Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 66224-66238 [2019-25972]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 84, Number 232 (Tuesday, December 3, 2019)]
[Notices]
[Pages 66224-66238]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2019-25972]


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NUCLEAR REGULATORY COMMISSION

[NRC-2019-0238]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to the Atomic Energy Act of 1954, as amended (the 
Act), the U.S. Nuclear Regulatory Commission (NRC) is publishing this 
regular biweekly notice. The Act requires the Commission to publish 
notice of any amendments issued, or proposed to be issued, and grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license or combined license, as 
applicable, upon a determination by the Commission that such amendment 
involves no significant hazards consideration, notwithstanding the 
pendency before the Commission of a request for a hearing from any 
person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from November 5, 2019 to November 18, 2019. The 
last biweekly notice was published on November 19, 2019.

DATES: Comments must be filed by January 2, 2020. A request for a 
hearing must be filed by February 3, 2020.

ADDRESSES: You may submit comments by any of the following methods:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2019-0238. Address 
questions about NRC dockets IDs in Regulations.gov to Jennifer Borges; 
telephone: 301-287-9127; email: [email protected]. For technical 
questions, contact the individual listed in the FOR FURTHER INFORMATION 
CONTACT section of this document.
     Mail comments to: Office of Administration, Mail Stop: 
TWFN-7-A60M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, ATTN: Program Management, Announcements and Editing Staff.
    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: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2019-0238, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2019-0238.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``Begin Web-based ADAMS 
Search.'' For problems with ADAMS, please contact the NRC's Public 
Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or 
by email to [email protected]. The ADAMS accession number for each 
document referenced (if it is available in ADAMS) is provided the first 
time that it is mentioned in this document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2019-0238, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at 
https://www.regulations.gov as well as enter the comment submissions 
into ADAMS. The NRC does not routinely edit comment submissions to 
remove identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

II. Background

    Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the 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.

[[Page 66225]]

III. 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 section 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 https://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 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

[[Page 66226]]

its boundaries. Alternatively, a State, local governmental body, 
Federally-recognized Indian Tribe, or agency thereof may participate as 
a non-party under 10 CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
the discretion of the presiding officer, be permitted to make a limited 
appearance pursuant to the provisions of 10 CFR 2.315(a). A person 
making a limited appearance may make an oral or written statement of 
his or her position on the issues but may not otherwise participate in 
the proceeding. A limited appearance may be made at any session of the 
hearing or at any prehearing conference, subject to the limits and 
conditions as may be imposed by the presiding officer. Details 
regarding the opportunity to make a limited appearance will be provided 
by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate under 10 CFR 2.315(c), must be filed in accordance with the 
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 
46562; August 3, 2012). The E-Filing process requires participants to 
submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC website at https://www.nrc.gov/site-help/e-submittals.html. Participants may not submit 
paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at 301-415-1677, to (1) request a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign submissions and access the E-Filing 
system for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a petition or 
other adjudicatory document (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public website at https://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 https://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 https://www.nrc.gov/site-help/e-submittals.html, by 
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 6 p.m., Eastern Time, Monday through Friday, excluding government 
holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
stating why there is good cause for not filing electronically and 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, 11555 
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and 
Adjudications Staff. Participants filing adjudicatory documents in this 
manner are responsible for serving the document on all other 
participants. Filing is considered complete by first-class mail as of 
the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service. A presiding officer, having granted an 
exemption request from using E-Filing, may require a participant or 
party to use E-Filing if the presiding officer subsequently determines 
that the reason for granting the exemption from use of E-Filing no 
longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the 
Commission or the presiding officer. If you do not have an NRC-issued 
digital ID certificate as described above, click ``cancel'' when the 
link requests certificates and you will be automatically directed to 
the NRC's electronic hearing dockets where you will be able to access 
any publicly available documents in a particular hearing docket. 
Participants are requested not to include personal privacy information, 
such as social security numbers, home addresses, or personal phone 
numbers in their filings, unless an NRC regulation or other law 
requires submission of such information. For example, in some 
instances, individuals provide home addresses in order to demonstrate 
proximity to a facility or site. With respect to copyrighted works, 
except for limited excerpts that serve the purpose of the adjudicatory 
filings and would constitute a Fair Use application, participants are 
requested not to include copyrighted materials in their submission.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing

[[Page 66227]]

information related to this document, see the ``Obtaining Information 
and Submitting Comments'' section of this document.

DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: September 5, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19248C571.
    Description of amendment request: The amendment would revise the 
Fermi 2 Technical Specification (TS) 2.1.1, ``Reactor Core SLs [safety 
limits],'' reactor steam dome pressure from 785 psig [pounds per square 
inch gauge] to 686 psig and TS Table 3.3.6.1-1, ``Primary Containment 
Isolation Instrumentation,'' Function 1.b, ``Main Steam Line Pressure--
Low,'' isolation function allowable value from 736 psig to 801 psig.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because decreasing the reactor steam dome pressure in TS Safety 
Limits 2.1.1.1 and 2.1.1.2 for reactor thermal power ranges and 
increasing the trip set point and allowable value for main steam 
line low pressure isolation effectively expands the validity range 
for GEXL critical power correlation and the calculation of minimum 
critical power ratio. The critical power ratio rises during the 
pressure reduction following the scram that terminates the PRFO 
[pressure regulator failure--Open] transient. The reduction in 
reactor steam dome pressure value in the SL and the increase in trip 
set point and the reactor steam dome pressure value in the SL and 
the increase in the trip set point and the allowable value for the 
main steam line low pressure isolation provides adequate margin to 
accommodate the pressure reduction during the PRFO transient within 
the revised TS limit.
    The proposed changes do not alter the use of the analytical 
methods used to determine the safety limits that have been 
previously reviewed and approved by the NRC. The proposed changes 
are in accordance with an NRC approved critical power correlation 
methodology and do not adversely affect accident initiators or 
precursors.
    The proposed changes do not alter or prevent the ability of 
structures, systems, and components from performing their intended 
function to mitigate the consequences of an initiating event within 
the applicable acceptance limits. The proposed changes are 
consistent with the safety analysis and resultant consequences.
    Based on the above, DTE has concluded that the proposed change 
will not result in 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 changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed reduction in the reactor dome steam pressure 
value in the safety limit in conjunction with the increase in the 
trip setpoint and the allowable value for the main steam line low 
pressure isolation reflects a wider range of applicability for the 
GEXL critical power correlation which is approved by the NRC.
    In addition, no new failure modes are being introduced. There 
are no changes in the method by which any plant systems perform a 
safety function. No new accident scenarios, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
changes.
    The proposed changes do not introduce any new accident 
precursors, nor do they involve any changes in the methods governing 
normal plant operation. The proposed changes do not alter the 
outcome of the safety analysis.
    Based on the above, DTE has concluded that the proposed TS 
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 margin of safety is established through the design of the 
plant structures, systems, and components, and through the 
parameters for safe operation and setpoints for actuation of 
equipment relied upon to respond to transients and design basis 
accidents. Evaluation of the 10 CFR part 21 condition by General 
Electric determined that since the Minimum Critical Power Ratio 
improves during the PRFO transient, there is no decrease in the 
safety margin and therefore there is not a threat to fuel cladding 
integrity. The proposed change in reactor steam dome pressure limits 
supports the current safety margin, which protects the fuel cladding 
integrity during a depressurization transient, but does not change 
the requirements governing operation or availability of safety 
equipment assumed to operate to preserve the margin of safety. The 
change does not alter the behavior of plant equipment, which remains 
unchanged. By raising the MSL LPIS AV [main steamline, low-pressure 
injection system, allowable value] in conjunction with lowering the 
Reactor Steam Dome Pressure SL, there is an increase in margin which 
increases protection of the MCPR [maximum critical power ratio].
    The proposed change to Reactor Core SLs 2.1.1.1 and 2.1.1.2 is 
consistent with and within the capabilities of the applicable NRC 
approved critical power correlation for the fuel designs in use at 
Fermi 2. The proposed change does not alter the manner in which the 
SLs are determined. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    The reduction in value of the reactor steam dome pressure safety 
limit and the increase in the trip setpoint and allowable value for 
main steam line low pressure isolation provides adequate margin to 
accommodate the pressure reduction during the PRFO transient within 
the revised TS limit.
    Based on the above, DTE has concluded that the proposed TS 
change does 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Jon P. Christinidis, DTE Energy, 688 WCB, 
One Energy Plaza, Detroit, MI 48226.
    NRC Branch Chief: Nancy L. Salgado.

