[Federal Register Volume 83, Number 10 (Tuesday, January 16, 2018)]
[Notices]
[Pages 2224-2234]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2018-00386]


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NUCLEAR REGULATORY COMMISSION

[NRC-2018-0005]


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 December 19, 2017 to December 29, 2017. The 
last biweekly notice was published on January 2, 2018.

DATES: Comments must be filed by February 15, 2018. A request for a 
hearing must be filed by March 19, 2018.

ADDRESSES: You may submit comments by any of the following methods.
    Federal Rulemaking website: Go to http://www.regulations.gov and 
search for Docket ID NRC-2018-0005. Address questions about NRC dockets 
to Carol Gallagher; telephone: 301-415-3463; email: 
<a href="/cdn-cgi/l/email-protection#682b091a0704462f090404090f000d1a28061a0b460f071e"><span class="__cf_email__" data-cfemail="115270637e7d3f56707d7d7076797463517f63723f767e67">[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: May Ma, Office of Administration, Mail 
Stop: OWFN-2-A13, 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: Kay Goldstein, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-1506, email: <a href="/cdn-cgi/l/email-protection#2a614b53046d45464e595e4f43446a445849044d455c"><span class="__cf_email__" data-cfemail="672c061e4920080b031413020e092709150449000811">[email protected]</span></a>.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2018-0005, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
    <bullet> Federal Rulemaking website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0005.
    <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#2a5a4e5804584f59455f58494f6a445849044d455c"><span class="__cf_email__" data-cfemail="28584c5a065a4d5b475d5a4b4d68465a4b064f475e">[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-2018-0005, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering 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

[[Page 2225]]

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

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if 
appropriate, a notice of a hearing will be issued.
    As required by 10 CFR 2.309(d) the petition should specifically 
explain the reasons why intervention should be permitted with 
particular reference to the following general requirements for 
standing: (1) The name, address, and telephone number of the 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
petitioner's property, financial, or other interest in the proceeding; 
and (4) the possible effect of any decision or order which may be 
entered in the proceeding on the petitioner's interest.
    In accordance with 10 CFR 2.309(f), the petition must also set 
forth the specific contentions which the petitioner seeks to have 
litigated in the proceeding. Each contention must consist of a specific 
statement of the issue of law or fact to be raised or controverted. In 
addition, the petitioner must provide a brief explanation of the bases 
for the contention and a concise statement of the alleged facts or 
expert opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to the specific sources and 
documents on which the petitioner intends to rely to support its 
position on the issue. The petition must include sufficient information 
to show that a genuine dispute exists with the applicant or licensee on 
a material issue of law or fact. Contentions must be limited to matters 
within the scope of the proceeding. The contention must be one which, 
if proven, would entitle the petitioner to relief. A petitioner who 
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene. 
Parties have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that party's admitted 
contentions, including the opportunity to present evidence, consistent 
with the NRC's regulations, policies, and procedures.
    Petitions must be filed no later than 60 days from the date of 
publication of this notice. Petitions and motions for leave to file new 
or amended contentions that are filed after the deadline will not be 
entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in 
accordance with the filing instructions in the ``Electronic Submissions 
(E-Filing)'' section of this document.
    If a hearing is requested, and the Commission has not made a final 
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

[[Page 2226]]

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 website at 
http://www.nrc.gov/site-help/e-submittals.html. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at <a href="/cdn-cgi/l/email-protection#89e1ece8fbe0e7eea7ede6eae2ecfdc9e7fbeaa7eee6ff"><span class="__cf_email__" data-cfemail="38505d594a51565f165c575b535d4c78564a5b165f574e">[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 website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a 
digital ID certificate and a docket has been created, the participant 
can then submit adjudicatory documents. Submissions must be in Portable 
Document Format (PDF). Additional guidance on PDF submissions is 
available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the 
time the document is submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
document on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before adjudicatory documents are 
filed so that they can obtain access to the documents via the E-Filing 
system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public website at http://www.nrc.gov/site-help/e-submittals.html, by 
email to <a href="/cdn-cgi/l/email-protection#420f110a066c1027312d37302127022c30216c252d34"><span class="__cf_email__" data-cfemail="78352b303c562a1d0b170d0a1b1d38160a1b561f170e">[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.

[[Page 2227]]

