responsible for serving the document on all other 
participants. Filing is considered complete by first-class mail as of 
the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service. A presiding officer, having granted an 
exemption request from using E-Filing, may require a participant or 
party to use E-Filing if the presiding officer subsequently determines 
that the reason for granting the exemption from use of E-Filing no 
longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the 
Commission or the presiding officer. If you do not have an NRC-issued 
digital ID certificate as described above, click cancel when the link 
requests certificates and you will be automatically directed to the 
NRC's electronic hearing dockets where you will be able to access any 
publicly-available documents in a particular hearing docket. 
Participants are requested not to include personal privacy information, 
such as social security numbers, home addresses, or personal phone 
numbers in their filings, unless an NRC regulation or other law 
requires submission of such information. For example, in some 
instances, individuals provide home addresses in order to demonstrate 
proximity to a facility or site. With respect to copyrighted works, 
except for limited excerpts that serve the purpose of the adjudicatory 
filings and would constitute a Fair Use application, participants are 
requested not to include copyrighted materials in their submission.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick 
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina
    Date of amendment request: June 29, 2017. A publicly available 
version is in ADAMS under Accession No. ML17180A538.
    Description of amendment request: The amendments would adopt 
changes, with variations, based on the NRC-approved safety evaluation 
of Technical Specifications Task Force (TSTF) Traveler TSTF-542, 
Revision 2, ``Reactor Pressure Vessel Water Inventory Control,'' dated 
December 20, 2016 (ADAMS Package Accession No. ML16343B066). The 
revisions would replace existing technical specification (TS) 
requirements related to ``operations with a potential for draining the 
reactor vessel'' (OPDRVs) with new requirements on reactor pressure 
vessel water inventory control (RPV WIC) to protect Safety Limit 
2.1.1.3, which requires reactor vessel water level to be greater than 
the top of active irradiated fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. Draining of RPV [reactor pressure vessel] water 
inventory in Mode 4 (i.e., cold shutdown) and Mode 5 (i.e., 
refueling) is not an accident previously evaluated and, therefore, 
replacing the existing TS controls to prevent or mitigate such an 
event with a new set of controls has no effect on any accident 
previously evaluated. RPV water inventory control in Mode 4 or Mode 
5 is not an initiator of any accident previously evaluated. The 
existing OPDRV controls or the proposed RPV WIC controls are not 
mitigating actions assumed in any accident previously evaluated.

[[Page 42847]]

    The proposed change reduces the probability of an unexpected 
draining event, which is not a previously evaluated accident, by 
imposing new requirements on the limiting time in which an 
unexpected draining event could result in the reactor vessel water 
level dropping to the top of the active fuel (TAF). These controls 
require cognizance of the plant configuration and control of 
configurations with unacceptably short drain times. These 
requirements reduce the probability of an unexpected draining event. 
The current TS requirements are only mitigating actions and impose 
no requirements that reduce the probability of an unexpected 
draining event. The proposed change reduces the consequences of an 
unexpected draining event, which is not a previously evaluated 
accident, by requiring an Emergency Core Cooling System (ECCS) 
subsystem to be operable at all times in Modes 4 and 5. The current 
TS requirements do not require any water injection systems, ECCS or 
otherwise, to be operable in certain conditions in Mode 5. The 
change in requirement from two ECCS subsystems to one ECCS subsystem 
in Modes 4 and 5 does not significantly affect the consequences of 
an unexpected draining event because the proposed Actions ensure 
equipment is available within the limiting drain time that is as 
capable of mitigating the event as the current requirements. The 
proposed controls provide escalating compensatory measures to be 
established as calculated drain times decrease, such as verification 
of a second method of water injection and additional confirmations 
that containment and/or filtration would be available if needed. The 
proposed change reduces or eliminates some requirements that were 
determined to be unnecessary to manage the consequences of an 
unexpected draining event, such as automatic initiation of an ECCS 
subsystem and control room ventilation. These changes do not affect 
the consequences of any accident previously evaluated since a 
draining event in Modes 4 and 5 is not a previously evaluated 
accident and the requirements are not needed to adequately respond 
to a draining event.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. The proposed change will not alter the design 
function of the equipment involved. Under the proposed change, some 
systems that are currently required to be operable during OPDRVs 
would be required to be available within the limiting drain time or 
to be in service depending on the limiting drain time. Should those 
systems be unable to be placed into service, the consequences are no 
different than if those systems were unable to perform their 
function under the current TS requirements. The event of concern 
under the current requirements and the proposed change is an 
unexpected draining event. The proposed change does not create new 
failure mechanisms, malfunctions, or accident initiators that would 
cause a draining event or a new or different kind of accident not 
previously evaluated or included in the design and licensing bases.
    Thus, based on the above, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC. The current requirements do 
not have a stated safety basis and no margin of safety is 
established in the licensing basis. The safety basis for the new 
requirements is to protect Safety Limit 2.1.1.3. New requirements 
are added to determine the limiting time in which the RPV water 
inventory could drain to the top of the fuel in the reactor vessel, 
should an unexpected draining event occur. Plant configurations that 
could result in lowering the RPV water level to the TAF within one 
hour are now prohibited. New escalating compensatory measures based 
on the limiting drain time replace the current controls. The 
proposed TS establish a safety margin by providing defense-in-depth 
to ensure that the Safety Limit is protected and to protect the 
public health and safety. While some less restrictive requirements 
are proposed for plant configurations with long calculated drain 
times, the overall effect of the change is to improve plant safety 
and to add safety margin.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, 
550 South Tryon Street, M/C DEC45A, Charlotte, NC 28202.
    NRC Branch Chief: Undine Shoop.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant (PNP), Van Buren County, Michigan
    Date of amendment request: July 27, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17208A428.
    Description of amendment request: The proposed amendment would 
revise certain staffing and training requirements, reports, programs, 
and editorial changes in the Technical Specifications (TSs) Table of 
Contents; Section 1.0, ``Use and Application''; and Section 5.0, 
``Administrative Controls,'' that will no longer be applicable once PNP 
is permanently defueled.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would not take effect until the PNP 
Certified Fuel Handler Training and Retraining Program has been 
approved by the NRC, and PNP has permanently ceased operation and 
entered a permanently defueled condition. The proposed changes would 
revise the PNP TS by modifying the definitions, in TS Section 1.0, 
and administrative controls, in TS Section 5.0, to correspond to the 
permanently defueled condition. Additionally, certain portions of 
the administrative control sections are deleted because they are no 
longer applicable to a permanently defueled facility.
    The proposed deletion and modification of provisions of the 
administrative controls do not directly affect the design of 
structures, systems, and components (SSCs) necessary for safe 
storage of spent nuclear fuel or the methods used for handling and 
storage of such fuel in the spent fuel pool (SFP). The proposed 
changes to the administrative controls are administrative in nature 
and do not affect any accidents applicable to the safe management of 
spent nuclear fuel or the permanently shutdown and defueled 
condition of the reactor. Thus, the consequences of an accident 
previously evaluated are not increased.
    In a permanently defueled condition, the only credible accidents 
are the fuel handling accident (FHA), the failure of tanks 
containing radioactive liquids, and a spent fuel cask drop accident. 
The probability of occurrence of previously evaluated accidents is 
not increased, because extended operation in a permanently defueled 
condition will be the only operation allowed. This mode of operation 
is bounded by the existing analyses. Additionally, the occurrence of 
postulated accidents associated with reactor operation are no longer 
credible in a permanently defueled reactor. This significantly 
reduces the scope of applicable accidents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment has no impact on facility systems, 
structures, and components (SSCs) affecting the safe storage of 
spent nuclear fuel, or on the methods of operation of such SSCs, or 
on the handling