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

    Date of amendment request: July 29, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19210D020.
    Description of amendment request: The amendment would revise H. B. 
Robinson Steam Electric Plant, Unit No. 2, Technical Specification (TS) 
3.7.3 regarding main feedwater isolation valves, main feedwater 
regulation valves, and bypass valves, by making the TS applicable to 
three additional feedwater bypass valves. The amendment would also 
revise the condition and completion time associated with the feedwater 
bypass valves.
    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 does not modify the feedwater system, nor 
does it make any physical or operational changes to the facility. 
The new non-safety BVs [bypass valves] are being installed under 10 
CFR 50.59 to provide a backup isolation function to the existing 
safety grade BVs, consistent with NUREG-0138 and Section 6.2.1.4 of 
the NRC's Standard Review Plan. The new BVs will receive the same 
Engineered Safety Features signals to close and they will be

[[Page 66228]]

subject to the same testing as the existing safety grade BVs. The 
proposed change has no impact on the containment or accident 
analyses. Inclusion of the new BVs within the scope of TS 3.7.3 
subjects them to the same TS LCO [limiting condition for operation] 
and Surveillance Requirements as the existing BVs and allows them to 
be credited as backups to the existing BVs.
    Extending the Completion Time of TS 3.7.3, Required Action C.1 
from 8 hours to 72 hours is not an accident initiator and thus does 
not change the probability that an accident will occur; however, it 
could potentially affect the consequences of an accident if the 
accident occurred during the extended unavailability of an 
inoperable BV. The new BVs provide redundant isolation in the 
feedwater bypass flow paths. This represents a safety improvement 
over the original single BV (per flow path) design. The proposed 
increase in time an inoperable BV is allowed to remain open/
unisolated is small and the probability of an event requiring 
isolation of the feedwater flow path occurring during this period, 
coincident with a failure of the redundant BV in that flow path, is 
low.
    Therefore, the proposed TS changes do not significantly increase 
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 modify the feedwater system, nor 
does it make any physical or operational changes to the facility. 
Neither the inclusion of the new BVs in TS 3.7.3 nor the extension 
of the Completion Time for TS 3.7.3 Required Action C.1 results in 
any new failure modes or affects. The new non-safety BVs are being 
installed under 10 CFR 50.59 to provide a backup isolation function 
to the existing safety grade BVs. Closure of the BVs is required to 
mitigate the consequences of steam line and feedwater line break 
events. The proposed changes allow for the new BVs to be credited in 
plant analyses for the isolation feedwater flow in the event of a 
failure of the existing BVs to close.
    Therefore, the possibility of a new or different kind of 
accident from any kind of accident previously evaluated is not 
created.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment does not involve: (1) A physical 
alteration of the plant, (2) a change to any set points for 
parameters associated with protection or mitigation actions nor (3) 
any impact on the fission product barriers or parameters associated 
with licensed safety limits. The new BVs are being installed under 
10 CFR 50.59 to provide a backup isolation function to the existing 
BVs. There are no changes to either the containment analysis or to 
the analysis for any design basis event.
    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: Kathryn B. Nolan, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tryon Street, DEC45A, Charlotte, NC 
28202.
    NRC Branch Chief: Undine Shoop.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
2 (ANO-2), Pope County, Arkansas

    Date of amendment request: August 29, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19241A264.
    Description of amendment request: The proposed amendment would 
modify multiple Technical Specifications (TSs) for ANO-2 to address 
non-conservative TSs associated with the movement of fuel assemblies. 
This proposed change is necessary due to the previous adoption of the 
Alternate Source Terms, which included an update to the ANO-2 fuel 
handling accident (FHA) analysis. This update created a new requirement 
to address the movement of new (unirradiated) fuel assemblies over 
irradiated fuel assemblies. The proposed amendment would also adopt 
certain changes to gain greater consistency with NUREG-1432, Revision 
4, ``Standard Technical Specifications, Combustion Engineering 
Plants.'' The changes necessary to support the revised FHA affect 
similar TSs associated with Technical Specifications Task Force (TSTF) 
Standard Technical Specification Change Travelers TSTF-51, Revision 2, 
``Revise Containment Requirements During Handling Irradiated Fuel and 
Core Alterations''; TSTF-272, Revision 1, ``Refueling Boron 
Concentration Clarification''; TSTF-286, Revision 2, ``Operations 
Involving Positive Reactivity Additions''; TSTF 471, Revision 1, 
``Eliminate Use of Term Core Alterations in ACTIONS and Notes''; and 
TSTF-571-1, Revision 0, ``Revise Actions for Inoperable Source Range 
Neutron Flux Monitor.'' Therefore, the licensee proposes to adopt these 
TSTFs in conjunction with changes necessary to support the revised FHA 
analysis. Additionally, the proposed amendment would incorporate 
specified administrative and editorial changes associated with the TS 
pages affected by the aforementioned proposed changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. Each of the six items described above is addressed under 
each of the three standards, 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.

Updated FHA [Analysis]

    TS changes associated with the updated FHA analysis ensure the 
initial assumptions of the FHA are maintained and, therefore, act to 
minimize the consequences of an accident by ensuring TS required 
features are operable during the movement of fuel assemblies. The 
updated FHA analysis was previously accepted by the NRC during 
adoption of Alternate Source Terms (AST) for ANO-2. The probability 
of a fuel assembly drop (or any load drop) is unchanged by the 
updated FHA analysis. Therefore, the updated FHA analysis does not 
involve a significant increase in the probability of an accident 
previously evaluated.
    Entergy has reviewed station procedures and controls in order to 
verify that no other loads, other than a new or irradiated fuel 
assembly, need be addressed with regard to an FHA (i.e., no other 
known load carried over irradiated fuel assemblies exists which 
would not be bounded by the fuel drop analysis or be expected to 
cause fuel damage if dropped). The proposed TS changes ensure 
required systems are operable during operations that could lead to 
an FHA. As previously approved by the NRC via the adoption of AST 
for ANO-2, the updated FHA analysis adequately bounds Control Room 
and offsite dose within federal limitations. Based on the above, the 
proposed FHA-related changes to the TSs do not result in a 
significant increase in the consequences of an accident previously 
evaluated. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

TSTF-51 and TSTF 471

    The design basis accident (DBA) assumed for ANO-2 related to the 
proposed changes is the FHA. The boron dilution event is evaluated 
in the ANO-2 Safety Analysis Report (SAR), but [is] considered an 
unlikely event due to the time available for operator detection and 
response, along with prevalent administrative controls. A loss of 
Shutdown Cooling (SDC) event has little relationship to and minimal 
impact with regard to an FHA. TSTF-51 and TSTF-471 replace the use 
of the previously defined ``core alterations'' term with 
requirements associated with the movement of fuel assemblies, since 
the drop of a fuel assembly is the only event that could reasonably 
lead to an FHA or a significant challenge to the plant.
    In addition, TSTF-51 reduces restrictions following sufficient 
radioactive decay of fuel assemblies since the offsite dose 
consequences of an FHA following this decay period (100 hours for 
ANO-2) would remain within 10 CFR 50.67 limits. Note that this 
allowance is not adopted for TS Control