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2 (CNS), York County, South Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2 (MNS), Mecklenburg County, North 
Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3 (ONC), Oconee County, South 
Carolina
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear 
Power Plant, Unit 1 (HNP), Wake County, North Carolina
Duke Energy Progress, LLC, Docket No. 50-261, H.B. Robinson Steam 
Electric Plant, Unit No. 2 (RNP), Darlington County, South Carolina
Date of amendment request: November 7, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17312A362.
Description of amendment request: The amendments would revise the 
technical specifications (TSs) based on Technical Specification Task 
Force (TSTF) Traveler TSTF-545, Revision 3, ``TS Inservice Testing 
[IST] Program Removal & Clarify SR [Surveillance Requirement] Usage 
Rule Application to Section 5.5 Testing'' (ADAMS Accession No. 
ML15294A555), with some variations. For each plant, the changes include 
deleting the current TS for the IST Program, adding a new defined term, 
``INSERVICE TESTING PROGAM,'' to the TSs, and revising other TSs to 
reference this new defined term instead of the deleted TS.
    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 revises TS Chapter 5 (TS Chapter 6 for HNP), 
``Administrative Controls,'' Section 5.5 (Section 6.8.4 for HNP), 
``Programs and Manuals,'' by replacing the current contents of the 
``Inservice Testing Program'' specification with a note referring to 
the TS Definition of ``INSERVICE TESTING PROGRAM.'' Most 
requirements in the Inservice Testing Program are removed, as they 
are duplicative of requirements in the ASME OM Code [American 
Society of Mechanical Engineers Code for Operations and Maintenance 
of Nuclear Power Plants], as clarified by Code Case OMN-20, 
``Inservice Test Frequency.'' The remaining requirements in the 
Section 5.5 (Section 6.8.4 for HNP) IST Program are eliminated 
because the NRC has determined their inclusion in the TS is contrary 
to regulations. A new defined term, ``INSERVICE TESTING PROGRAM,'' 
is added to the TS, which references the requirements of 10 CFR 
50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. Inservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than or equal to 2 
years may be extended by up to 6 months to facilitate test 
scheduling and consideration of plant operating conditions that may 
not be suitable for performance of the required testing. The testing 
frequency extension will not affect the ability of the components to 
mitigate any accident previously evaluated as the components are 
required to be operable during the testing period extension. 
Performance of inservice tests utilizing the allowances in OMN-20 
will not significantly affect the reliability of the tested 
components. As a result, the availability of the affected 
components, as well as their ability to mitigate the consequences of 
accidents previously evaluated, is not affected.
    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 does not alter the design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The proposed change does not alter the types of 
inservice testing performed. In most cases, the frequency of 
inservice testing is unchanged. However, the frequency of testing 
would not result in a new or different kind of accident from any 
previously evaluated since the testing methods are not altered.
    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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than or equal to 2 years to be extended by 6 
months to facilitate test scheduling and consideration of plant 
operating conditions that may not be suitable for performance of the 
required testing. The testing frequency extension will not affect 
the ability of the components to respond to an accident as the 
components are required to be operable during the testing period 
extension. The proposed change will eliminate the existing TS SR 
3.0.3 allowance to defer performance of missed inservice tests up to 
the duration of the specified testing frequency, and instead will 
require an assessment of the missed test on equipment operability. 
This assessment will consider the effect on a margin of safety 
(equipment operability). Should the component be inoperable, the 
Technical Specifications provide actions to ensure that the margin 
of safety is protected. The proposed change also eliminates a 
statement that nothing in the ASME Code should be construed to 
supersede the requirements of any TS. The NRC has determined that 
statement to be incorrect. However, elimination of the statement 
will have no effect on plant operation or safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tryon Street, Mail Code DEC45A, 
Charlotte NC 28202.
    NRC Branch Chief: Undine Shoop.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of amendment request: October 23, 2017, as supplemented by 
letter dated November 15, 2017. Publicly-available versions are in 
ADAMS under Accession Nos. ML17296B380, and ML17320A314, respectively.
    Description of amendment request: The proposed amendment would 
adopt Technical Specification Task Force (TSTF) traveler TSTF-542, 
Revision 2, ADAMS Accession No. ML16343B008 ``Reactor Pressure Vessel 
Water Inventory Control.'' The proposed amendment 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 2.1.1.3. Safety Limit 2.1.1.3 
requires the 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