[[Page 42848]]

and storage of spent nuclear fuel itself. The proposed amendment 
does not result in different or more adverse failure modes or 
accidents than previously evaluated because the reactor will be 
permanently shutdown and defueled, and PNP will no longer be 
authorized to operate the reactor or retain or place fuel in the 
reactor vessel.
    The proposed amendment does not affect systems credited in the 
PNP accident analysis for a[n] FHA, or for mitigating accident 
releases from the failure of tanks containing radioactive liquids or 
from a spent fuel cask drop. The proposed changes will continue to 
require proper control and monitoring of safety significant 
parameters and activities.
    The proposed amendment does not result in any new mechanisms 
that could damage the remaining relevant safety barriers that 
support maintaining the plant in a permanently shutdown and defueled 
condition (e.g., fuel cladding and SFP cooling). Since extended 
operation in a defueled condition will be the only operation 
allowed, and this condition is bounded by existing analyses, such a 
condition does not create the possibility of a new or different kind 
of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment involves deleting and/or modifying 
certain TS requirements once the PNP has been permanently shutdown 
and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 
license for PNP will no longer authorize operation of the reactor or 
emplacement or retention of fuel into the reactor vessel following 
submittal of the certifications required by 10 CFR 50.82(a)(1). 
Therefore, the occurrence of postulated accidents associated with 
reactor operation are no longer credible.
    The only remaining credible accidents are the fuel handling 
accident (FHA), the failure of tanks containing radioactive liquids, 
and a spent fuel cask drop accident. The proposed amendment does not 
adversely affect the inputs or assumptions of any of the design 
basis analyses that impact these analyzed conditions.
    The proposed changes are limited to those portions of the TS 
that are not related to the SSCs that are important to the safe 
storage of spent nuclear fuel. The requirements that are proposed to 
be revised or deleted from the PNP TS are not credited in the 
existing accident analysis for the remaining applicable postulated 
accidents, and as such, do not contribute to the margin of safety 
associated with the accident analysis. Postulated design basis 
accidents involving the reactor are no longer possible because the 
reactor will be permanently shutdown and defueled, and PNP will no 
longer be authorized to operate the reactor or retain or place fuel 
in the reactor vessel.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. William Dennis, Assistant General 
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White 
Plains, NY 10601.
    NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No. 1, DeWitt County, Illinois
    Date of amendment request: July 18, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17199F854.
    Description of amendment request: The proposed change would revise 
the design value for the spent fuel storage pool in Technical 
Specification (TS) 4.3.2, ``Drainage,'' to an appropriate value, 
consistent with the original design basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    No physical changes to the facility will occur as a result of 
this proposed amendment. The proposed changes will not alter the 
physical design. The proposed change will revise the current TS 
4.3.2 value for the SFP [spent fuel pool] level design to be 
consistent with the original design basis value and the applicable 
regulatory requirements. The proposed value will continue to ensure 
that inadvertent draining of the SFP will not result in the 
uncovering of spent fuel, as well as provide adequate shielding for 
personnel protection.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not alter the physical design, safety 
limits, or safety analysis assumptions associated with the operation 
of the plant. Accordingly, the change does not introduce any new 
accident initiators, nor does it reduce or adversely affect the 
capabilities of any plant structure, system, or component to perform 
their safety function. The proposed change will revise the current 
TS 4.3.2 value for the SFP level design to be consistent with the 
original design basis value and the applicable regulatory 
requirements. The proposed value will continue to ensure that 
inadvertent draining of the SFP will not result in the uncovering of 
spent fuel, as well as provide adequate shielding for personnel 
protection.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change conforms to NRC regulatory guidance 
regarding the content of plant Technical Specifications. The 
proposed change does not alter the physical design, safety limits, 
or safety analysis assumptions associated with the operation of the 
plant.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: David J. Wrona.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
    Date of amendment request: July 19, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17200D096.
    Description of amendment request: The amendments would replace 
existing technical specification (TS) requirements related to 
``operations with a potential for draining the reactor vessel'' 
(OPDRVs) with new requirements on reactor pressure vessel (RPV) water 
inventory control (WIC) to protect Safety Limit 2.1.4. Safety Limit 
2.1.4 requires RPV water level to be greater than the top of active 
irradiated fuel. The proposed changes are based on Technical 
Specifications Task Force (TSTF) Traveler TSTF-542, ``Reactor Pressure 
Vessel Water Inventory Control,'' Revision 2 (ADAMS Package Accession 
No. ML16250A231).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 42849]]