[[Page 66229]]

Room ventilation or radiation monitoring systems (associated with 
meeting 10 CFR 50, appendix A, General Design Criteria (GDC) 19).
    The removal of references to ``core alterations'' in favor of 
restrictions associated with the movement of fuel assemblies 
eliminates current restrictions associated with the manipulation of 
other core components (i.e., sources or reactivity control 
components within the core) since such manipulation cannot result in 
an FHA, boron dilution event, or loss of SDC. In addition, 
manipulation of these other components cannot present a significant 
challenge to shutdown margin (SDM) because the TS required RCS boron 
concentration for Mode 6 operation provides substantial margin to 
criticality.
    Changes associated with TSTF-51 and TSTF-471, as adopted, do not 
modify limitations in such a way that the consequences of an FHA 
would be greater than that assumed in the updated FHA analysis 
(i.e., 10 CFR 50.67 and GDC 19 limitations are not exceeded 
following an FHA).
    Based on the above, the proposed changes associated with the 
adoption of TSTF-51 and TSTF-471 do not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.

TSTF-272

    Changes associated with TSTF-272 place additional restrictions 
on Mode 6 operations by ensuring the boron concentration of the 
water in the refueling canal meets the same TS limits required for 
the Reactor Coolant System (RCS) when the RCS is in direct hydraulic 
communication with the refueling canal (i.e., reactor vessel head 
removed and refueling canal filled). These changes are unrelated to 
any accident initiator and further prohibit any challenge to the 
fuel in the reactor vessel by ensure sufficient boron concentration 
is maintained during Mode 6 operations. Therefore, these changes do 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.

TSTF-286

    Changes associated with TSTF-286 permit operator control of RCS 
inventory and temperature when certain TS requirements are not met, 
provide[d] the overall required SDM of the RCS is maintained. The 
activities that involve inventory makeup from sources with boron 
concentrations less than the current RCS concentration (i.e., boron 
dilution) need not be precluded in the TSs provided the required SDM 
is maintained for the worst-case overall effect on the core. Note 
that an unexpected boron dilution event is considered unlikely for 
ANO-2 due to the significant period of time for operator detection 
and response before SDM would be significantly challenged (reference 
ANO-2 Safety Analysis Report Section 15.1.4.3). In addition, while a 
boron dilution event is evaluated in the accident analysis, the only 
``accident'' assumed for ANO-2 during Mode 6 operations is the FHA. 
Permitting RCS inventory and temperature adjustments is unrelated to 
any assumptions associated with an FHA. Therefore, these changes do 
not result in a significant increase in the probability an accident 
(or a boron dilution event) previously evaluated. Because an 
unexpected boron dilution event provides sufficient opportunity for 
detection and recovery, the proposed changes associated with TSTF-
286 likewise do not result in a significant increase in the 
consequences of an accident (or boron dilution event) previously 
evaluated.

TSTF-571-T

    The proposed change revises the Actions for inoperable source 
range neutron flux monitors to prohibit the movement of fuel 
assemblies, sources, and reactivity control components when [a] 
monitor is inoperable. The Actions taken when a monitor is 
inoperable are not initiators to any accident previously evaluated. 
The monitors are not credited to mitigate any previously evaluated 
accident. The proposed change restricts the licensee's actions while 
a monitor is inoperable beyond the current requirements. Therefore, 
the consequences of an accident previously evaluated are not 
significantly increased.

Administrative/Editorial/Miscellaneous Changes

    Enhancements and administrative changes proposed for TSs 
affected by the previously discussed updated FHA or changes 
associated with increasing consistency with the ITS [improved 
technical specifications] are unrelated to any accident initiator. 
Administrative changes likewise cannot impact the consequences of 
any accident previously evaluated.
    The following is a listing of other changes proposed in this 
amendment request which modify the TSs (not considered within the 
editorial/administrative realm).
     A new Note 3 is proposed that clarifies the original 
intent of the TS requirements for radiation monitoring and automatic 
isolation of the Containment Purge system. As written, the TS would 
require the radiation monitoring and isolation capability to remain 
operable even when the Containment Purge system is secured. The 
addition of Note 3 specifies that operability is required only 
during (1) Containment Purge operations, or (2) ongoing Containment 
Building continuous ventilation operations when moving recently 
irradiated fuel assemblies or moving new fuel assemblies over 
irradiated fuel assemblies in the Containment Building, consistent 
with the updated FHA and TSTF-51. Other associated enhancements are 
made to the Containment Purge requirements in support of the above 
changes or to provide additional clarification.
     The phrase ``elevation corresponding to the'' top of 
irradiated fuel is added to the Limiting Condition for Operation 
(LCO) of TS 3.9.9, ``Water Level--Reactor Vessel.'' This ensures 
that proper water level is established prior to initiating refueling 
of the reactor core following a defueled condition.
     The movement of fuel ``within the reactor vessel'' 
contained in the Applicability and Action of TS 3.9.9 is revised to 
``within the Containment Building.'' This reference is also added to 
the Surveillance Requirement. The required water level should be met 
even when fuel is being moved in other areas of the refueling canal, 
not just in the reactor vessel. In addition, the phrase ``while in 
Mode 6'' is deleted from the Applicability since fuel assemblies 
cannot physically be removed from the reactor until Mode 6 has been 
achieved.
    Enhancements associated with the Containment Purge system 
radiation instrumentation ensure Surveillance testing is performed 
when the system is in service, regardless if an actual Purge is 
taking place. In addition, the proposed changes ensure appropriate 
testing is performed prior to placing the system in service each 
refueling outage. The proposed changes are neutral or more 
restrictive and, therefore, cannot increase the consequences of an 
accident previously evaluated.
    Clarifications to limitations on refueling water level and the 
location of fuel assemblies are more restrictive changes, ensuring 
proper controls have been established before activities are 
commenced. No impact to the consequences of any accident result from 
these changes. The changes to these TSs, in addition to the 
aforementioned changes to Containment Purge requirements, do not 
increase the probability of an accident occurring.
    Based on the above, the proposed changes do not represent 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.

Updated FHA [Analysis]

    TS changes associated with the updated FHA [analysis] involve no 
physical changes to the plant. These changes act to ensure required 
structures, systems, and components (SSCs) are operable when moving 
irradiated fuel assemblies or new fuel assemblies over irradiated 
fuel assemblies to limit any Control Room or offsite dose 
consequences to within acceptable limits. Therefore, these changes 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.

