[Federal Register Volume 82, Number 58 (Tuesday, March 28, 2017)]
[Notices]
[Pages 15377-15392]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-05990]


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NUCLEAR REGULATORY COMMISSION

[NRC-2017-0080]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, 
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from February 28, 2017 to March 13, 2017. The 
last biweekly notice was published on March 14, 2017.

DATES: Comments must be filed by April 27, 2017. A request for a 
hearing must be filed by May 30, 2017.

[[Page 15378]]


ADDRESSES: You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0080. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Paula Blechman, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-2242; email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2017-0080, 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 Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0080.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [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 the 
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-2017-0080, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

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

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

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if 
appropriate, a notice of a hearing will be issued.
    As required by 10 CFR 2.309(d) the petition should specifically 
explain the reasons why intervention should be permitted with 
particular reference to the following general requirements for 
standing: (1) The name, address, and telephone number of the 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
petitioner's property, financial, or other interest in the proceeding; 
and (4) the possible effect of any decision or order which may be 
entered in the proceeding on the petitioner's interest.
    In accordance with 10 CFR 2.309(f), the petition must also set 
forth the specific contentions which the

[[Page 15379]]

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 by May 
30, 2017. The petition must be filed in accordance with the filing 
instructions in the ``Electronic Submissions (E-Filing)'' section of 
this document, and should meet the requirements for petitions set forth 
in this section, except that under 10 CFR 2.309(h)(2) a State, local 
governmental body, or federally recognized Indian Tribe, or agency 
thereof does not need to address the standing requirements in 10 CFR 
2.309(d) if the facility is located within its boundaries. 
Alternatively, a State, local governmental body, Federally-recognized 
Indian Tribe, or agency thereof may participate as a non-party under 10 
CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
the discretion of the presiding officer, be permitted to make a limited 
appearance pursuant to the provisions of 10 CFR 2.315(a). A person 
making a limited appearance may make an oral or written statement of 
his or her position on the issues but may not otherwise participate in 
the proceeding. A limited appearance may be made at any session of the 
hearing or at any prehearing conference, subject to the limits and 
conditions as may be imposed by the presiding officer. Details 
regarding the opportunity to make a limited appearance will be provided 
by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate under 10 CFR 2.315(c), must be filed in accordance with the 
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 
46562, August 3, 2012). The E-Filing process requires participants to 
submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC's Web site at 
http://www.nrc.gov/site-help/e-submittals.html. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [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 Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a 
digital ID certificate and a docket has been created, the participant 
can then submit adjudicatory documents. Submissions must be in Portable 
Document Format (PDF). Additional guidance on PDF submissions is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the 
time the document is submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who

[[Page 15380]]