    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.4. Draining of RPV water inventory in OPERATIONAL 
CONDITION 4 (i.e., cold shutdown) and OPERATIONAL CONDITION 5 (i.e., 
refueling), is not an accident previously evaluated and, therefore, 
replacing the existing TS controls to prevent or mitigate such an 
event with a new set of controls has no effect on any accident 
previously evaluated. RPV water inventory control in OPERATIONAL 
CONDITION 4 or 5 is not an initiator of any accident previously 
evaluated. The existing OPDRV controls or the proposed RPV WIC 
controls are not mitigating actions assumed in any accident 
previously evaluated.
    The proposed changes reduce the probability of an unexpected 
draining event (which is not a previously evaluated accident) by 
imposing new requirements on the limiting time in which an 
unexpected draining event could result in the reactor vessel water 
level dropping to the top of the active fuel (TAF). These controls 
require cognizance of the plant configuration and control of 
configurations with unacceptably short drain times. These 
requirements reduce the probability of an unexpected draining event. 
The current TS requirements are only mitigating actions and impose 
no requirements that reduce the probability of an unexpected 
draining event.
    The proposed changes reduce the consequences of an unexpected 
draining event (which is not a previously evaluated accident) by 
requiring an Emergency Core Cooling System (ECCS) subsystem to be 
operable at all times in OPERATIONAL CONDITIONS 4 and 5. The current 
TS requirements do not require any water injection systems, ECCS or 
otherwise, to be Operable in certain conditions in OPERATIONAL 
CONDITION 5. The change in requirement from two ECCS subsystems to 
one ECCS subsystem in OPERATIONAL CONDITIONS 4 and 5 does not 
significantly affect the consequences of an unexpected draining 
event because the proposed Actions ensure equipment is available 
within the limiting drain time that is as capable of mitigating the 
event as the current requirements. The proposed controls provide 
escalating compensatory measures to be established as calculated 
drain times decrease, such as verification of a second method of 
water injection and additional confirmations that containment and/or 
filtration would be available if needed.
    The proposed changes reduce or eliminate some requirements that 
were determined to be unnecessary to manage the consequences of an 
unexpected draining event, such as automatic initiation of an ECCS 
subsystem and control room ventilation. These changes do not affect 
the consequences of any accident previously evaluated since a 
draining event in OPERATIONAL CONDITIONS 4 and 5 is not a previously 
evaluated accident and the requirements are not needed to adequately 
respond to a draining event.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.4. The proposed changes will not alter the design function 
of the equipment involved. Under the proposed changes, some systems 
that are currently required to be operable during OPDRVs would be 
required to be available within the limiting drain time or to be in 
service depending on the limiting drain time. Should those systems 
be unable to be placed into service, the consequences are no 
different than if those systems were unable to perform their 
function under the current TS requirements.
    The event of concern under the current requirements and the 
proposed changes is an unexpected draining event. The proposed 
changes do not create new failure mechanisms, malfunctions, or 
accident initiators that would cause a draining event or a new or 
different kind of accident not previously evaluated or included in 
the design and licensing bases.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC. The current requirements do 
not have a stated safety basis and no margin of safety is 
established in the licensing basis. The safety basis for the new 
requirements is to protect Safety Limit 2.1.4. New requirements are 
added to determine the limiting time in which the RPV water 
inventory could drain to the TAF in the reactor vessel should an 
unexpected draining event occur. Plant configurations that could 
result in lowering the RPV water level to the TAF within one hour 
are now prohibited. New escalating compensatory measures based on 
the limiting drain time replace the current controls. The proposed 
TS establish a safety margin by providing defense-in-depth to ensure 
that the Safety Limit is protected and to protect the public health 
and safety. While some less restrictive requirements are proposed 
for plant configurations with long calculated drain times, the 
overall effect of the change is to improve plant safety and to add 
safety margin.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: James G. Danna.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
    Date of amendment request: January 23, 2017, as supplemented by 
letter dated July 3, 2017. Publicly-available versions are in ADAMS 
under Accession Nos. ML17025A399 and ML17184A176, respectively.
    Description of amendment request: The license amendment request was 
originally noticed in the Federal Register on March 28, 2017 (82 FR 
15383). The notice is being reissued in its entirety to include the 
revised scope, description of the amendment request, and proposed no 
significant hazards consideration determination. As a result of the 
revised scope, updates to the ``Basis for proposed no significant 
hazards consideration determination'' section of this notice are 
delineated by brackets.
    The amendments would modify the Technical Specifications (TSs) by 
limiting the MODE of applicability for the Reactor Protection System 
(RPS), Startup, and Operating Rate of Change of Power--High, functional 
unit trip. Additionally, the proposed amendments add new Limiting 
Condition for Operation (LCO) 3.0.5 and relatedly modifies LCO 3.0.1 
and LCO 3.0.2, to provide for placing inoperable equipment under 
administrative control for the purpose of conducting testing required 
to demonstrate OPERABILITY.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Limiting the MODE 1 applicability for RPS functional unit, 
Startup and Operating Rate of Change of Power--High, to Power Range 
Neutron Flux Power <=15% of RATED THERMAL POWER, is an 
administrative change in nature and does not alter the manner in 
which the functional unit is operated or maintained. The proposed 
changes do not represent any physical

[[Page 42850]]

change to plant [structures, systems, and components (SSC(s))], or 
to procedures established for plant operation. The subject RPS 
functional unit is not an event initiator nor is it credited in the 
mitigation of any event or credited in the [probabilistic risk 
assessment (PRA)]. As such, the initial conditions associated with 
accidents previously evaluated and plant systems credited for 
mitigating the consequences of accidents previously evaluated remain 
unchanged.
    The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 
and Unit 2 TS and related modification to [LCO 3.0.1 and] LCO 3.0.2 
is consistent with the guidance provided in NUREG-1432, Volume 1 
[ADAMS Accession No. ML12102A165] (Reference 6.1 [of the amendment 
request]) and thereby has been previously evaluated by the 
Commission with a determination that the proposed change does not 
involve a significant hazards consideration.
    Therefore, facility operation in accordance with the proposed 
license amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Limiting the MODE 1 applicability for the RPS functional unit, 
Startup and Operating Rate of Change of Power--High, to Power Range 
Neutron Flux Power <=15% of RATED THERMAL POWER, is an 
administrative change in nature and does not involve the addition of 
any plant equipment, methodology or analyses. The proposed changes 
do not alter the design, configuration, or method of operation of 
the subject RPS functional unit or of any other SSC. More 
specifically, the proposed changes neither alter the power rate-of-
change trip function nor its ability to bypass and reset as 
required. The subject RPS functional unit remains capable of 
performing its design function.
    The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 
and Unit 2 TS and related modification to [LCO 3.0.1 and] LCO 3.0.2 
is consistent with the guidance provided in NUREG-1432, Volume 1 
(Reference 6.1 [of the amendment request]) and thereby has been 
previously evaluated by the Commission with a determination that the 
proposed change does not involve a significant hazards 
consideration.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Limiting the MODE 1 applicability for RPS functional unit, 
Startup and Operating Rate of Change of Power--High, to Power Range 
Neutron Flux Power <=15% of RATED THERMAL POWER is an administrative 
change in nature. The proposed changes neither involve changes to 
any safety analyses assumptions, safety limits, or limiting safety 
system settings nor do they adversely impact plant operating margins 
or the reliability of equipment credited in safety analyses.
    The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 
and Unit 2 TS and related modification to [LCO 3.0.1 and] LCO 3.0.2 
is consistent with the guidance provided in NUREG-1432, Volume 1 
(Reference 6.1 [of the amendment request]) and thereby has been 
previously evaluated by the Commission with a determination that the 
proposed change does not involve a significant hazards 
consideration.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Undine Shoop.
Florida Power & Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County, Florida
    Date of amendment request: June 29, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17195A569.
    Description of amendment request: The amendments would modify the 
Technical Specification (TS) requirements for mode change limitations 
in TS 3.0.4 and TS 4.0.4 based on Technical Specifications Tasks Force 
(TSTF) Improved Standard Technical Specifications Change Traveler, 
TSTF-359, Revision 9, ``Increase Flexibility in MODE Restraints'' 
(ADAMS Accession No. ML031190607).
    The NRC issued a notice of opportunity for comment in the Federal 
Register on August 2, 2002 (67 FR 50475), on possible amendments 
concerning TSTF-359, including a model safety evaluation and model no 
significant hazards consideration determination, using the consolidated 
line item improvement process (CLIIP). Subsequently, on April 4, 2003, 
the NRC published a Notice of Availability for TSTF-359, Revision 8, in 
the Federal Register (68 FR 16579). That notice announced the 
availability of this TS improvement through the CLIIP. The NRC 
subsequently made two modifications in response to comments, as well as 
one editorial change, which have been incorporated into TSTF-359, 
Revision 9. The changes proposed in the licensee's submittal are, 
therefore, based on TSTF-359, Revision 9.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS Action. Being 
in a TS Action is not an initiator of any accident previously 
evaluated. Therefore, the probability of an accident previously 
evaluated is not significantly increased. The consequences of an 
accident while relying on Actions as allowed by the proposed LCO 
3.0.4 are no different than the consequences of an accident while 
relying on Actions for other reasons, such as equipment 
inoperability. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS while in a TS Action, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS while in a TS Action. The TS 
allow operation of the plant without the full complement of 
equipment through the Actions for not meeting the TS Limiting 
Conditions for Operation (LCO). The risk associated with this 
allowance is managed by the imposition of Actions that must be 
performed within the prescribed completion times. The net effect of 
being in a TS Action on the margin of safety is not considered