TSTF-51 and TSTF 471

    TS changes associated with ITS improvements related to these 
TSTFs involve no physical changes to the plant. The removal of 
references to ``core alterations'' in favor of restrictions 
associated with the movement of fuel assemblies eliminates current 
restrictions associated with the manipulation of other core 
components (i.e., sources or reactivity control components within 
the core). Such manipulations cannot result in an FHA, boron 
dilution event, or loss of SDC. In addition, such manipulations 
cannot result in an appreciable change in core reactivity due to the 
high RCS boron concentration required during refueling operations by 
the TSs. TSTF-51 changes associated with a reduction in restrictions 
following sufficient radioactive decay of fuel assemblies are not 
considered accident precursors. The proposed changes do not 
introduce a new accident initiator, accident precursor, or accident-
related malfunction

[[Page 66230]]

mechanism. Therefore, these changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

TSTF-272

    Changes associated with TSTF-272 place additional restrictions 
on Mode 6 operations by ensuring the boron concentration of the 
water in the refueling canal meets the same TS limits required for 
the RCS when the RCS is in direct hydraulic communication with the 
refueling canal (i.e., reactor vessel head removed and refueling 
canal filled). These changes are unrelated to any accident initiator 
and further prohibit any challenge to the fuel in the reactor vessel 
by [ensuring] sufficient boron concentration is maintained during 
Mode 6 operations. The proposed changes do not introduce a new 
accident initiator, accident precursor, or accident-related 
malfunction mechanism. Therefore, these changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

TSTF-286

    Changes associated with TSTF-286 permit operator control of RCS 
inventory and temperature when certain TS requirements are not met, 
provide[d] the overall required SDM of the RCS is maintained. No 
physical plant changes are related to these TS changes. The only 
accident or event that could be affected by this change is the boron 
dilution event, which has been previously evaluated. The proposed 
changes do not introduce a new accident initiator, accident 
precursor, or accident-related malfunction mechanism. Therefore, 
these changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

TSTF-571-T

    The proposed change revises the Actions for inoperable source 
range neutron flux monitors to prohibit the movement of fuel 
assemblies, sources, and reactivity control components when a 
monitor is inoperable. The proposed change does not involve a 
physical alteration of the plant (no new or different type of 
equipment will be installed). No credible new failure mechanisms, 
malfunctions, or accident initiators that would have been considered 
a design basis accident in the ANO-2 Safety Analysis Report (SAR) 
are created.

Administrative/Editorial/Miscellaneous Changes

    Enhancements and administrative changes proposed for TSs 
affected by the above updated FHA or ITS improvements are unrelated 
to any accident initiator and involve no physical changes to the 
plant.
    Enhancements associated with the Containment Purge system 
radiation instrumentation ensure Surveillance testing is performed 
when the system is in service, regardless if an actual Purge is 
taking place. In addition, the proposed changes ensure appropriate 
testing is performed prior to placing the system in service each 
refueling outage. Clarifications to limitations on refueling water 
level and the location of fuel assemblies are more restrictive 
changes, ensuring proper controls have been established before 
activities are commenced.
    The proposed changes do not introduce a new accident initiator, 
accident precursor, or accident-related malfunction mechanism. Based 
on the above, these changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.

Updated FHA [Analysis]

    TS changes associated with the updated FHA [analysis] act to 
ensure required SSCs are operable when moving irradiated fuel 
assemblies or new fuel assemblies over irradiated fuel assemblies to 
limit any Control Room or offsite dose consequences to within 
acceptable limits. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

TSTF-51 and TSTF 471

    The removal of references to ``core alterations'' in favor of 
restrictions associated with the movement of fuel assemblies 
eliminates current restrictions associated with the manipulation of 
other core components (i.e., sources or reactivity control 
components within the core). Such manipulations cannot result in an 
FHA, boron dilution event, or loss of SDC. In addition, such 
manipulations cannot result in an appreciable change in core 
reactivity due to the high RCS boron concentration required during 
refueling operations by the TSs. TSTF-51 also reduces restrictions 
following sufficient radioactive decay of fuel assemblies since the 
consequence of an FHA following this decay period would remain 
within 10 CFR 50.67 limits. Note that this allowance is not adopted 
for Control Room ventilation or radiation monitoring systems 
(governed under GDC 19). Changes associated with TSTF-51 and TSTF-
471, as adopted, do not modify limitations in such a way that the 
consequences of an FHA would be greater than that assumed in the FHA 
analysis (i.e., 10 CFR 50.67 and GDC 19 limitations are not exceeded 
following an FHA). Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

TSTF-272

    Changes associated with TSTF-272 place additional restrictions 
on Mode 6 operations by ensuring the boron concentration of the 
water in the refueling canal meets the same TS limits required for 
the RCS when the RCS is in direct hydraulic communication with the 
refueling canal (i.e., reactor vessel head removed and refueling 
canal filled). These changes are more restrictive than the current 
TS and, therefore, do not involve a significant reduction in a 
margin of safety.

TSTF-286

    Changes associated with TSTF-286 permit operator control of RCS 
inventory and temperature when certain TS requirements are not met, 
provide the overall required SDM of the RCS is maintained. The only 
accident or event that could be affected by this change is the boron 
dilution event which has been previously evaluated. While the margin 
between existing boron concentration and that required to meet SDM 
requirements may be reduced, margin is gained by permitting 
operators to take corrective action to maintain RCS inventory and 
temperature within limits during periods when such operations are 
otherwise prohibited. While not quantifiable, the changes associated 
with TSTF-286 have a general balanced effect in relation to the 
margin of safety. Because an unexpected boron dilution event 
provides sufficient opportunity for detection and recovery, the 
proposed changes associated with TSTF-286 do not involve a 
significant reduction in a margin of safety.

TSTF-571-T

    The proposed change revises the Actions for inoperable source 
range neutron flux monitors to prohibit the movement of fuel 
assemblies, sources, and reactivity control components when a 
monitor is inoperable. No safety limits are affected. No Limiting 
Conditions for Operation or Surveillance limits are affected. The 
design, operation, surveillance methods, and acceptance criteria 
specified in applicable codes and standards (or alternatives 
approved for use by the NRC) continue to be met as described in the 
plants' [plant's] licensing basis. The proposed change does not 
adversely affect existing plant safety margins, or the reliability 
of the equipment assumed to operate in the safety analysis. As such, 
there are no changes being made to safety analysis assumptions, 
safety limits, or limiting safety system settings that would 
adversely affect plant safety. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.

Administrative/Editorial/Miscellaneous Changes

    Enhancements and administrative changes proposed for TSs 
affected by the above updated FHA or ITS improvements are unrelated 
to any accident initiator or mitigation strategy. Enhancements 
associated with the Containment Purge system radiation 
instrumentation ensure Surveillance testing is performed when the 
system is in service, regardless if an actual Purge is taking place. 
In addition, the proposed changes ensure appropriate testing is 
performed prior to placing the system in service each refueling 
outage. Clarifications to limitations on refueling water level and 
the location of fuel assemblies are more restrictive changes, 
ensuring proper controls have been established before activities are 
commenced. Based on the above, these proposed changes do not involve 
a significant reduction in a margin of safety.
    Therefore, the proposed changes contained within this amendment 
request do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 66231]]

    Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, 
Entergy Services, LLC, 101 Constitution Avenue NW, Suite 200 East, 
Washington, DC 20001.
    NRC Branch Chief: Jennifer L. Dixon-Herrity.