have advised the Office of the Secretary that they wish to participate 
in the proceeding, so that the filer need not serve the document on 
those participants separately. Therefore, applicants and other 
participants (or their counsel or representative) must apply for and 
receive a digital ID certificate before adjudicatory documents are 
filed so that they can obtain access to the documents via the E-Filing 
system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public Web site at http://www.nrc.gov/site-help/e-submittals.html, by 
email to [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 information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station (PNPS), Plymouth County, Massachusetts
    Date of amendment request: February 14, 2017. A publicly available 
version is in ADAMS under Accession No. ML17053A468.
    Description of amendment request: The amendment would revise 
certain staffing and training requirements, reports, programs, and 
editorial changes in the Technical Specifications (TS) Table of 
Contents; Section 1.0, ``Definitions''; Section 4.0, ``Design 
Features''; and Section 5.0, ``Administrative Controls'' that will no 
longer be applicable once PNPS is permanently defueled.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would not take effect until PNPS has 
permanently ceased operation and entered a permanently defueled 
condition and the Certified Fuel Handler Training and Retraining 
Program is approved by the NRC. The proposed amendment would modify 
the PNPS TS by deleting the portions of the TS that are no longer 
applicable to a permanently defueled facility, while modifying the 
other sections to correspond to the permanently defueled condition.
    The deletion and modification of provisions of the 
administrative controls do not directly affect the design of 
structures, systems, and components (SSCs) necessary for safe 
storage of irradiated fuel or the methods used for handling and 
storage of such fuel in the spent fuel pool. The changes to the 
administrative controls are administrative in nature and do not 
affect any accidents applicable to the safe management of irradiated 
fuel or the permanently shutdown and defueled condition of the 
reactor. Thus, the consequences of an accident previously evaluated 
are not increased.
    In a permanently defueled condition, the only credible accidents 
are the fuel handling accident (FHA) and those involving radioactive 
waste systems remaining in service. The probability of occurrence of 
previously evaluated accidents is not increased, because extended 
operation in a defueled condition will be the only operation 
allowed. This mode of operation is bounded by the existing analyses. 
Additionally, the occurrence of postulated accidents associated with 
reactor operation is no longer credible in a permanently defueled 
reactor. This significantly reduces the scope of applicable 
accidents.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes have no impact on facility SSCs affecting 
the safe storage of irradiated fuel, or on the methods of operation 
of such SSCs, or on the handling and storage of irradiated fuel 
itself. The administrative removal or modifications of the TS that 
are related only to administration of the facility cannot result in 
different or more adverse failure modes or accidents than previously 
evaluated because the reactor will be permanently shutdown and 
defueled and PNPS will no longer be authorized to operate the 
reactor or retain or place fuel in the reactor vessel.
    The proposed changes to the PNPS TS do not affect systems 
credited in the accident analysis for the FHA or radioactive waste 
system upsets at PNPS. The proposed TS will continue to require 
proper control and monitoring of safety significant parameters and 
activities.
    The proposed amendment does not result in any new mechanisms 
that could initiate damage to the remaining relevant safety barriers 
for defueled plants (fuel cladding and spent fuel cooling). Extended 
operation in a defueled condition will be the only

[[Page 15381]]

operation allowed, and it is 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 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.
    Since the 10 CFR part 50 license for PNPS will no longer 
authorize operation of the reactor or emplacement or retention of 
fuel into the reactor vessel once the certifications required by 10 
CFR 50.82(a)(1) are docketed, as specified in 10 CFR 50.82(a)(2), 
the occurrence of postulated accidents associated with reactor 
operation is no longer credible. The only remaining credible 
accidents are a FHA and those involving radioactive waste systems 
remaining in service. The proposed amendment does not adversely 
affect the inputs or assumptions of any of the design basis analyses 
that impact these analyzed conditions.
    The proposed changes are limited to those portions of the TS 
that are not related to the safe storage of irradiated fuel. The 