[[Page 42851]]

significant. The proposed change does not alter the required actions 
or completion times of the TS. The proposed change allows TS Actions 
to be entered and the associated required actions and completion 
times to be used in new circumstances. This use is predicated upon 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing Actions 
in similar circumstances without assessing and managing risk. The 
net change to the margin of safety is insignificant. Therefore, this 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Undine Shoop.
National Institute of Standards and Technology (NIST), Docket No. 50-
184, Center for Neutron Research Test Reactor, Montgomery County, 
Maryland
    Date of amendment request: March 2, 2017 (two letters), as 
supplemented by letters dated March 29, 2017, and May 25, 2017. 
Publicly-available versions are in ADAMS under Accession Nos. 
ML17068A163, ML17068A164, ML17097A243, and ML17153A172, respectively.
    Description of amendment request: The proposed amendment would 
modify the NIST test reactor's technical specifications (TSs) to remove 
limitations in the present version of the TSs that prohibit use of a 
test procedure and to change the organizational chart in the TSs. In 
addition, the proposed amendment would modify the NIST test reactor's 
license to allow transfer of instrumentation calibration and testing 
sources from the NIST's material license to the reactor license.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    No, the proposed amendment would not increase the probability or 
consequences of an accident previously evaluated. The proposed 
amendment removes conformance conflicts within the Technical 
Specifications that would occur when operating the reactor as 
permitted under TSs 2.2(4). The conflicts are removed from the TSs 
by adding exception statements. When the reactor is operated under 
the NRC approved conditions in TSs 2.2(4), steady state thermal 
hydraulic analysis shows that operation at less than 500 kW 
[kilowatt] with natural circulation results in a critical heat flux 
ratio and onset of flow instability ratio greater than 2. Transient 
analysis of reactivity insertion accidents shows that the fuel 
cladding temperature remains far below the safety limit. The limit 
of 10 kw was chosen since that was deemed adequate for any 
operational situation requiring natural circulation operation, such 
as testing of an unknown core loading.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    No, the proposed amendment would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed amendment removes conformance conflicts 
within the Technical Specifications that would occur when operating 
the reactor as permitted under TSs 2.2(4). The conflicts are removed 
from the TSs by adding exception statements. The accident analysis 
was discussed in the document, NIST Response to NRC Request for 
Information (TAC No. MD3410), August 19, 2008, ADAMS Accession 
Number ML082890338. The request from the NRC was: ``. . . Provide 
justification for 500 kW power operations under natural convection 
flow by demonstrating that no credible accidents would result in 
exceeding the safety limit. . . ,'' the following was the response 
by NIST. ``This analysis shows that there is ample margin between 
the maximum clad temperature in any credible accident and the safety 
limit of 450 [deg]C [degrees Centigrade].'' The details of the 
analysis are presented in the above reference.
    The intent with this amendment is to allow, without apparent TSs 
nonconformance, operation analyzed and evaluated by the NRC. This 
will allow the use of testing similar to that which was performed in 
the commissioning of NIST test reactor.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    No, the proposed amendment would not involve a significant 
reduction in a margin of safety. This amendment will allow testing 
when commissioning a core configuration that is unknown in the most 
conservative manner appropriate. It removes apparent TS conflicts 
that would force the licensee into situations that would be less 
conservative and with less margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Melissa J. Lieberman, Deputy Chief Counsel 
for NIST, National Institute of Standards and Technology, 100 Bureau 
Drive, Gaithersburg, MD 20899.
    NRC Branch Chief: Alexander Adams, Jr.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: April 21, 2017, as supplemented by 
letter dated August 15, 2017. Publicly-available versions are in ADAMS 
under Accession Nos. ML17111A958, and ML17227A775, respectively.
    Description of amendment request: The amendment request proposes to 
depart from approved AP1000 Design Control Document (DCD) Tier 2 
information (text, tables and figures) as incorporated into the Updated 
Final Safety Analysis Report (UFSAR) as plant-specific DCD information, 
and also proposes to depart from involved plant-specific Tier 1 
information (and associated Combined License (COL) Appendix C 
information). Specifically, the amendment request proposes changes to 
COL Appendix C (and plant-specific Tier 1) Table 2.2.4-1 and Figure 
2.2.4-1 to add two main feedwater thermal relief valves and two start-
up feedwater thermal relief valves. The proposed COL Appendix C (and 
plant-specific DCD Tier 1) changes require additional changes to 
corresponding Tier 2 information in UFSAR Chapters 3 and 10. Because 
this proposed change requires a departure from Tier 1 information in 
the Westinghouse Electric Company's AP1000 DCD, the licensee also 
requested an exemption from the requirements of the Generic DCD Tier 1 
in accordance with 10 CFR 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The changes to Combined License (COL) Appendix C (and plant-
specific Tier 1) Table 2.2.4-1 and Figure 2.2.4-1, and associated 
Updated Final Safety Analysis Report (USFAR) design information do 
not adversely impact previously evaluated accidents. The addition of 
the thermal relief valves to the feedwater lines does not adversely 
impact the ability to isolate the main and startup feedwater lines 
following a steam or