Exelon FitzPatrick, LLC and Exelon Generation Company, LLC, Docket No. 
50-333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New 
York

    Date of amendment request: September 12, 2019. A publicly available 
version is in ADAMS under Accession No. ML19255D988.
    Description of amendment request: The amendment would revise 
Technical Specifications related to primary containment hydrodynamic 
loads.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise operating limits for containment 
systems during normal operation that provide the initial conditions 
at which containment performance to mitigate loss-of-coolant 
accidents is evaluated. The affected parameters are unrelated to the 
Reactor Coolant Pressure Boundary or reactivity control systems and 
therefore are unrelated to accident initiation or probability of 
occurrence.
    Analysis has demonstrated that the containment will continue to 
operate within design limits in the event of an accident. Therefore, 
the consequences of an accident are not significantly affected by 
the proposed change.
    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 do not alter the protection system design, 
create new failure modes, or change any modes of operation. The 
proposed changes do not involve a physical alteration of the plant; 
and no new or different kind of equipment will be installed. 
Consequently, there are no new initiators that could result in a new 
or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes will eliminate the 1.7 psi [pounds per 
square inch] differential pressure requirement between the drywell 
and wetwell, raise the maximum torus water level to 14.25 ft, and 
raise the HPCI [high pressure coolant injection] ``Suppression Pool 
Water Level--High'' Allowable Value to <= [less than or equal to] 
14.75 ft. Technical Report ``13-0541-TR-002'' evaluated use of these 
operating parameters and determined that all structural elements 
continue to meet code requirements with adequate margin. Other 
design aspects such as Emergency Core Cooling System Pump Net 
Positive Suction Head, Equipment Qualification, and accident 
radiological dose impacted by the proposed changes were also 
evaluated and found to have negligible to no impact.
    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: Donald P. Ferraro, Assistant General 
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Suite 305, 
Kennett Square, PA 19348.
    NRC Branch Chief: James G. Danna.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station (CNS), Nemaha County, Nebraska

    Date of amendment request: August 19, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19238A065.
    Description of amendment request: The proposed amendment would 
revise CNS Technical Specification 5.5.12, ``Primary Containment 
Leakage Rate Testing Program,'' to allow for an exception to certain 
leak rate testing interval requirements of the program. Specifically, 
the proposed amendment would permit the 10 CFR part 50, appendix J, 
Option B leak testing of Type C residual heat removal system heat 
exchanger relief valves and their associated Type B testable discharge 
flange tests be performed at the same frequency as the visual 
examination, seat leakage testing, and set pressure testing performed 
for these valves under the requirements of the Inservice Testing 
Program per 10 CFR 50.55a(f).
    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 certain leak testing intervals 
required by the CNS primary containment leakage rate testing program 
to be aligned with certain testing intervals required by the 
Inservice Testing Program under 10 CFR50.55a(f). The containment 
function is solely to mitigate the consequences of an accident. No 
design basis accident is initiated by a failure of the containment 
leakage mitigation function. Aligning the testing interval 
requirements of the two programs does not create any adverse 
interactions with other systems that could result in initiation of a 
design basis accident. Continued containment integrity is assured by 
the established programs for local leakage rate testing and 
inservice testing which are unaffected by the proposed change.
    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 allows certain leak testing intervals 
required by the CNS primary containment leakage rate testing program 
to be aligned with certain testing intervals required by the 
Inservice Testing Program under 10 CFR 50.55a(f). This proposed 
change does not modify existing structures, systems, or components 
(SSC) of the plant, and it does not introduce new SSC's. The plant 
will continue to be operated in the same manner. Thus, it does not 
affect the design function or operation of SSC's involved, and it 
does not introduce a new accident initiator.
    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 allows certain leak testing intervals 
required by the CNS primary containment leakage rate testing program 
to be aligned with certain testing intervals required by the 
Inservice Testing Program under 10 CFR 50.55a(f). The proposed 
alignment of testing intervals will not result in a change to the 
design or operation of any plant SSC used to shutdown the plant, 
initiate Emergency Core Cooling systems, or isolate the ability of 
CNS to mitigate any accident or transient. There is no impact on 
safety limits or limiting safety system settings. The change does 
not affect any plant safety parameters or setpoints.
    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

[[Page 66232]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Jennifer Dixon-Herrity.

NextEra Energy Duane Arnold (NEDA), LLC, Docket No. 50-331, Duane 
Arnold Energy Center (DAEC), Linn County, Iowa

    Date of amendment request: June 20, 2019, as supplemented by 
letters dated September 12, 2019, and November 4, 2019. Publicly-
available versions are in ADAMS under Accession Nos. ML19176A356, 
ML19261A141, and ML19308A085, respectively.
    Description of amendment request: The NRC staff has previously made 
a proposed determination that the amendment request dated June 20, 
2019, involves no significant hazards consideration (84 FR 45544; 
August 29, 2019). Subsequently, the licensee provided additional 
information that expanded the scope of the amendment request as 
originally noticed. In the supplemental letter dated September 12, 
2019, the licensee provided no significant hazards consideration for 
the supplemental changes only. This notice combines the two no 
significant hazards considerations provided by the licensee. 
Accordingly, this notice supersedes the previous notice in its 
entirety.
    By letter dated June 20, 2019, NEDA submitted a request for an 
amendment to the operating license (OL) and technical specifications 
(TSs) for the DAEC. The submittal requested revisions to the OL and TSs 
consistent with the permanent cessation of reactor operation and 
permanent defueling of the reactor. The revised TSs will be identified 
as the DAEC post defueled technical specifications (PDTS). Following 
the June 20, 2019, submittal, the licensee supplemented the original 
application by letters dated September 12, 2019, and November 4, 2019. 
NEDA performed an analysis of a fuel handling accident (FHA) in the 
spent fuel pool (SFP). This analysis determined that, following a decay 
period of 19 days, control building emergency ventilation is not 
required to maintain FHA dose consequences for control room occupants 
below the acceptance criteria of 10 CFR 50.67(b)(2)(iii). Consequently, 
NEDA hereby requests supplemental changes to the DAEC TSs to reflect 
the revised FHA analysis. Specifically, those TSs associated with 
control building emergency ventilation are proposed for deletion by 
this supplemental submittal.
    The proposed supplemental changes to the DAEC TSs are in accordance 
with 10 CFR 50.36(c)(1) through (c)(5). The proposed supplemental 
changes also include administrative changes to content format and 
revised page numbering. The TS Table of Contents will be revised 
accordingly.
    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 changes would not take effect until DAEC has 
certified to the NRC that it has permanently ceased operation and 
entered a permanently defueled condition. Because the 10 CFR part 50 
license for DAEC will no longer authorize operation of the reactor, 
or emplacement or retention of fuel into the reactor vessel with the 
certifications required by 10 CFR part 50.82(a)(1) submitted, as 
specified in 10 CFR part 50.82(a)(2), the occurrence of postulated 
accidents associated with reactor operation is no longer credible. 
DAEC's accident analyses are contained in Chapter 15 of the Updated 
Final Safety Analysis Report (UFSAR). In a permanently defueled 
condition, the only credible UFSAR described accident that remains 
is the Fuel Handling Accident (FHA). Other Chapter 15 accidents will 
no longer be applicable to a permanently defueled reactor.
    The UFSAR-described FHA analyses for DAEC shows that, following 
the required decay time after reactor shutdown and provided the SFP 
water level requirement of TS LCO [limiting condition for operation] 
3.7.8 is met, the dose consequences are acceptable without relying 
on secondary containment or the Standby Gas Treatment System. The 
control building envelop is credited for reduction of operator dose. 
Consequently, the TS requirements for the Standby Filter Unit and 
Control Building Chillers are retained.
    The probability of occurrence of previously evaluated accidents 
is not increased, since safe storage and handling of fuel will be 
the only operations performed, and therefore, bounded by the 
existing analyses. Additionally, the occurrence of postulated 
accidents associated with reactor operation will no longer be 
credible in the permanently defueled condition. This significantly 
reduces the scope of applicable accidents. The deletion of TS 
definitions and rules of usage and application requirements that 
will not be applicable in a defueled condition has no impact on 
facility SSCs [structures, system, and components] or the methods of 
operation of such SSCs. The deletion of design features and safety 
limits not applicable to the permanently shut down and defueled DAEC 
has no impact on the remaining applicable DBA [design-basis 
accident].
    The removal of LCOs or SRs [surveillance requirements] 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 do not affect the applicable DBAs previously evaluated 
since these DBAs are no longer applicable in the permanently 
defueled condition.
    The proposed changes, as supplemented, would not take effect 
until DAEC has certified to the NRC that it has permanently ceased 
operation, entered a permanently defueled condition, and a period of 
19 days has transpired since shutdown. Because the 10 CFR part 50 
license for DAEC will no longer authorize operation of the reactor, 
or emplacement or retention of fuel into the reactor vessel with the 
certifications required by 10 CFR part 50.82(a)(1) submitted, as 
specified in 10 CFR part 50.82(a)(2), the occurrence of postulated 
accidents associated with reactor operation is no longer credible. 
DAEC's accident analyses are contained in Chapter 15 of the Updated 
Final Safety Analysis Report (UFSAR). In a permanently defueled 
condition, the only credible UFSAR described accident that remains 
is the Fuel Handling Accident (FHA). Other Chapter 15 accidents will 
no longer be applicable to a permanently defueled reactor.
    The UFSAR-described FHA analyses for DAEC shows that, provided 
the SFP water level requirement of TS LCO 3.7.8 is met, the dose 
consequences are acceptable without relying on secondary containment 
or the Standby Gas Treatment System.
    Once the DAEC has permanently shut down and defueled, the only 
credible FHA is a fuel drop in the SFP. NEDA performed an analysis 
of the SFP FHA. This analysis determined that, following a decay 
period of 19 days, Control Building emergency ventilation is not 
required to maintain FHA dose consequences for control room 
occupants below the acceptance criteria of 10 CFR 50.67(b)(2)(iii). 
Consequently, the TS requirements for the systems supporting the 
Control Building emergency ventilation are proposed for deletion.
    The probability of occurrence of previously evaluated accidents 
is not increased, since safe storage and handling of fuel will be 
the only operations performed, and therefore, bounded by the 
existing analyses. Additionally, the occurrence of postulated 
accidents associated with reactor operation will no longer be 
credible in the permanently defueled condition. This significantly 
reduces the scope of applicable accidents. The deletion of TS 
definitions and rules of usage and application requirements that 
will not be applicable in a defueled condition has no impact on 
facility SSCs or the methods of operation of such SSCs. The deletion 
of design features and safety limits not applicable to the 
permanently shut down and defueled DAEC has no impact on the 
remaining applicable DBA.
    The removal of LCOs or SRs that are related only to the 
operation of the nuclear reactor or only to the prevention, 
diagnosis, or mitigation of reactor-related transients or