requirements that are proposed to be revised or deleted from the 
PNPS TS are not credited in the existing accident analysis for the 
remaining applicable postulated accident; and as such, do not 
contribute to the margin of safety associated with the accident 
analysis. Postulated design basis accidents involving the reactor 
are no longer possible because the reactor will be permanently 
shutdown and defueled and PNPS will no longer be authorized to 
operate the reactor or retain or place fuel in the reactor vessel.
    Therefore, the proposed change does not involve a significant 
reduction in 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2 (NMP2), Oswego County, New York
    Date of amendment request: December 13, 2016, as supplemented by 
letter dated February 17, 2017. Publicly-available versions are in 
ADAMS under Accession Nos. ML16348A368 and ML17048A034, respectively.
    Description of amendment request: The amendment would revise the 
NMP2 technical specification (TS) safety limit (SL) to increase the low 
pressure isolation setpoint allowable value, which will result in 
earlier main steam line isolation. The revised main steam line low 
pressure isolation capability and the revised SL are intended to ensure 
that NMP2 remains within the TS SLs in the event of a pressure 
regulator failure maximum demand transient.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC staff edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because decreasing the reactor dome pressure in TS SL 2.1.1.1 and TS 
SL 2.1.1.2 for reactor RTP [rated thermal power] ranges and 
increasing the AV [allowable value] for the Main Steam Line 
Pressure-Low on TS Table 3.3.6.1-1, Function b, effectively expands 
the range of applicability for GEXL correlation and the calculation 
of MCPR [minimum critical power ratio]. The CPR [critical power 
ratio] rises during the pressure reduction following the scram that 
terminates the PRFO [pressure regulator failure--maximum demand 
(open)] transient. The reduction in the reactor dome pressure value 
in the SL from 785 psig [pounds per square inch gauge] to 700 psia 
[pounds per square inch absolute] and the increase in the AV from 
>=746 psig to >=814 psig adequately accommodate the pressure 
reduction during the PRFO transient within the revised TS limit 
without compromising fuel integrity.
    The expanded GEXL correlation range supports NMP2 revised low 
pressure safety limit of 700 psia. The proposed TS revision involves 
no significant changes to the operation of any systems or components 
in normal or accident or transient operating conditions.
    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 create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed reduction in the reactor dome pressure value in 
the SL from 785 psig to 700 psia reflects a wider range of 
applicability for the GEXL correlation which is approved by the NRC 
for both GE14 currently in NMP2 and GNF2 fuels proposed for NMP2. 
The proposed changes do not involve physical changes to the plant or 
its operating characteristics. In addition, the increase in the AV 
for the MSL [main steam line] low pressure from >=746 psig to >=814 
psig will result in the MSIV [main steam isolation valve] closure 
signal initiation at a higher temperature. As a result, no new 
failure modes are being introduced.
    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 do not involve a significant reduction in a 
margin of safety because 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 the 
actuation of equipment relied upon to respond to transients and 
design basis accidents. The proposed change in reactor dome pressure 
SLs and the AV for the MSL low pressure ensures the safety margin is 
maintained, which protects the fuel cladding integrity during steady 
state operation, normal operational transients, or AOOs [anticipated 
operational occurrences] such as 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 proposed changes do not involve physical changes to the 
plant or its operating characteristics. The reduction in the reactor 
dome pressure value in the SL from 785 psig to 700 psia and the 
increase to the AV for the MSL low pressure provides added margin to 
accommodate the pressure reduction during the PRFO transient within 
the revised TS limit without compromising fuel integrity.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Acting Branch Chief: Stephen S. Koenick.
Exelon Generation Company, LLC (Exelon), Docket No. 50-219, Oyster 
Creek Nuclear Generating Station (OCNGS), Ocean County, New Jersey
    Date of amendment request: February 20, 2017. A publicly-available 
version is available in ADAMS under Accession No. ML17051A003.
    Description of amendment request: The licensee proposes to delete 
from the Facility Operating License (FOL) certain license conditions, 
which impose specific requirements on the decommissioning trust 
agreement. The