[[Page 42852]]

feedwater line break or steam generator tube rupture. The new 
thermal relief valves are normally closed and required to open to 
prevent potential overpressure conditions when ambient temperatures 
increase in the area. Thermal relief valves added into the feedwater 
lines operate mechanically and are not activated upon a new 
engineered safety features (ESF) signal in response to design basis 
accidents. Isolation capabilities of the main and startup feedwater 
lines are not adversely affected as ESF signals are not changed. The 
proposed change does not reduce the temperature of feedwater and 
does not increase feedwater flow during any operational mode as main 
feedwater and startup feedwater isolation and control valves are not 
changed by this activity. Performance of overpressure relief 
supports the safety-related functions of the isolation and control 
valves in the main and startup feedwater lines when isolation is 
required.
    No safety-related structure, system, component (SSC) or function 
is adversely affected by this change. The change does not involve an 
interface with any SSC accident initiator or initiating sequence of 
events, and thus, the probabilities of the accidents evaluated in 
the plant-specific UFSAR are not affected. The proposed changes do 
not involve a change to the predicted radiological releases due to 
postulated accident conditions, thus, the consequences of the 
accidents evaluated in the UFSAR are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to COL Appendix C (and plant-specific Tier 
1) Table 2.2.4-1 and Figure 2.2.4-1, and associated UFSAR design 
information do not reduce the temperature of feedwater and do not 
increase feedwater flow during any operational mode such that it 
would result in a new or different kind of accident from accidents 
previously evaluated. Conclusions of existing analyses are not 
changed by this activity as existing feedwater isolation and control 
valves functions are not changed.
    The proposed changes to add thermal relief valves to the main 
and startup feedwater lines do not adversely affect any safety-
related equipment, and do not add any new interfaces to safety-
related SSCs that adversely affect safety functions. No system or 
design function or equipment qualification is adversely affected by 
these changes as the changes do not modify any SSCs that prevent 
safety functions from being performed by the existing main feedwater 
and startup feedwater valves. The changes do not introduce a new 
failure mode, malfunction or sequence of events that could adversely 
affect safety or safety-related equipment as feedwater isolation 
capabilities are not changed. Performance of overpressure relief 
supports the safety-related functions of the isolation and control 
valves in the main and startup feedwater lines when isolation is 
required.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to COL Appendix C (and plant-specific Tier 
1) Table 2.2.4-1 and Figure 2.2.4-1, and associated UFSAR design 
information add thermal relief valves to the main feedwater and 
startup feedwater lines. These valves are designed to the same codes 
and standards as the existing piping to which they are connected, 
including ASME Code Section III, Class C, seismic Category I. The 
proposed changes do not affect any other safety-related equipment or 
fission product barriers. The requested changes will not affect any 
design code, function, design analysis, safety analysis input or 
result, or design/safety margin. No safety analysis or design basis 
acceptance limit/criterion is challenged or exceeded by the 
requested changes. There are not any changes to operation of the 
main feedwater and startup feedwater isolation and control valves 
when isolation of the lines is required. Operation of the relief 
valves supports isolation capabilities for the main and feedwater 
isolation and control valves.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: July 14, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17195B047.
    Description of amendment request: The requested amendment proposes 
to depart from Tier 2 information in the Updated Final Safety Analysis 
Report (UFSAR) (which includes the plant-specific design control 
document (DCD) Tier 2 information) and involves related changes to 
plant-specific Tier 1 (and associated Combined License (COL) Appendix 
C) information, and COL Appendix A Technical Specifications. 
Specifically, the requested amendment proposes changes to add a second 
normal residual heat removal system (RNS) suction relief valve in 
parallel to the current RNS suction relief valve, with the necessary 
piping changes. Additionally, a change is proposed to Tier 1 Figure 
2.2.1-1, for penetration P19, to accurately depict the orientation of 
the class break of containment isolation valve RNS-PL-V061.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff's edits in 
square brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to Combined License (COL) Appendix C (and 
plant-specific Tier 1) Figures 2.2.1-1 and 2.3.6-1, Tables 2.3.6-1, 
2.3.6-2 and 2.3.6-4, COL Appendix A, Technical Specification 3.4.14 
and associated Updated Final Safety Analysis Report (UFSAR) design 
information to identify a new normal residual heat removal system 
(RNS) relief valve, RNS-PL-V020, do not adversely impact accidents 
previously evaluated in the safety analysis. Transients that are 
capable of overpressurizing the reactor coolant system (RCS) are 
categorized as either mass or heat input transients. The relief 
valves must be capable of passing flow greater than that required 
for the limiting low-temperature overpressure protection (LTOP) 
transients while maintaining RCS pressure less than the lowest 
pressure represented by the pressure/temperature limit curve, 110% 
of the design pressure of the RNS, or the acceptable RNS relief 
valve inlet pressure. The restrictions added to COL Appendix A, 
Technical Specification 3.4.14 to close chemical and volume control 
system (CVS) makeup line containment isolation valve, CVS-PL-V091, 
limit flow capacity when the RCS is aligned to the RNS to support 
LTOP functions and provide reliable operation of the RNS relief 
valves during mass and heat input transients. When CVS-PL-V091 is 
open, the RCS is depressurized and an RCS vent of >=4.15 square 
inches is established. Transient conditions including mass input and 
heat input are not changed and probability of events is not 
increased as the added RNS relief valve, RNS-PL-V020, supports LTOP 
functions as required by Technical Specification 3.4.14. The current 
3-inch RNS relief valve is sufficient to terminate identified 
transients; however, the added 1-inch RNS relief valve reduces 
chatter in the current valve during low flow scenarios.
    Responses to mass and heat input transients are not changed as 
LTOP functions to prevent overpressurization of the RCS are not 
changed by this activity. The added RNS relief valve, RNS-PL-V020, 
is designed in accordance with the same requirements as the current 
RNS relief valve, RNS-PL-V021, but with a lower flow capacity and 
functions at a lower setpoint pressure. Overpressure protection 
provided by the RNS is not

[[Page 42853]]