[[Page 66233]]

accidents do not affect the applicable DBAs previously evaluated 
since these DBAs are no longer applicable in the permanently 
defueled condition.
    Therefore, the proposed change, as supplemented, does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to delete or modify certain DAEC Operating 
License, TS, and current licensing bases (CLB) 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 the 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 shut down and defueled.
    The proposed modification or deletion of requirements of the 
DAEC Operating License, TS, and CLB do not affect systems credited 
in the accident analysis for the remaining credible DBA at DAEC. The 
proposed Operating License and PDTS will continue to require proper 
control and monitoring of safety significant parameters and 
activities. The TS regarding SFP water level and spent fuel storage 
is retained to preserve the current requirements for safe storage of 
irradiated fuel. The proposed amendment does not result in any new 
mechanisms that could initiate damage to the remaining relevant 
safety barriers for defueled plants (fuel cladding, spent fuel 
racks, SFP integrity, and SFP water level). Since extended operation 
in a defueled condition and safe fuel handling will be the only 
operation allowed, and therefore bounded by the existing analyses, 
such a condition does not create the possibility of a new or 
different kind of accident.
    The proposed changes, as supplemented, to delete or modify 
certain DAEC TS, and current licensing bases (CLB) 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 the 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 shut down and defueled.
    The proposed modification or deletion of requirements of the 
DAEC TS, and CLB do not affect systems credited in the accident 
analysis for the remaining credible DBA at DAEC. The proposed TS 
will continue to require proper control and monitoring of safety 
significant parameters and activities. The TS regarding SFP water 
level is retained to preserve the current requirements for safe 
storage of irradiated fuel. The proposed amendment, as supplemented, 
does not result in any new mechanisms that could initiate damage to 
the remaining relevant safety barriers for defueled plants (fuel 
cladding, spent fuel racks, SFP integrity, and SFP water level). 
Since extended operation in a defueled condition and safe fuel 
handling will be the only operation allowed, and therefore bounded 
by the existing analyses, such a condition does not create the 
possibility of a new or different kind of accident.
    Therefore, the proposed change, as supplemented, 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 changes are to delete or modify certain Operating 
License, TS and CLB once the DAEC facility has been permanently shut 
down and defueled. As specified in 10 CFR 50.82(a)(2), the 10 CFR 50 
license for DAEC will no longer authorize operation of the reactor 
or emplacement or retention of fuel into the reactor vessel 
following submittal of the certifications required by 10 CFR 
50.82(a)(1). As a result, the occurrence of certain design basis 
postulated accidents are no longer considered credible when the 
reactor is permanently defueled.
    The only remaining credible UFSAR described accident is a[n] 
FHA. The proposed changes do not adversely affect the inputs or 
assumptions of any of the design basis analyses that impact the FHA.
    The proposed changes are limited to those portions of the 
Operating License, TS, and CLB that are not related to the safe 
storage of irradiated fuel. The requirements proposed to be revised 
or deleted from the Operating License, TS, and CLB are not credited 
in the existing accident analysis for the remaining postulated 
accident (i.e., FHA); and, as such, do not contribute to the margin 
of safety associated with the accident analysis. Certain postulated 
DBAs involving the reactor are no longer possible because the 
reactor will be permanently shut down and defueled and DAEC will no 
longer be authorized to operate the reactor.
    The proposed changes, as supplemented, are to delete or modify 
certain TS and CLB once the DAEC facility has been permanently shut 
down and defueled and a period of no less than 19 days has 
transpired since shutdown. As specified in 10 CFR 50.82(a)(2), the 
10 CFR 50 license for DAEC will no longer authorize operation of the 
reactor or emplacement or retention of fuel into the reactor vessel 
following submittal of the certifications required by 10 CFR 
50.82(a)(1). As a result, the occurrence of certain design basis 
postulated accidents are no longer considered credible when the 
reactor is permanently defueled.
    The only remaining credible UFSAR described accident is a[n] 
FHA. Further, an FHA in the reactor core is no longer credible. An 
FHA in the SFP is the only remaining credible accident. NEDA has 
performed a revised analysis for an FHA in the SFP. This analysis 
determined that, following a decay period of 19 days, Control 
Building emergency ventilation is not required to maintain FHA dose 
consequences for control room occupants below the acceptance 
criteria of 10 CFR 50.67(b)(2)(iii). Consequently, TS LCOs and SRs 
associated with CBEV [Control Building emergency ventilation] and 
support equipment are proposed for deletion. The proposed changes, 
as supplemented, do not adversely affect the inputs or assumptions 
of the revised FHA analysis.
    The proposed changes, as supplemented, are limited to those 
portions of the TS, and CLB that are not related to the safe storage 
of irradiated fuel. The requirements proposed to be revised or 
deleted from the TS, and CLB are not credited in the existing 
accident analysis for the remaining postulated accident (i.e., FHA 
in the SFP); and, as such, do not contribute to the margin of safety 
associated with the accident analysis. Certain postulated DBAs 
involving the reactor are no longer possible because the reactor 
will be permanently shut down and defueled and DAEC will no longer 
be authorized to operate the reactor.
    Therefore, the proposed changes, as supplemented, have no impact 
to 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: Steven Hamrick, Managing Attorney--Nuclear, 
Florida Power Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Nancy L. Salgado.