[[Page 15382]]

licensee proposes to meet the provisions of 10 CFR 50.75(h) for OCNGS.
    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 requested changes delete License Conditions 3.F through 3.K 
pertaining to Decommissioning Trust Agreements currently in the 
OCNGS FOL. The requested changes are consistent with the types of 
license amendments [identified] in 10 CFR 50.75(h)(4).
    The regulations of 10 CFR 50.75(h)(4) state ``Unless otherwise 
determined by the Commission with regard to a specific application, 
the Commission has determined that any amendment to the license of a 
utilization facility that does no more than delete specific license 
conditions relating to the terms and conditions of decommissioning 
trust agreements involves ``no significant hazard considerations.''
    This request involves changes that are administrative in nature. 
No actual plant equipment or accident analyses will be affected by 
the proposed changes.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Does the [p]roposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This request involves administrative changes to the license that 
will be consistent with the NRC's regulations at 10 CFR 50.75(h).
    No actual plant equipment or accident analyses will be affected 
by the proposed change and no failure modes not bounded by 
previously evaluated accidents will be created.
    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.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers to limit the level of radiation dose to 
the public.
    This request involves administrative changes to the license that 
will be consistent with the NRC's regulations at 10 CFR 50.75(h).
    No actual plant equipment or accident analyses will be affected 
by the proposed change. Additionally, the proposed changes will not 
relax any criteria used to establish safety limits, will not relax 
any safety systems settings, or will not relax the bases for any 
limiting conditions of operation.
    Therefore, the proposed 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 
and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and 
Lancaster Counties, Pennsylvania
    Date of amendment request: January 30, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17030A302.
    Description of amendment request: The amendments would replace 
existing Technical Specification (TS) requirements related to 
``operations with a potential for draining the reactor vessel'' 
(OPDRVs) with new requirements on reactor pressure vessel (RPV) water 
inventory control (WIC) to protect Safety Limit 2.1.1.3. Safety Limit 
2.1.1.3 requires RPV water level to be greater than the top of active 
irradiated fuel. The proposed changes are based on TS Task Force (TSTF) 
Traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water 
Inventory Control.''
    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 replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold 
shutdown) and Mode 5 (i.e., refueling) is not an accident previously 
evaluated and, therefore, replacing the existing TS controls to 
prevent or mitigate such an event with a new set of controls has no 
effect on any accident previously evaluated. RPV water inventory 
control in Mode 4 or Mode 5 is not an initiator of any accident 
previously evaluated. The existing OPDRV controls or the proposed 
RPV WIC controls are not mitigating actions assumed in any accident 
previously evaluated.
    The proposed changes reduce the probability of an unexpected 
draining event (which is not a previously evaluated accident) by 
imposing new requirements on the limiting time in which an 
unexpected draining event could result in the reactor vessel water 
level dropping to the top of the active fuel (TAF). These controls 
require cognizance of the plant configuration and control of 
configurations with unacceptably short drain times. These 
requirements reduce the probability of an unexpected draining event. 
The current TS requirements are only mitigating actions and impose 
no requirements that reduce the probability of an unexpected 
draining event.
    The proposed changes reduce the consequences of an unexpected 
draining event (which is not a previously evaluated accident) by 
requiring an Emergency Core Cooling System (ECCS) subsystem to be 
operable at all times in Modes 4 and 5. The current TS requirements 
do not require any water injection systems, ECCS or otherwise, to be 
Operable in certain conditions in Mode 5. The change in requirement 
from two ECCS subsystems to one ECCS subsystem in Modes 4 and 5 does 
not significantly affect the consequences of an unexpected draining 
event because the proposed Actions ensure equipment is available 
within the limiting drain time that is as capable of mitigating the 
event as the current requirements. The proposed controls provide 
escalating compensatory measures to be established as calculated 
drain times decrease, such as verification of a second method of 
water injection and additional confirmations that containment and/or 
filtration would be available if needed.
    The proposed changes reduce or eliminate some requirements that 
were determined to be unnecessary to manage the consequences of an 
unexpected draining event, such as automatic initiation of an ECCS 
subsystem and control room ventilation. These changes do not affect 
the consequences of any accident previously evaluated since a 
draining event in Modes 4 and 5 is not a previously evaluated 
accident and the requirements are not needed to adequately respond 
to a draining event.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC that will protect Safety 
Limit 2.1.1.3. The proposed changes will not alter the design 
function of the equipment involved. Under the proposed changes, some 
systems that are currently required to be operable during OPDRVs 
would be required to be available within the limiting drain time or 
to be in service depending on the limiting drain time. Should those 
systems be unable to be placed into service, the consequences are no 
different than if those systems were unable to perform their 
function under the current TS

[[Page 15383]]