changed. The change does not adversely impact the capability of the 
RNS to protect the RCS from exceeding pressure and temperature 
limits in accordance with 10 CFR 50, Appendix G or 110% of the 
design pressure of the RNS. Changes in piping to accommodate the 
addition of the valve and reduce inlet piping losses do not impact 
the consequences or probabilities of previously evaluated accidents. 
The class break correction for valve RNS-PL-V061, in COL Appendix C 
(and plant-specific Tier 1) Figure 2.2.1-1 does not impact accidents 
previously evaluated.
    No safety-related structure, system, component (SSC) or function 
is adversely affected by this change. The change does not involve an 
interface with any structure, system, or component (SSC) accident 
initiator or initiating sequence of events, and thus, the 
probabilities of the accidents evaluated in the plant-specific UFSAR 
are not affected. The proposed changes do not involve a change to 
the predicted radiological releases due to postulated accident 
conditions, thus, the consequences of the accidents evaluated in the 
UFSAR are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Conclusions of existing analyses are not changed by the proposed 
change as LTOP functions provided by both the current and added RNS 
relief valves continue to provide the assumed protection for LTOP 
events. RCS pressure is maintained within limits by the use of both 
RNS relief valves. The closure of CVS-PL-V091 limits flow and 
reduces the impact of mass and heat input transients when RNS relief 
valves are relied upon for overpressure protection.
    The proposed change to add the smaller RNS relief valve, RNS-PL-
V020, does not adversely affect safety-related equipment, and does 
not add any new interfaces to safety-related SSCs that adversely 
affect safety functions. The added RNS relief valve, functions in 
the same manner as the current RNS relief valve, but has a lower 
capacity and lifts at a lower pressure. The added RNS relief valve 
also discharges to the liquid radwaste system (WLS) containment 
sump. No system or design function or equipment qualification is 
adversely affected by these changes as the change does not modify 
any SSCs that prevent safety functions from being performed by the 
RNS and the current relief valve. The changes do not introduce a new 
failure mode, malfunction or sequence of events that could adversely 
affect safety or safety-related equipment. Piping changes to 
accommodate the installation of the new valve do not create the 
potential for a new or different kind of accident as the piping 
requirements are consistent with those of the current relief valve, 
and subject to the same pipe rupture evaluation requirements. LTOP 
functions are not changed. The class break correction for valve RNS-
PL-V061 does not impact accident analysis or create a new or 
different kind of accident as the function of the affected equipment 
and piping is not changed.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not affect safety-related equipment or 
fission product barriers. LTOP functions are not adversely impacted 
as both the current and added RNS relief valves continue to provide 
protection from overpressurization. The added RNS relief valve is 
designed in accordance with [American Society of Mechanical 
Engineers (ASME)] Code Section III, Class 2, requirements consistent 
with the current RNS relief valve. Modified piping is constructed 
consistent with current design requirements for RNS piping. The 
addition of the valve adds safety margin in regards to transients as 
the new valve lifts at a lower set pressure than the current valve, 
causing flow rates to be lower through the RNS piping. Therefore, 
margin of safety is not reduced. The requested changes will not 
affect any design code, function, design analysis, safety analysis 
input or result, or design/safety margin. No safety analysis or 
design basis acceptance limit/criterion is challenged or exceeded by 
the requested changes. Transient conditions, including mass input 
and heat input, are not changed and margin of safety is not reduced 
as the added RNS relief valve supports LTOP functions in the same 
manner as the current RNS relief valve.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: May 31, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17151A296.
    Description of amendment request: The requested amendment proposes 
to depart from approved AP1000 Design Control Document (DCD) Tier 2 
information (text, tables, and figures) as incorporated into the 
Updated Final Safety Analysis Report (UFSAR) as plant-specific DCD 
information, and from involved plant-specific Technical Specifications 
as incorporated in Appendix A of the combined license. Specifically, 
the proposed changes support the addition of chemicals necessary to 
achieve proper reactor coolant system (RCS) water quality by allowing 
an unborated water source through the chemical mixing tank to be 
unisolated for <=1 hour for chemical addition to the pressurizer to be 
performed with reactor coolant pumps (RCPs) not in operation. In order 
to perform chemical addition to the pressurizer without the mixing 
provided by forced reactor coolant system (RCS) flow, administrative 
controls are established such that coolant introduced into the RCS is 
at a boron concentration greater than or equal to that required to meet 
the shutdown margin (SDM) boron concentration.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Updated Final Safety Analysis Report (UFSAR) 15.4.6, Chemical 
and Volume Control System Malfunction that Results in a Decrease in 
the Boron Concentration in the Reactor Coolant, addresses 
inadvertent boron dilution events. The principal means of positive 
reactivity insertion to the core is the addition of unborated, 
primary-grade water from the demineralized water transfer and 
storage system (DWS) into the reactor coolant system (RCS) through 
the reactor makeup portion of the chemical and volume control system 
(CVS).
    These events are primarily evaluated with one or more reactor 
coolant pumps (RCPs) in operation providing adequate mixing. The 
changes proposed by this amendment request do not involve operations 
where the RCPs are in operation. Therefore, there is no increase in 
the probability or consequences of inadvertent boron dilution events 
with RCPs operating.
    UFSAR Subsection 15.4.6 also describes that when a reactor 
coolant pump is not operating, the demineralized water isolation 
valves are closed and an uncontrolled boron dilution transient 
cannot occur. The proposed amendment adds provisions to allow a 
specific CVS unborated water source flow path to be opened through 
the chemical mixing tank to the RCS pressurizer when RCPs are not in 
operation for the purpose of chemical addition to the pressurizer. 
The administrative control provisions proposed provide adequate 
assurance that any injection to the RCS pressurizer would only occur 
such that injected water is limited to

[[Page 42854]]

boron concentrations greater than the required concentrations to 
meet the SDM. With no reduction in SDM, there would be no means of 
positive reactivity insertion to the core leading to an adverse 
reactivity event. As such, there is no significant increase in the 
probability of a previously evaluated boron dilution event as a 
result of this change.
    Since the proposed change does not lead to any positive 
reactivity insertion, there are no increased consequences of an 
accident previously evaluated.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The administrative control provisions proposed provide adequate 
assurance that any injection to the pressurizer would only occur 
such that injected water is limited to boron concentrations greater 
than the required concentrations to meet the SDM. With no reduction 
in SDM, there would be no means of positive reactivity insertion to 
the core leading to an adverse reactivity event. Failure modes 
involving procedural controls and operator actions are considered in 
evaluating inadvertent boron dilution events. The possibility of a 
new or different kind of failure, malfunction, or sequence of events 
has been evaluated with these proposed changes; events are precluded 
with the proposed administrative controls and defense in depth 
features inherent in the AP1000 design.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety is established by maintaining the required 
SDM during shutdown activities. The proposed changes to the UFSAR 
and Technical Specifications do not adversely affect the safety-
related functions of the RCS or CVS in maintaining adequate SDM. 
Provisions are proposed for a specific CVS unborated water source 
flow path to be opened through the chemical mixing tank to the RCS 
pressurizer when RCPs are not in operation; however, this activity 
is performed under administrative controls that preclude the 
potential for a reduction in SDM.
    The changes do not affect containment penetrations or any other 
safety-related equipment or fission product barriers. The requested 
changes will not affect any design code, function, design analysis, 
safety analysis input or result, or design/safety margin. No safety 
analysis or design basis acceptance limit/criterion is challenged or 
exceeded by the requested changes. The existing design and operation 
of the associated systems are adequate to preclude an inadvertent 
boron dilution from occurring when RCPs are not in operation.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazard consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: July 28, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17209A185.
    Description of amendment request: The requested amendment proposes 
to depart from approved AP1000 Design Control Document (DCD) Tier 2 
information as incorporated into the Updated Final Safety Analysis 
Report (UFSAR) as plant-specific DCD information, and also proposes to 
depart from involved plant-specific Tier 1 information and the 
associated combined license (COL) Appendix C information. Specifically, 
the amendment, if approved, would revise the COL documents mentioned 
previously to reflect the proposed changes to update Reactor Coolant 
System (RCS) requirements for the reactor vessel head vent (RVHV) mass 
flow rate. Pursuant to the provisions of 10 CFR 52.63(b)(1), an 
exemption from elements of the design as certified in the 10 CFR part 
52, Appendix D, design certification rule is also requested for the 
plant-specific DCD Tier 1 material departures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    UFSAR Subsections 15.2.7, 15.5.1, and 15.5.2 describe analyses 
performed for an increase in reactor coolant inventory due to a loss 
of normal feedwater flow, and for malfunctions of the chemical and 
volume control system and the core makeup tanks. In each of these 
evaluated accidents, it is assumed that the operators are alerted to 
the event due to a high pressurizer water level and take subsequent 
action to open the reactor vessel head vent valves. When the head 
vent is opened, the pressurizer water level increase slows and 
eventually decreases.
    Changing the required mass flow rate from 8.2 lbm/sec at a 
Reactor Coolant System (RCS) pressure of 1250 psia [pounds per 
square inch absolute] to 9.0 lbm/sec [pounds mass per second] at an 
RCS pressure of 2500 psia for the reactor vessel head vent (RVHV) 
flow path does not change the probability of these events occurring. 
The valves are used to mitigate the events. They are not an 
initiator of these accidents, or any other accident previously 
evaluated. Changing the required mass flow rate does not change the 
consequences of these accidents. The proposed flow rate change is 
made to be consistent with the latest AP1000 safety analysis. This 
change does not lead to an increase in the probability of a loss of 
coolant accident, nor does it cause the RVHV to exceed the 
capability of the normal makeup system. The changes described above 
continue to ensure the design is capable of providing adequate flow 
rate for emergency letdown and the prevention of long term 
pressurizer overfill.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes impact the acceptance criteria for RVHV 
mass flow rate. The required mass flow rate is changed from 8.2 lbm/
sec at an RCS pressure of 1250 psia to 9.0 lbm/sec at an RCS 
pressure of 2500 psia to align with the events evaluated in the 
current safety analysis. The proposed changes do not result in a new 
accident initiator and do not impact a current accident initiator.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes impact the acceptance criteria for RVHV 
mass flow rate. The required mass flow rate is changed from 8.2 lbm/
sec at an RCS pressure of 1250 psia to 9.0 lbm/sec at an RCS 
pressure of 2500 psia. The proposed changes are made to reflect the 
updated AP1000 plant safety analysis; the changes are conservative 
and bound the expected performance of the as-built equipment.
    COL Appendix C (plant-specific Tier 1) is proposed to be updated 
to reflect the new mass flow rate through the RVHV line and the 
associated system pressure. COL Appendix C (plant-specific Tier 1) 
is updated to reflect the latest safety analysis, which credits an 
emergency letdown mass flow rate of 9.0 lbm/sec at an RCS pressure 
of 2500