NextEra Energy Duane Arnold (NEDA), LLC, Docket No. 50-331, Duane 
Arnold Energy Center (DAEC), Linn County, Iowa

    Date of amendment request: September 25, 2019, as supplemented by 
letter dated November 4, 2019. Publicly-available versions are in ADAMS 
under Accession Nos. ML19290G447, and ML19308A085, respectively.
    Description of amendment request: The amendment would delete the 
DAEC Operating License Condition 2.C.(3), ``Fire Protection Program,'' 
which requires that NEDA implement and maintain a fire protection 
program that complies with the requirements of 10 CFR 50.48(a) and 10 
CFR 50.48(c). NEDA will maintain a Fire Protection Program in 
accordance with 10 CFR 50.48(f), as required for licensees that have 
submitted certification of permanent cessation of operations.
    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

[[Page 66234]]

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 does not alter, degrade or prevent action 
described or assumed in any accident in the UFSAR [updated final 
safety analysis report] from being performed. The proposed change 
does not alter any assumptions previously made in evaluating 
radiological consequences. The proposed change does not affect the 
integrity of any fission product barrier.
    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 any safety limits or safety 
analysis assumptions associated with the operation of the plant. The 
proposed change does not introduce any new accident initiators, nor 
does the change reduce or adversely affect the capabilities of any 
plant structure or system in the performance of its safety function.
    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 does not alter the manner in which safety 
limits or limiting safety system settings are determined. The safety 
analysis acceptance criteria are not affected by the proposed 
change. The proposed change does not change the design function of 
any equipment assumed to operate in the event of an accident.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Steven Hamrick, Managing Attorney--Nuclear, 
Florida Power Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Nancy L. Salgado.
    Northern States Power Company--Minnesota (NSPM), Docket Nos. 50-282 
and 50-306, Prairie Island Nuclear Generating Plant (PINGP), Unit Nos.1 
and 2, Goodhue County, Minnesota
    Date of amendment request: October 7, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19280B335.
    Description of amendment request: The amendments would revise 
technical specifications (TSs) for the PINGP, Units 1 and 2. The 
proposed change revises TS 5.5.14, ``Containment Leakage Rate Testing 
Program,'' to increase the containment integrated leakage rate test 
program Type A test interval from 10 to 15 years and extend the 
containment isolation valve Type C leakage rate test frequency from 60 
to up to 75 months.
    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 adopts the NRC-accepted guidelines of NEI 
[Nuclear Energy Institute] 94-01 for the development of the NSPM 
performance-based containment testing program for PINGP Units 1 and 
2. NEI 94-01 allows, based on risk and performance, an extension of 
the Type A and Type C containment leak test intervals. 
Implementation of these guidelines continues to provide adequate 
assurance that during design basis accidents, the primary 
containment and its components will limit leakage rates to less than 
the values assumed in the plant safety analyses.
    The findings of the PINGP risk assessment confirm the general 
findings of previous studies that the risk impact with extending the 
containment leak rate is small. In accordance with the guidance 
provided in Regulatory Guide 1.174, ``An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' an extension of the leak 
test interval in accordance with NEI 94-01, Revision 3-A results in 
an estimated change within the very small change region.
    Since the change is implementing a performance-based containment 
testing program, the proposed amendment does not involve either a 
physical change to the plant or a change in the manner in which the 
plant is operated or controlled. The requirement for containment 
leakage rate acceptance will not be changed by this amendment. 
Therefore, the containment will continue to perform its design 
function as a barrier to fission product releases.
    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 implement a performance-based containment 
testing program, associated with integrated leakage rate test 
frequency, does not change the design or operation of structures, 
systems, or components of the plant. The proposed change would 
continue to ensure containment integrity and would ensure operation 
within the bounds of existing accident analyses. There are no 
accident initiators created or affected by this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to confidence in the ability of the 
fission product barriers (fuel cladding, reactor coolant system, and 
primary containment) to perform their design functions during and 
following postulated accidents. The proposed change to implement a 
performance-based containment testing program, associated with 
integrated leakage rate test and local leak rate testing frequency, 
does not affect plant operations, design functions, or any analysis 
that verifies the capability of a structure, system, or component of 
the plant to perform a design function. In addition, this change 
does not affect safety limits, limiting safety system setpoints, or 
limiting conditions for operation.
    The specific requirements and conditions of the TS Containment 
Leakage Rate Testing Program exist to ensure that the degree of 
containment structural integrity and leak-tightness that is 
considered in the plant safety analysis is maintained. The overall 
containment leak rate limit specified by the TSs is maintained. This 
ensures that the margin of safety in the plant safety analysis is 
maintained. The design, operation, testing methods and acceptance 
criteria for Type A, B, and C containment leakage tests specified in 
applicable codes and standards would continue to be met with the 
acceptance of this proposed change since these are not affected by 
implementation of a performance-based containment testing program.
    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: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Nancy L. Salgado.

Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, 
Georgia

    Date of amendment request: September 30, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19273A953.

[[Page 66235]]

    Description of amendment request: The amendment request proposes 
changes to the Combined License (COL) Numbers NPF-91 and NPF-92 for 
VEGP, Units 3 and 4, and proposes to depart from Updated Final Safety 
Analysis Report (UFSAR) Tier 2 information (which includes the plant-
specific Design Control Document (DCD) Tier 2 information). The 
proposed changes involve related changes to plant-specific Tier 1 
information, with corresponding changes to the associated COL Appendix 
C information, and involves related changes to COL Appendix A, 
Technical Specifications. Specifically, the requested amendment 
proposes changes to reflect revisions in the design parameters of (a) 
the maximum stroke times for the automatic depressurization system 
(ADS) Stages 1, 2 and 3 valves, (b) the minimum effective flow areas 
for the ADS Stages 2 and 3 valves, and (c) the core makeup tank minimum 
volume. 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.
    The proposed revisions to the automatic depressurization system 
(ADS) and core makeup tank (CMT) design parameters have been found 
to continue to provide the required functional capability of the 
safety systems for previously evaluated accidents and anticipated 
operational occurrences. The ADS and CMT design parameters are not 
an initiator of any accident analyzed in the Updated Final Safety 
Analysis Report (UFSAR), nor do the changes 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 UFSAR are not affected. The proposed 
changes do not involve a change to any mitigation sequence or the 
predicted radiological releases due to postulated accident 
conditions, thus, the consequences of the accidents evaluated in the 
UFSAR are not affected.
    The UFSAR describes the analyses of various design basis 
transients and accidents to demonstrate compliance of the design 
with the acceptance criteria for these events. The acceptance 
criteria for the various events are based on meeting the relevant 
regulations, general design criteria, and the Standard Review Plan, 
and are a function of the anticipated frequency of occurrence of the 
event and potential radiological consequences to the public. The 
revised accident analyses maintain their plant conditions, and thus 
their frequency designation and consequence level as 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 proposed revisions to the ADS and CMT design parameters have 
been found to continue to provide the required functional capability 
of the safety systems for previously evaluated accidents and 
anticipated operational occurrences. The proposed revisions to the 
ADS and CMT design parameters do not change the function of the 
related systems, and thus, the changes do not introduce a new 
failure mode, malfunction or sequence of events that could adversely 
affect safety or safety-related equipment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed revisions to the ADS and CMT design parameters have 
been found to continue to provide the required functional capability 
of the safety systems for previously evaluated accidents and 
anticipated operational occurrences. The proposed revisions to the 
ADS and CMT design parameters does not change the function of the 
related systems nor significantly affect the margins provided by the 
systems. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the requested changes.
    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: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Victor Hall.