[[Page 2228]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements RPV WIC that will protect Safety Limit 
2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold 
shutdown) and Mode 5 (i.e., refueling) is not an accident previously 
evaluated and, therefore, replacing the existing TS controls to 
prevent or mitigate such an event with a new set of controls has no 
effect on any accident previously evaluated. RPV water inventory 
control in Mode 4 or Mode 5 is not an initiator of any accident 
previously evaluated. The existing OPDRV controls or the proposed 
RPV WIC controls are not mitigating actions assumed in any accident 
previously evaluated.
    The proposed change reduces the probability of an unexpected 
draining event (which is not a previously evaluated accident) by 
imposing new requirements on the limiting time in which an 
unexpected draining event could result in the reactor vessel water 
level dropping to the top of the active fuel (TAF). These controls 
require cognizance of the plant configuration and control of 
configurations with unacceptably short drain times. These 
requirements reduce the probability of an unexpected draining event. 
The current TS requirements are only mitigating actions and impose 
no requirements that reduce the probability of an unexpected 
draining event.
    The proposed change reduces the consequences of an unexpected 
draining event (which is not a previously evaluated accident) by 
requiring an Emergency Core Cooling System (ECCS) subsystem to be 
operable at all times in Modes 4 and 5. The current TS requirements 
do not require any water injection systems, ECCS or otherwise, to be 
Operable in certain conditions in Mode 5. The change in requirement 
from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does 
not significantly affect the consequences of an unexpected draining 
event because the proposed Actions ensure equipment is available 
within the limiting drain time that is as capable of mitigating the 
event as the current requirements. The proposed controls provide 
escalating compensatory measures to be established as calculated 
drain times decrease, such as verification of a second method of 
water injection and additional confirmations that containment and/or 
filtration would be available if needed.
    The proposed change reduces or eliminates some requirements that 
were determined to be unnecessary to manage the consequences of an 
unexpected draining event, such as automatic initiation of an ECCS 
subsystem and control room ventilation. These changes do not affect 
the consequences of any accident previously evaluated since a 
draining event in Modes 4 and 5 is not a previously evaluated 
accident and the requirements are not needed to adequately respond 
to a draining event.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. The proposed change will not alter the design 
function of the equipment involved. Under the proposed change, some 
systems that are currently required to be operable during OPDRVs 
would be required to be available within the limiting drain time or 
to be in service depending on the limiting drain time. Should those 
systems be unable to be placed into service, the consequences are no 
different than if those systems were unable to perform their 
function under the current TS requirements.
    The event of concern under the current requirements and the 
proposed change is an unexpected draining event. The proposed change 
does not create new failure mechanisms, malfunctions, or accident 
initiators that would cause a draining event or a new or different 
kind of accident not previously evaluated or included in the design 
and licensing bases.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC. The current requirements do 
not have a stated safety basis 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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW, Washington, DC 20006-3817.
    NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana
    Date of amendment request: November 28, 2017, as supplemented by 
letter dated December 7, 2017. Publicly-available versions are in ADAMS 
under Accession Nos. ML17332A898, and ML17341B295, respectively.
    Description of amendment request: The proposed amendment would 
revise Section 4.3.3 of the Waterford 3 Updated Final Safety Analysis 
Report to indicate that the RAPTOR-M3G code is used for reactor vessel 
fluence calculations. The use of the RAPTOR-M3G code would meet the 
criteria present in Regulatory Guide (RG) 1.190, ``Calculational and 
Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,'' 
dated March 2001.
    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 probability of occurrence of an accident previously 
evaluated for Waterford 3 is not altered by the proposed license 
amendment. The accidents currently analyzed in the Waterford 3 Final 
Safety Analysis Report (FSAR) remain the same. The proposed change 
does not impact the integrity of the reactor coolant pressure 
boundary (RCPB) (i.e., there is no change to the operating pressure, 
materials, loadings, etc.). The proposed change does not affect the 
probability nor consequences of any design basis accident (DBA). The 
proposed neutron fluence calculational methodology meets the 
criteria in RG 1.190 and will be used to ensure that the P/T 
[pressure-temperature] limit curves, maximum heatup and cooldown 
rates, and LTOP [low-temperature overpressure protection] enable 
temperature remain acceptable to maintain reactor pressure vessel 
integrity.
    Fracture toughness test data are obtained from material 
specimens contained in capsules that are periodically withdrawn from 
the reactor vessel. These data, combined

[[Page 2229]]