[[Page 42855]]

psia. At these conditions, long term pressurizer overfill is 
prevented. RCS calculations show that the expected mass flow rate 
through the emergency letdown path is 12.34 lbm/sec. Therefore, the 
safety analysis calculation, and the corresponding mass flow rate 
and RCS pressure values used in the proposed changes, is 
conservative and bounded by the expected mass flow rate.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue, North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas
    Date of amendment request: July 31, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17212A842.
    Description of amendment request: The amendment would revise the 
staffing and staff augmentation times described in the South Texas 
Project Emergency Plan. The proposed amendment would increase the 
Emergency Response Organization (ERO) response times and would modify 
minimum staffing functions and requirements of the ERO and Operations 
Support Center staff. The changes also include formatting, 
clarification, and editorial modifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment has no effect on normal plant operation 
or on any accident initiator or precursors and does not impact the 
function of plant structures, systems, or components. The proposed 
changes do not alter or prevent the ability of the Emergency 
Response Organization to perform their intended functions to 
mitigate the consequences of an accident or event.
    Therefore, the proposed STPEGS [South Texas Project Electric 
Generating Station] Emergency Plan change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not impact any accident analysis. 
The change does not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed), a 
change in the method of plant operation, or new operator actions. 
The proposed change does not introduce failure modes that could 
result in a new accident, and the change does not alter assumptions 
made in the safety analysis. The proposed change revises the on-
shift staffing and staff augmentation response times in the STPEGS 
Emergency Plan. The proposed changes do not alter or prevent the 
ability of the Emergency Response Organization to perform their 
intended functions to mitigate the consequences of an accident or 
event.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed change is 
associated with the STPEGS Emergency Plan staff and staff 
augmentation and does not impact operation of the plant or its 
response to transients or accidents. The change does not affect the 
Technical Specifications. The proposed change does not involve a 
change in the method of plant operation and no accident analyses 
will be affected by the proposed change. Safety analysis acceptance 
criteria are not affected by the proposed change. The revised STPEGS 
Emergency Plan will continue to provide the necessary response staff 
with the proposed change. Therefore, the proposed change is 
determined to not adversely affect the ability to meet the 
requirements of 10 CFR 50.54(q)(2), 10 CFR 50 Appendix E, or the 
emergency planning standards described in 10 CFR 50.47(b).
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kym Harshaw, General Counsel, STP Nuclear 
Operating Company, P.O. Box 289, Wadsworth, TX 77483.
    NRC Branch Chief: Robert J. Pascarelli.

III. Notice of Issuance of Amendments to Facility Operating Licenses 
and Combined Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.

[[Page 42856]]

Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick 
Steam Electric Plant, Units 1 and 2 (Brunswick), Brunswick County, 
North Carolina

Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear 
Power Plant, Unit 1 (Harris), Wake County, North Carolina

Duke Energy Progress, LLC, Docket No. 50-261, H.B. Robinson Steam 
Electric Plant Unit No. 2 (Robinson), Darlington County, South Carolina

Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3 (Oconee), Oconee County, 
South Carolina