IV. Previously Published Notices of Consideration of Issuance of 
Amendments to Facility Operating Licenses and Combined Licenses, 
Proposed No Significant Hazards Consideration Determination, and 
Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar 
Nuclear Plant, Units 1 and 2, Rhea County, Tennessee

    Date of amendment request: October 23, 2019. A publicly-available 
version is in ADAMS under Accession No. ML19296C538.
    Description of amendment request: The amendments would revise the 
Watts Bar Nuclear Plant, Units 1 and 2, Technical Specification Table 
3.3.5-1, ``LOP [Loss of Power] DG [Diesel Generator] Start 
Instrumentation,'' Function 5, ``6.9 kV [kilovolt] Emergency Bus 
Undervoltage (Unbalanced Voltage),'' to correct the values for the 
allowable value for the unbalanced voltage relay (UVR) low trip 
voltage, the allowable value for the UVR high trip time delay, and the 
trip setpoint for the UVR high trip time delay.
    Date of publication of individual notice in Federal Register: 
November 6, 2019 (84 FR 59846).
    Expiration date of individual notice: December 6, 2019 (public 
comments); January 6, 2020 (hearing requests).

V. 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.

[[Page 66236]]

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.

DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: February 8, 2019.
    Brief description of amendment: The amendment adopted Technical 
Specifications Task Force (TSTF)-564, ``Safety Limit MCPR (Minimum 
Critical Power Ratio),'' Revision 2, and revises the Fermi 2 technical 
safety limit on MCPR to reduce the need for cycle-specific changes to 
the value while still meeting the regulatory requirement for a safety 
limit. In addition, TS 5.6.5, Core Operating Limits Report (COLR), was 
revised to require the current safety limit MCPR value to be included 
in the cycle specific COLR.
    Date of issuance: November 5, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 214. A publicly-available version is in ADAMS under 
Accession No. ML19189A004; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-43: The amendment 
revised the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 9, 2019 (84 FR 
14144).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 5, 2019.
    No significant hazards consideration comments received: No.

NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit 
No. 1, Rockingham County, New Hampshire

    Date of amendment request: October 4, 2018, as supplemented by 
letter dated September 30, 2019.
    Description of amendment request: The amendment revised the 
technical specifications to adopt changes provided in Technical 
Specifications Task Force (TSTF)-234, ``Add Action for More than One 
(Digital Rod Position Indication) [D]RPI Inoperable''; TSTF-547, 
``Clarification of Rod Position Requirements''; and made various other 
changes to align the Seabrook TSs more closely with NUREG-1431, 
``Standard Technical Specifications Westinghouse Plants.''
    Date of issuance: November 18, 2019.
    Effective date: As of its date of issuance and shall be implemented 
by May 28, 2020.
    Amendment No.: 162. A publicly-available version is in ADAMS under 
Accession No. ML19224A563; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-86: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: April 9, 2019 (84 FR 
14151). The supplemental letter dated September 30, 2019, 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 November 18, 2019.
    No significant hazards consideration comments received: No.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, 
Goodhue County, Minnesota

    Date of amendment request: July 20, 2018, as supplemented by 
letters dated April 29, 2019 and August 5, 2019.
    Brief description of amendment: The amendments added a condition to 
the PINGP, Units 1 and 2, renewed facility operating licenses to allow 
the implementation of 10 CFR 50.69, ``Risk informed categorization and 
treatment of structures, systems and components for nuclear power 
reactors.''
    Date of issuance: November 12, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 230 (Unit 1); 218 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML19276F684; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-42 and DPR-60: The 
amendments revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: September 11, 2018 (83 
FR 45986). The supplemental letters dated April 29, 2019 and August 5, 
2019, 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 November 12, 2019.
    No significant hazards consideration comments received: No.

Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota

    Date of amendment request: October 2, 2018, as supplemented by 
letter dated December 4, 2018.
    Brief description of amendment: The amendments revised the design 
basis accident dose threshold for designation of certain fuel handling 
equipment as Quality Type I (safety-related) to greater than 10 percent 
of the dose limits specified in 10 CFR part 100, ``Reactor Site 
Criteria.''
    Date of issuance: November 7, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 229 (Unit 1); 217 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML19232A151; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.

[[Page 66237]]

    Renewed Facility Operating License Nos. DPR-42 and DPR-60: The 
amendments revised the Updated Safety Analysis Report.
    Date of initial notice in Federal Register: January 31, 2019 (84 FR 
812).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 7, 2019.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: February 27, 2019.
    Brief description of amendment: The amendment adopted Technical 
Specifications Task Force (TSTF) Traveler TSTF-546, ``Revise APRM 
[Average Power Range Monitor] Channel Adjustment Surveillance 
Requirement,'' which revises the Hope Creek Generating Station 
technical specification surveillance requirement to verify that 
calculated power is no more than 2 percent greater than the APRM 
channel output. This change revised the surveillance requirement to 
distinguish between APRM indications that are consistent with the 
accident analyses and those that provide additional margin.
    Date of issuance: November 7, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 220. A publicly-available version is in ADAMS under 
Accession No. ML19289A886; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-57: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: April 9, 2019 (84 FR 
14152).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 7, 2019.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC and Exelon Generation Company, LLC, Docket Nos. 50-272 
and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey

    Date of amendment request: February 4, 2019, as supplemented by 
letter dated June 11, 2019.
    Brief description of amendments: The amendments revised the 
Technical Specification requirements on control and shutdown rods and 
rod and bank position indication, consistent with NRC-approved 
Technical Specifications Task Force (TSTF) Traveler TSTF-547, Revision 
1, ``Clarification of Rod Position Requirements,'' dated March 4, 2016.
    Date of issuance: November 18, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 330 (Unit No. 1) and 311 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML19275D694; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-70 and DPR-75: The 
amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: March 26, 2019 (84 FR 
11339). The supplemental letter dated June 11, 2019, 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 November 18, 2019.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia

    Date of amendment request: July 23, 2019.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) actions for inoperable residual heat 
removal (RHR) shutdown cooling subsystems in the RHR shutdown cooling 
system limiting conditions for operation. The proposed changes are 
based on Technical Specifications Task Force (TSTF) traveler TSTF-566, 
Revision 0, ``Revise Actions for Inoperable RHR Shutdown Cooling 
Subsystems,'' dated January 19, 2018.
    Date of issuance: November 13, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 300 (Unit No. 1) and 245 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML19267A023; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: The 
amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: September 10, 2019 (84 
FR 47551).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 13, 2019.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: February 1, 2019.
    Brief description of amendments: The amendments adopted Technical 
Specifications Task Force (TSTF) Traveler TSTF-563, Revision 0, 
``Revise Instrument Testing Definitions to Incorporate the Surveillance 
Frequency Control Program.''
    Date of issuance: November 18, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 347 (Unit 1) and 341 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML19281B554; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-77 and DPR-79: The 
amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: April 9, 2019 (84 FR 
14153).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 18, 2019.
    No significant hazards consideration comments received: No.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: November 19, 2018, as supplemented by 
letter dated August 22, 2019.
    Brief description of amendments: The amendments approved 
installation of

[[Page 66238]]

two non-safety-related water headers within a safety-related flood 
protection dike.
    Date of issuance: November 13, 2019.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 283 (Unit No. 1) and 266 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML19274C998; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Facility Operating License No. NPF-4 and NPF-7: The amendments 
revised the Renewed Facility Operating Licenses and Technical 
Specifications.
    Date of initial notice in Federal Register: March 26, 2019 (84 FR 
11342). The supplement dated August 22, 2019, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 13, 2019.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 25th day of November 2019.

    For the Nuclear Regulatory Commission.
Craig G. Erlanger,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2019-25972 Filed 12-2-19; 8:45 am]
 BILLING CODE 7590-01-P