with the neutron fluence calculations, permit determination of the 
conditions under which the vessel can be operated with adequate 
safety margins against brittle fracture throughout its service life. 
For each analyzed transient and steady state condition, the 
allowable pressure is determined as a function of reactor coolant 
temperature considering postulated flaws in the reactor vessel 
beltline, inlet nozzle, outlet nozzle, and closure head.
    The predicted radiation induced [Delta]RT<INF>NDT</INF> [delta 
reference temperature nil ductility transition] is calculated using 
the respective reactor vessel beltline materials' copper and nickel 
contents and the neutron fluence determination. The RT<INF>NDT</INF> 
and, in turn, the operating limits for Waterford 3 are adjusted, if 
necessary, to account for the effects of irradiation on the fracture 
toughness of the reactor vessel materials and maintain reactor 
vessel integrity within design assumptions.
    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 to the neutron fluence calculational method 
will not create a new accident scenario. The requirements to have P/
T limits and LTOP protection are part of the licensing basis for 
Waterford 3. The neutron fluence calculation method will validate, 
and when necessary, provide input to the development of new 
operating limits. The data analysis for the vessel surveillance 
specimens are used to confirm that the vessel materials are 
responding as predicted based on previous neutron fluence 
projections.
    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 change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change to the neutron fluence calculational method 
conforms to the criteria presented in RG 1.190 and will ensure that 
Waterford 3 continues to operate within the operating margins 
allowed by 10 CFR 50.60 and the ASME [American Society of Mechanical 
Engineers] Code.
    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: Ms. Anna Vinson Jones, Senior Counsel, 
Entergy Services, Inc., 101 Constitution Avenue NW, Suite 200 East, 
Washington, DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC (Exelon), Docket No. 50-219, Oyster 
Creek Nuclear Generating Station (OCNGS), Ocean County, New Jersey
    Date of amendment request: November 16, 2017. A publicly-available 
version is available in ADAMS under Accession No. ML17320A411.
    Description of amendment request: The amendment would revise the 
OCNGS renewed facility operating license (RFOL) and the associated 
Technical Specifications (TSs) to Permanently Defueled Technical 
Specifications (PDTS) consistent with the permanent cessation of 
reactor operation and permanent defueling of the reactor. By letter 
dated January 7, 2011 (ADAMS Accession No. ML110070507), Exelon 
provided formal notification to the NRC of Exelon's contingent 
determination to permanently cease operations at OCNGS no later than 
December 31, 2019. The amendment would eliminate those TSs applicable 
in operating modes or modes where fuel is placed in the reactor vessel. 
The amendment would change other TS limiting conditions for operation 
(LCOs), definitions, surveillance requirements (SRs), administrative 
controls, as well as several license conditions.
    Basis for proposed no significant hazards consideration (NSHC) 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of NSHC, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes would not take effect until OCNGS has 
permanently ceased operation, entered a permanently defueled 
condition, and at least 60 days of irradiated fuel decay time after 
reactor shutdown. The proposed changes would revise the OCNGS RFOL 
and TS by deleting or modifying certain portions of the TS that are 
no longer applicable to a permanently shutdown and defueled 
facility. This change is consistent with the criteria set forth in 
10 CFR 50.36 for the contents of TS.
    Chapter 15 of the OCNGS Updated Final Safety Analysis Report 
(UFSAR) described the design basis accident (DBA) and transient 
scenarios applicable to OCNGS during power operations. The analyzed 
accidents that remains applicable to OCNGS in the permanently shut 
down and defueled condition is a Fuel Handling Accident (FHA) in the 
[spent fuel pool (SFP)] (a dropped fuel assembly onto the top of the 
core will no longer be applicable) and the Postulated Radioactive 
Tank Failure and Release of Radioactive Liquid Waste while 
radioactive liquids are still present. The FHA is the remaining 
accident with radiological consequences and has been revised for the 
permanently shutdown and defueled condition. The liquid tank 
accidents analysis remains bounding and unchanged; therefore, is not 
discussed further in this NSHC evaluation.
    Once the reactor is in a permanently defueled condition, the 
spent fuel pool (SFP) and its cooling systems will be dedicated only 
to spent fuel storage. In this condition, the spectrum of credible 
accidents will be much smaller than for an operational plant. Once 
the certifications are docketed by OCNGS pursuant to 10 CFR 
50.82(a)(1), and the consequent removal of authorization to operate 
the reactor or to place or retain fuel in the reactor vessel 
pursuant to 10 CFR 50.82(a)(2), the majority of the accident 
scenarios previously postulated in the UFSAR will no longer be 
possible and will be removed from the UFSAR under the provisions of 
10 CFR 50.59.
    The deletion of TS definitions and rules of usage and 
application, that will not be applicable in a defueled condition, 
has no impact on facility structures, systems, and components (SSCs) 
or the methods of operation of such SSCs. The deletion of design 
features and safety limits not applicable to the permanently 
shutdown and defueled status of OCNGS has no impact on the remaining 
applicable DBA. The removal of LCOs or SRs that are related to only 
the operation of the nuclear reactor or to only the prevention, 
diagnosis, or mitigation of reactor-related transients or accidents 
do not affect the applicable DBAs previously evaluated since these 
DBAs are no longer applicable in the defueled mode. The safety 
functions involving core reactivity control, reactor heat removal, 
reactor coolant system inventory control, and containment integrity 
are no longer applicable at OCNGS as a permanently defueled plant. 
The analyzed accidents involving damage to the reactor coolant 
system, main steam lines, reactor core, and the subsequent release 
of radioactive material will no longer be possible at OCNGS.
    After OCNGS permanently ceases operation, the future generation 
of fission products will cease and the remaining source term will 
decay. The radioactive decay of the irradiated fuel following 
shutdown of the reactor will have reduced the consequences of the 
FHA in the SFP below those previously analyzed. The relevant 
parameter (water level) associated with the fuel pool provides an 
initial condition for the FHA analysis and is included in the PDTS.
    The SFP water level and spent fuel storage TSs are retained to 
preserve the current requirements for safe storage of irradiated 
fuel. SFP cooling and makeup related equipment and support equipment 
(e.g., electrical power systems) are not required to be continuously 
available since there will be sufficient time to effect repairs, 
establish alternate sources of makeup flow, or establish alternate 
sources of cooling in the event of a loss of cooling and makeup flow 
to the SFP.
    The deletion and modification of provisions of the 
administrative controls do not directly affect the design of SSCs 
necessary for safe storage of irradiated fuel or the methods used 
for handling and storage of

[[Page 2230]]