    Date of amendment request: April 29, 2016, as supplemented by 
letters dated October 3, 2016, and January 16, 2017.
    Brief description of amendments: The amendments (1) consolidated 
the Emergency Operations Facilities (EOFs) for Brunswick, Harris, and 
Robinson with the Duke Energy Progress, LLC (Duke Energy) corporate EOF 
in Charlotte, North Carolina; (2) decreased the frequency for a 
multisite drill at Oconee from once per 6 years to once per 8 years; 
(3) allowed the multisite drill performance with sites other than the 
Catawba Nuclear Station, McGuire Nuclear Station, or Oconee; (4) 
changed the Brunswick, Harris, and Robinson augmentation times to be 
consistent with those of the sites currently supported by the Duke 
Energy corporate EOF; and (5) decreased the frequency of the 
unannounced augmentation drill at Brunswick from twice per year to once 
per year.
    Date of issuance: August 21, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment Nos.: 279 and 307 for Brunswick, Units 1 and 2; 160 for 
Harris, Unit 1; 254 for Robinson Unit No. 2; and 405, 407, and 406 for 
Oconee, Units 1, 2, and 3. A publicly-available version is in ADAMS 
under Accession No. ML17188A387; documents related to these amendments 
are listed in the Safety Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-71 and DPR-62 for 
Brunswick, Units 1 and 2; NPF-63 for Harris, Unit 1; DPR-23 for 
Robinson Unit No. 2; and DPR-38, DPR-47, and DPR-55 for Oconee, Units 
1, 2, and 3: The amendments revised the emergency plans.
    Date of initial notice in Federal Register: July 5, 2016 (81 FR 
43650). The supplemental letters dated October 3, 2016, and January 16, 
2017, provided additional information that expanded the scope of the 
application as originally noticed and changed the NRC staff's original 
proposed no significant hazards consideration determination as 
published in the Federal Register. Accordingly, the NRC published a 
second proposed no significant hazards consideration determination in 
the Federal Register on February 14, 2017 (82 FR 10594). This notice 
superseded the original notice in its entirety.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 21, 2017.
    No significant hazards consideration comments received: No.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: July 28, 2016, as supplemented 
by letters dated February 23, 2017, and June 21, 2017.
    Brief description of amendment: The amendment revised the current 
emergency action level scheme to one based on Nuclear Energy Institute 
(NEI) guidance in NEI 99-01, Revision 6, ``Development of Emergency 
Action Levels for Non-Passive Reactors'' (ADAMS Accession No. 
ML12326A805). Revision 6 of NEI 99-01 was endorsed by the NRC in a 
letter dated March 28, 2013.
    Date of issuance: August 28, 2017.
    Effective date: As of its date of issuance and shall be implemented 
within 180 days from the date of issuance.
    Amendment No.: 244. A publicly-available version is in ADAMS under 
Accession No. ML17188A230; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-21: The amendment 
revised the Operating License.
    Date of initial notice in Federal Register: September 27, 2016 (81 
FR 66305). The supplemental letters dated February 23, 2017, and June 
21, 2017, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 28, 2017.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of application for amendments: October 7, 2016, as 
supplemented by letter dated March 20, 2017.
    Brief description of amendments: The amendments revised the Updated 
Final Safety Analysis Report (UFSAR) to identify the TORMIS Computer 
Code as the methodology used for assessing tornado-generated missile 
protection of unprotected plant structures, systems and components 
(SSCs) and to describe the results of the Byron Station site-specific 
tornado hazard analysis.
    Date of issuance: August 10, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of the date of issuance. The UFSAR changes shall be 
filed with the NRC in the next periodic update to the UFSAR scheduled 
for December 15, 2018.
    Amendment Nos.: 199 for NPF-37 and 199 for NPF-66. A publicly-
available version is in ADAMS under Accession No. ML17188A155; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Facility Operating License Nos. NPF-37, and NPF-66: The amendments 
revised the current licensing basis as described in the UFSAR.
    Date of initial notice in Federal Register: December 6, 2016 (81 FR 
87969). The March 20, 2017, supplement contained clarifying information 
and did not change the scope of the proposed action or affect the NRC 
staff's initial proposed finding of no significant hazards 
consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 10, 2017.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station (Beaver Valley), Unit Nos. 1 and 2, 
Beaver County, Pennsylvania

    Date of amendment request: June 30, 2017.
    Brief description of amendments: The amendments modified 
requirements on control and shutdown rods, and rod and bank position 
indication for Beaver Valley, Unit No. 2. The changes are consistent 
with Technical Specifications Task Force (TSTF) Traveler TSTF-547, 
Revision 1, ``Clarification of Rod

[[Page 42857]]

Position Requirements.'' Additional supporting changes to Beaver 
Valley, Unit Nos. 1 and 2, Technical Specifications were also made.
    Date of Issuance: August 16, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 299 (Unit No. 1) and 188 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML17221A280; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-66 and NPF-73: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: July 11, 2017 (82 FR 
32017).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 16, 2017.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama

    Date of amendment request: September 21, 2015, as supplemented by 
letters dated November 13, December 15 (two letters), and December 18, 
2015; February 16, March 8, March 9, March 24, March 28, April 4, April 
5, April 14, April 22 (two letters), April 27, May 11, May 20 (two 
letters), May 27, June 9, June 17, June 20, June 24, July 13 (two 
letters), July 27, July 29 (two letters), August 3 (three letters), 
September 12, September 21, September 23, October 13, October 28, and 
October 31, 2016; and January 20, February 3, March 3, and June 12, 
2017.
    Brief description of amendments: The amendments revised Renewed 
Facility Operating Licenses (RFOLs) and Technical Specifications (TSs) 
to authorize an increase of maximum reactor core thermal power level 
for Browns Ferry Nuclear Plant, Units 1, 2, and 3 to 3,952 megawatt 
thermal (MWt). These license amendments represent an increase of 
approximately 14.3 percent above the current licensed thermal power 
level of 3,458 MWt, which is an increase of approximately 20 percent 
above the original licensed thermal power level of 3,293 MWt. The NRC 
considers the requested increase in power level to be an extended power 
uprate.
    Date of issuance: August 14, 2017.
    Effective date: As of the date of issuance and shall be implemented 
prior to startup from the refueling outages of fall 2018 (Unit 1), 
spring 2019 (Unit 2), and spring 2018 (Unit 3).
    Amendment Nos.: 299 (Unit 1), 323 (Unit 2), and 283 (Unit 3). A 
publicly-available version is in ADAMS under Accession No. ML17032A120; 
documents related to these amendments are listed in the Safety 
Evaluation (SE) enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the RFOLs and TSs.
    Date of initial notice in Federal Register: July 5, 2016 (81 FR 
43666). The supplemental letters dated April 22 (two letters), April 
27, May 11, May 20 (two letters), May 27, June 9, June 17, June 20, 
June 24, July 13, (two letters); July 27, July 29 (two letters), August 
3 (three letters), September 12, September 21, September 23, October 
13, October 28, and October 31, 2016; and January 20, February 3, March 
3, and June 12, 2017, provided additional information that clarified 
the application, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in the SE dated August 14, 2017.
    No significant hazards consideration comments received: Yes, refer 
to Section 6.0, ``Public Comments,'' of the SE.

Wolf Creek Nuclear Operating Corporation (WCNOC), Docket No. 50-482, 
Wolf Creek Generating Station (WCGS), Coffey County, Kansas

    Date of amendment request: September 30, 2016, as supplemented by 
letters dated March 16 and April 26, 2017.
    Brief description of amendment: The amendment revised the emergency 
action level (EAL) scheme used at WCGS. The currently approved EAL 
scheme is based on Nuclear Management and Resources Council/National 
Environmental Studies Project (NUMARC/NESP)-007, Revision 2, 
``Methodology for Development of Emergency Action Levels,'' January 
1992. The amendment allows WCNOC to adopt an EAL scheme, which is based 
on the guidance established in Nuclear Energy Institute (NEI) 99-01, 
Revision 6, ``Development of Emergency Action Levels for Non-Passive 
Reactors,'' November 2012. Revision 6 of NEI 99-01 has been endorsed by 
the NRC by letter dated March 28, 2013.
    Date of issuance: August 28, 2017.
    Effective date: As of its date of issuance and shall be implemented 
by September 30, 2018.
    Amendment No.: 218. A publicly-available version is in ADAMS under 
Accession No. ML17166A409; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Operating License.
    Date of initial notice in Federal Register: December 6, 2016 (81 FR 
87974). The supplemental letters dated March 16 and April 26, 2017, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 28, 2017.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 30th day of August 2017.

    For the Nuclear Regulatory Commission.
Eric J. Benner,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2017-19214 Filed 9-11-17; 8:45 am]
 BILLING CODE 7590-01-P


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