such fuel in the fuel pool. The changes to the administrative 
controls are administrative in nature and do not affect any 
accidents applicable to the safe management of irradiated fuel or 
the permanently shutdown and defueled condition of the reactor.
    The probability of occurrence of previously evaluated accidents 
is not increased, since extended operation in a defueled condition 
will be the only operation allowed, and therefore bounded by the 
existing analyses. Additionally, the occurrence of postulated 
accidents associated with reactor operation will no longer be 
credible in a permanently defueled reactor. This significantly 
reduces the scope of applicable accidents.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to delete and/or modify certain TS have no 
impact on facility SSCs affecting the safe storage of spent 
irradiated fuel, or on the methods of operation of such SSCs, or on 
the handling and storage of spent irradiated fuel itself. The 
removal of TS that are related only to the operation of the nuclear 
reactor or only to the prevention, diagnosis, or mitigation of 
reactor related transients or accidents, cannot result in different 
or more adverse failure modes or accidents than previously evaluated 
because the reactor will be permanently shutdown and defueled and 
OCNGS will no longer be authorized to operate the reactor.
    The proposed deletion of requirements of the OCNGS RFOL and TS 
do not affect systems credited in the accident analysis for the FHA 
in the SFP at OCNGS. The proposed RFOL and PDTS will continue to 
require proper control and monitoring of safety significant 
parameters and activities.
    The TS regarding SFP water level and spent fuel storage is 
retained to preserve the current requirements for safe storage of 
irradiated fuel. The restriction on the SFP water level is fulfilled 
by normal operating conditions and preserves initial conditions 
assumed in the analyses of the postulated DBA.
    The proposed amendment does not result in any new mechanisms 
that could initiate damage to the remaining relevant safety barriers 
for defueled plants (fuel cladding and spent fuel cooling). Since 
extended operation in a defueled condition will be the only 
operation allowed, and therefore bounded by the existing analyses, 
such a condition does not create the possibility of a new or 
different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes involve deleting and/or modifying certain 
TS once the OCNGS facility has been permanently shutdown, defueled, 
and at least 60 days of irradiated fuel decay time after reactor 
shutdown. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 license 
for OCNGS will no longer authorize operation of the reactor or 
emplacement or retention of fuel into the reactor vessel following 
submittal of the certifications required by 10 CFR 50.82(a)(1). As a 
result, the occurrence of certain design basis postulated accidents 
associated with reactor operation is no longer considered credible. 
The only remaining credible accidents are a FHA and the Postulated 
Radioactive Releases Due to Liquid Radwaste Tank Failures. The 
proposed amendment does not adversely affect the inputs or 
assumptions of any of the design basis analyses that impact either 
accident.
    The proposed changes are limited to those portions of the RFOL 
and TS that are not related to the safe storage of irradiated fuel. 
The requirements that are proposed to be revised or deleted from the 
OCNGS RFOL and 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 
will no longer be possible because the reactor will be permanently 
shutdown and defueled and OCNGS will no longer be authorized to 
operate the reactor.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves NSHC.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: Douglas A. Broaddus.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center, Linn County, Iowa
    Date of amendment request: June 9, 2017, as supplemented by letter 
dated November 1, 2017. Publicly-available versions are in ADAMS under 
Accession Nos. ML17164A076 and ML17305A910, respectively.
    Description of amendment request: The proposed amendment would 
replace the existing technical specification (TS) requirements related 
to ``operations with a potential for draining the reactor vessel'' 
(OPDRVs) with requirements for reactor pressure vessel (RPV) water 
inventory control (WIC) to protect Safety Limit 2.1.1.3. Safety Limit 
2.1.1.3 requires RPV water level to be greater than the top of active 
irradiated fuel. The proposed amendment is based on Technical 
Specification Task Force (TSTF) traveler TSTF-542, Revision 2, 
``Reactor Pressure Vessel Water Inventory Control,'' which was approved 
by the NRC by letter dated December 20, 2016.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold 
shutdown) and Mode 5 (i.e., refueling) is not an accident previously 
evaluated and, therefore, replacing the existing TS controls to 
prevent or mitigate such an event with a new set of controls has no 
effect on any accident previously evaluated. RPV water inventory 
control in Mode 4 or Mode 5 is not an initiator of any accident 
previously evaluated. The existing OPDRV controls or the proposed 
RPV WIC controls are not mitigating actions assumed in any accident 
previously evaluated.
    The proposed change reduces the probability of an unexpected 
draining event (which is not a previously evaluated accident) by 
imposing new requirements on the limiting time in which an 
unexpected draining event could result in the reactor vessel water 
level dropping to the top of the active fuel (TAF). These controls 
require cognizance of the plant configuration and control of 
configurations with unacceptably short drain times. These 
requirements reduce the probability of an unexpected draining event. 
The current TS requirements are only mitigating actions and impose 
no requirements that reduce the probability of an unexpected 
draining event.
    The proposed change reduces the consequences of an unexpected 
draining event (which is not a previously evaluated accident) by 
requiring an Emergency Core Cooling System (ECCS) subsystem to be 
operable at all times in Modes 4 and 5. The current TS requirements 
do not require any water injection systems, ECCS or otherwise, to be 
Operable in certain conditions in Mode 5. The change in requirement 
from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does 
not significantly affect the consequences of an unexpected draining 
event because the proposed Actions ensure equipment is available 
within the limiting drain time that is as capable of mitigating the 
event as the current requirements. The proposed controls provide 
escalating compensatory measures to be established as calculated 
drain times decrease, such as verification of a second method of 
water injection and additional confirmations that containment and/or 
filtration would be available if needed.

[[Page 2231]]

    The proposed change reduces or eliminates some requirements that 
were determined to be unnecessary to manage the consequences of an 
unexpected draining event, such as automatic initiation of an ECCS 
subsystem and control room ventilation. These changes do not affect 
the consequences of any accident previously evaluated since a 
draining event in Modes 4 and 5 is not a previously evaluated 
accident and the requirements are not needed to adequately respond 
to a draining event.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. The proposed change will not alter the design 
function of the equipment involved. Under the proposed change, some 
systems that are currently required to be operable during OPDRVs 
would be required to be available within the limiting drain time or 
to be in service depending on the limiting drain time. Should those 
systems be unable to be placed into service, the consequences are no 
different than if those systems were unable to perform their 
function under the current TS requirements.
    The event of concern under the current requirements and the 
proposed change is an unexpected draining event. The proposed change 
does not create new failure mechanisms, malfunctions, or accident 
initiators that would cause a draining event or a new or different 
kind of accident not previously evaluated or included in the design 
and licensing bases.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change replaces existing TS requirements related to 
OPDRVs with new requirements on RPV WIC. The current requirements do 
not have a stated safety basis 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: William Blair, P. O. Box 14000, Juno Beach, 
FL 33408-0420.
    NRC Branch Chief: David J. Wrona.
Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260, and 50-
296, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, Limestone 
County, Alabama
TVA, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Plant (SQN), Units 
1 and 2, Hamilton County, Tennessee
TVA, Docket Nos. 50-390 and 50-391, Watts Bar Nuclear Plant (WBN), 
Units 1 and 2, Rhea County, Tennessee
    Date of amendment request: November 17, 2017. A publicly-available 
version is in ADAMS under Accession Nos. ML17324A349.
    Description of amendment request: The amendments would add a new 
level of protection, ``Unbalanced Voltage,'' to the Technical 
Specifications for the loss of power instrumentation. The NRC issued 
Bulletin 2012-01, ``Design Vulnerability in Electric Power System,'' 
which requested addressees to submit specific information regarding 
plant design and operating configurations relative to the regulatory 
requirements of General Design Criterion (GDC) 17, ``Electric power 
systems.'' The Nuclear Energy Institute notified the NRC that the 
nuclear industry's chief nuclear officers approved a formal initiative 
to address the open phase condition (OPC). It further stated that the 
initiative represented a formal commitment among nuclear power plant 
licensees to address the OPC design vulnerability for operating 
reactors.
    The licensee stated, in its November 17, 2017, submittal, that the 
primary reason for the proposed change is to provide equipment 
protection from the effects of an unbalanced voltage in a similar 
fashion to the existing degraded and loss of voltage protection 
schemes. The identification of the vulnerability was based on industry 
operating experience and subsequent commitment to meet the voluntary 
Nuclear Strategic Issues Advisory Committee Open Phase Industry 
Initiative, also known as the ``Voluntary Industry Initiative'' (VII) 
for GDC 17 Compliance.
    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 consequence of an accident previously evaluated?
    Response: No.
    The proposed change to add a new unbalanced voltage relay (UVR) 
function at BFN, SQN, and WBN provides another level of undervoltage 
protection for the Class 1E electrical equipment. The new relay 
setpoints ensure that the normally operating Class 1E motors and 
equipment, which are powered from the Class 1E buses, are 
appropriately isolated from the normal offsite power source and 
would not be damaged in the event of sustained unbalanced voltage. 
The addition of the UVR function continues to allow the existing 
undervoltage protection circuitry to function as originally designed 
(i.e., degraded and loss of voltage protection remain in place and 
are unaffected by this change). The addition of the new UVR function 
has no impact on accident initiators or precursors; does not alter 
the accident analysis assumptions or the manner in which the plant 
is operated or maintained; and does not affect the probability of 
operator error.
    Based on the above, 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 to add a new UVR function at BFN, SQN, and 
WBN provides another level of undervoltage protection for the Class 
1E electrical equipment. This change ensures that the assumption in 
the previously evaluated accidents, which may involve a degraded 
voltage condition, continue to be valid. The proposed change does 
not result in the creation of any new accident precursors; does not 
result in changes to any existing accident scenarios; and does not 
introduce any operational changes or mechanisms that would create 
the possibility of a new or different kind of accident. The UVR 
function would not affect the existing loss of voltage and degraded 
voltage protection schemes, would not affect the number of 
occurrences of degraded voltage conditions that would cause the 
actuation of the existing Loss of Voltage Relays, Degraded Voltage 
Relays or the new UVRs; would not affect the failure rate of the 
existing protection relays; and would not impact the assumptions in 
any existing accident scenario.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

[[Page 2232]]

    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The current undervoltage protection circuitry is designed to 
isolate the normally operating Class 1E motors/equipment, which are 
powered from the Class 1E buses, from the offsite power source such 
that the subject equipment would not be damaged in the event of 
sustained degraded bus voltage. After the Class 1E buses are 
isolated from the offsite power supply, the Class 1E motors would be 
sequenced back on the Class 1E bus powered by the diesel generators 
(DGs) and continue to perform their design basis function to 
mitigate the consequences of an accident, with a specified margin of 
safety. With the addition of the new level of undervoltage 
protection, the capability of the Class 1E equipment is assured. 
Thus the equipment would continue to perform its design basis 
function to mitigate the consequences of the previously analyzed 
accidents and maintain the existing margin to safety currently 
assumed in the accident analyses. A DG start due to a safety 
injection signal (i.e., loss of coolant accident) and the subsequent 
sequencing of Class 1E loads back onto the Class 1E buses, powered 
by the DG, are not adversely affected by this change. If an actual 
loss of voltage condition were to occur on the Class 1E buses, the 
loss of voltage time delays would continue to isolate the Class 1E 
distribution system from the offsite power source prior to the DG 
assuming the Class 1E loads. The Class 1E loads would sequence back 
on the bus in a specified order and timer interval, again ensuring 
that the existing accident analysis assumptions remain valid and the 
existing margin to safety is unaffected.
    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Branch Chief: Undine Shoop.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3 (Indian Point 2 and 3), 
Westchester County, New York
    Date of amendment request: December 14, 2016, as supplemented by 
letters dated April 19, 2017; August 16, 2017; and October 2, 2017.
    Brief description of amendments: The amendments revised the 
Appendix A Technical Specifications Limiting Condition for Operation 
(LCO) 3.7.13, ``Spent Fuel Pit Storage,'' for Indian Point 2 and 
Appendix C Technical Specifications LCO 3.1.2, ``Shielded Transfer 
Canister (STC) Loading,'' for Indian Point 2 and 3. These LCOs ensure 
that the fuel to be loaded into the STC meets the design basis for the 
STC and has an acceptable rack location in the Indian Point 2 spent 
fuel pool before the STC is loaded with fuel. The proposed changes 
increase the population of Indian Point 3 fuel eligible for transfer 
via the STC to the Indian Point 2 spent fuel pool.
    Date of issuance: December 22, 2017.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 287 (Unit No. 2) and 264 (Unit No. 3). A publicly-
available version is in ADAMS under Accession No. ML17320A354; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Facility Operating License Nos. DPR-26 and DPR-64: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: June 19, 2017 (82 FR 
27885). The supplemental letters dated August 16, 2017, and October 2, 
2017, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the NRC staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 22, 2017.
    No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee 
Nuclear Power Station, Vernon, Vermont
    Date of amendment request: May 1, 2017, as supplemented by letter 
dated June 13, 2017.
    Brief description of amendment: The amendment revised the 
completion date for Milestone 8, full implementation of the Cyber 
Security Plan, from December 15, 2017, to July 31, 2019.
    Date of issuance: December 15, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 266. A publicly-available version is in ADAMS under 
Accession No. ML17339A097; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Operating License No. DPR-28: The amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: August 15, 2017 (82 FR 
38717). The supplemental letter dated June 13, 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 December 15, 2017.

[[Page 2233]]

    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 
and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and 
Lancaster Counties, Pennsylvania
    Date of amendment request: January 30, 2017, as supplemented by 
letters dated August 11, 2017, September 8, 2017, and December 20, 
2017.
    Brief description of amendments: The amendments replaced existing 
Technical Specification requirements related to ``operations with a 
potential for draining the reactor vessel'' with new requirements on 
reactor pressure vessel water inventory control to protect Safety Limit 
2.1.1.3. Safety Limit 2.1.1.3 requires reactor pressure vessel water 
level to be greater than the top of active irradiated fuel. The changes 
are based on Technical Specifications Task Force (TSTF) Traveler TSTF-
542, Revision 2, ``Reactor Pressure Vessel Water Inventory Control.''
    Date of issuance: December 27, 2017.
    Effective date: As of the date of issuance and shall be implemented 
prior to the Unit 2 fall 2018 refueling outage (P2R22).
    Amendments Nos.: 317 (Unit 2) and 320 (Unit 3). A publicly-
available version is in ADAMS under Accession No. ML17325B708; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: March 28, 2017 (82 FR 
15382). The supplemental letters dated August 11, 2017, September 8, 
2017, and December 20, 2017, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 27, 2017.
    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York
    Date of amendment request: April 5, 2017.
    Brief description of amendment: The amendment revised the Nine Mile 
Point Nuclear Station, Unit 2, Technical Specifications to allow 
greater flexibility in performing surveillance testing in Modes 1, 2, 
or 3 of emergency diesel generators. The changes are based on Technical 
Specifications Task Force (TSTF) Traveler TSTF-283A, Revision 3, 
``Modify Section 3.8 Mode Restrictions Notes.''
    Date of issuance: December 21, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 165. A publicly-available version of the amendment 
is in ADAMS under Accession No. ML17324B178; documents related to this 
amendment are listed in the Safety Evaluation enclosed with the 
amendment.
    Renewed Facility Operating License No. NPF-69: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: June 19, 2017 (82 FR 
27887).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 21, 2017.
    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station (OCNGS), Ocean County, New Jersey
    Date of amendment request: April 10, 2017, as supplemented by 
letters dated October 4 and December 15, 2017.
    Brief description of amendment: The amendment revised the OCNGS 
renewed facility operating license for the Cyber Security Plan (CSP) 
Milestone 8 full implementation completion date, as set forth in the 
CSP implementation schedule, and revised the physical protection 
license condition. The amendment revised the CSP Milestone 8 completion 
date from December 31, 2017, to August 31, 2021.
    Date of issuance: December 22, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 292. A publicly-available version is in ADAMS under 
Accession No. ML17289A222; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-16: The amendment 
revised the renewed facility operating license.
    Date of initial notice in Federal Register: May 23, 2017 (82 FR 
23626). The supplemental letters dated October 4 and December 15, 2017, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's 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 December 22, 2017.
    No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
    Date of amendment request: March 24, 2017.
    Brief description of amendments: The amendments revised the Donald 
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Technical Specification (TS) 
3.7.2, ``Steam Generator Stop Valves (SGSVs),'' to incorporate the SGSV 
actuator trains into the Limiting Condition for Operation statement and 
to provide associated Conditions, Required Actions, and Completion 
Times to the ACTIONS table. In addition, Surveillance Requirement (SR) 
3.7.2.2 was revised to clearly identify that the SGSV actuator trains 
are required to be tested in accordance with the SR.
    Date of issuance: December 19, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: Unit No. 1-338; Unit No. 2-320. A publicly-
available version is in ADAMS under Accession No. ML17312B030; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-58 and DPR-74: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: May 23, 2017 (82 FR 
23626).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 19, 2017.
    No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Houston County, 
Alabama
    Date of amendment request: November 22, 2016, as supplemented by 
letters dated May 23, 2017; June 8, 2017;

[[Page 2234]]

September 7, 2017; November 21, 2017; and December 18, 2017.
    Brief description of amendments: The amendments revised the 
licensing basis of FNP to support a full scope application of an 
Alternative Source Term methodology and modified Technical 
Specifications (TSs) 3.7.10, 3.9.3, and TS 5.5.18, consistent with 
Technical Specifications Task Force (TSTF) Travelers TSTF-448-A, 
``Control Room Habitability,'' Revision 3, and TSTF-312, 
``Administratively Control Containment Penetrations.''
    Date of issuance: December 20, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 216 (Unit 1) and 213 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17271A265; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-2 and NPF-8: The 
amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: January 3, 2017 (82 FR 
160). The supplemental letters dated May 23, 2017; June 8, 2017; 
September 7, 2017; November 21, 2017; and December 18, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 20, 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 (BFN), Limestone County, 
Alabama
    Date of amendment request: June 7, 2017. As supplemented by letters 
dated September 18 and October 23, 2017.
    Brief description of amendment: The amendments revised fire 
protection license condition 2.C.(13) for Unit 1, license condition 
2.C.(14) for Unit 2, and license condition 2.C.(7) for Unit 3.
    Date of issuance: December 19, 2017.
    Effective date: As of the date of issuance and shall be implemented 
as indicated in Items 2 and 3 under ``Transition License Conditions'' 
of the Operating Licenses, as shown in the attachment to the license 
amendments.
    Amendment Nos.: 302 (Unit 1), 326 (Unit 2), and 286 (Unit 3). A 
publicly-available version is in ADAMS under Accession No. ML17317A422; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the licenses.
    Date of initial notice in Federal Register: September 5, 2017 (82 
FR 41997). The supplemental letters dated September 18 and October 23, 
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 evaluations of the amendments are 
contained in Safety Evaluations dated December 19, 2017.
    No significant hazards consideration comments received: No.
Tennessee Valley Authority (TVA) Docket Nos. 50-259, 50-260, 50-296, 
and 72-052, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, 
Limestone County, Alabama
TVA Docket Nos. 50-327, 50-328, and 72-034, Sequoyah Nuclear Plant 
(SQN), Units 1 and 2, Hamilton County, Tennessee
TVA Docket Nos. 50-390, 50-391, and 72-1048, Watts Bar Nuclear Plant 
(WBN), Units 1 and 2, Rhea County, Tennessee
    Date of amendment request: January 4, 2017, as supplemented by 
letter dated July 7, 2017.
    Brief description of amendments: The amendments revised TVA 
Emergency Plans for the above nuclear plants. Specifically, they 
adopted the NRC-endorsed Radiological Emergency Plan Emergency Action 
Level schemes developed by the Nuclear Energy Institute (NEI 99-01, 
Revision 6, ``Development of Emergency Action Levels for Non-Passive 
Reactors'').
    Date of issuance: December 22, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days from the date of its issuance or July 3, 2018, 
whichever comes later.
    Amendment Nos.: BFN, 303 (Unit 1), 327 (Unit 2), and 287 (Unit 3); 
SQN, 339 (Unit 1) and 332 (Unit 2); and WBN, 118 (Unit 1) and 18 (Unit 
2). A publicly-available version is in ADAMS under Accession No. 
ML17289A032; documents related to these amendments are listed in the 
Safety Evaluations (SEs) enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-33, DPR-52, DPR-68, 
DPR-77, DPR-79 and Facility Operating License Nos, NPF-90 and NPF-96: 
Amendments revised the licenses.
    Date of initial notice in Federal Register: June 19, 2017 (82 FR 
27891). The supplemental letter dated July 7, 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 SEs dated December 22, 2017.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, on January 8, 2018.

    For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Acting Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2018-00386 Filed 1-12-18; 8:45 am]
 BILLING CODE 7590-01-P


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