requirements. The event of concern under the current requirements 
and the proposed changes are an unexpected draining event. The 
proposed changes do not create new failure mechanisms, malfunctions, 
or accident initiators that would cause a draining event or a new or 
different kind of accident not previously evaluated or included in 
the design and licensing bases.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes replace existing TS requirements related to 
OPDRVs with new requirements on RPV WIC. The current requirements do 
not have a stated safety basis and no margin of safety is 
established in the licensing basis. The safety basis for the new 
requirements is to protect Safety Limit 2.1.1.3. New requirements 
are added to determine the limiting time in which the RPV water 
inventory could drain to the top of the fuel in the reactor vessel 
should an unexpected draining event occur. Plant configurations that 
could result in lowering the RPV water level to the TAF within one 
hour are now prohibited. New escalating compensatory measures based 
on the limiting drain time replace the current controls. The 
proposed TS establish a safety margin by providing defense-in-depth 
to ensure that the Safety Limit is protected and to protect the 
public health and safety. While some less restrictive requirements 
are proposed for plant configurations with long calculated drain 
times, the overall effect of the change is to improve plant safety 
and to add safety margin.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL 
60555.
    NRC Branch Chief: James G. Danna.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
    Date of amendment request: January 23, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17025A399.
    Description of amendment request: The amendments would modify the 
St. Lucie Plant, Unit Nos. 1 and 2, Technical Specifications (TSs) by 
limiting the MODE of applicability for the Reactor Protection System 
(RPS), Startup, and Operating Rate of Change of Power--High, functional 
unit trip. Additionally, the proposed license amendments add new 
Limiting Condition for Operation (LCO) 3.0.5 and relatedly modifies LCO 
3.0.2, to provide for placing inoperable equipment under administrative 
control for the purpose of conducting testing required to demonstrate 
OPERABILITY.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Limiting the MODE 1 applicability for RPS functional unit, 
Startup and Operating Rate of Change of Power--High, to Power Range 
Neutron Flux Power <=15% of RATED THERMAL POWER, is an 
administrative change in nature and does not alter the manner in 
which the functional unit is operated or maintained. The proposed 
changes do not represent any physical change to plant [structures, 
systems, and components (SSC(s))], or to procedures established for 
plant operation. The subject RPS functional unit is not an event 
initiator nor is it credited in the mitigation of any event or 
credited in the [probabilistic risk assessment (PRA)]. As such, the 
initial conditions associated with accidents previously evaluated 
and plant systems credited for mitigating the consequences of 
accidents previously evaluated remain unchanged.
    The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 
and Unit 2 TS and related modification to LCO 3.0.2 is consistent 
with the guidance provided in NUREG-1432, Volume 1 [ADAMS Accession 
No. ML12102A165] (Reference 6.1 [of the amendment request]) and 
thereby has been previously evaluated by the Commission with a 
determination that the proposed change does not involve a 
significant hazards consideration.
    Therefore, facility operation in accordance with the proposed 
license amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Limiting the MODE 1 applicability for the RPS functional unit, 
Startup and Operating Rate of Change of Power--High, to Power Range 
Neutron Flux Power <= 5% of RATED THERMAL POWER, is an 
administrative change in nature and does not involve the addition of 
any plant equipment, methodology or analyses. The proposed changes 
do not alter the design, configuration, or method of operation of 
the subject RPS functional unit or of any other SSC. More 
specifically, the proposed changes neither alter the power rate-of-
change trip function nor its ability to bypass and reset as 
required. The subject RPS functional unit remains capable of 
performing its design function.
    The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 
and Unit 2 TS and related modification to LCO 3.0.2 is consistent 
with the guidance provided in NUREG-1432, Volume 1 (Reference 6.1 
[of the amendment request]) and thereby has been previously 
evaluated by the Commission with a determination that the proposed 
change does not involve a significant hazards consideration.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Limiting the MODE 1 applicability for RPS functional unit, 
Startup and Operating Rate of Change of Power--High, to Power Range 
Neutron Flux Power <=15% of RATED THERMAL POWER is an administrative 
change in nature. The proposed changes neither involve changes to 
any safety analyses assumptions, safety limits, or limiting safety 
system settings nor do they adversely impact plant operating margins 
or the reliability of equipment credited in safety analyses.
    The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1 
and Unit 2 TS and related modification to LCO 3.0.2 is consistent 
with the guidance provided in NUREG-1432, Volume 1 (Reference 6.1 
[of the amendment request]) and thereby has been previously 
evaluated by the Commission with a determination that the proposed 
change does not involve a significant hazards consideration.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Benjamin G. Beasley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1 (FCS), Washington County, Nebraska
    Date of amendment request: December 16, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16351A464.

[[Page 15384]]

    Description of amendment request: The proposed amendment would 
revise the FCS Emergency Plan and Emergency Action Level (EAL) scheme 
for the permanently defueled condition. The proposed permanently 
defueled Emergency Plan and EAL scheme are commensurate with the 
significantly reduced spectrum of credible accidents that can occur in 
the permanently defueled condition and are necessary to properly 
reflect the conditions of the facility while continuing to preserve the 
effectiveness of the emergency plan.
    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 to the FCS Emergency Plan and EAL scheme do 
not impact the function of facility structures, systems, or 
components. The proposed changes do not affect accident initiators 
or precursors, nor does it alter design assumptions. The proposed 
changes do not prevent the ability of the on-shift staff and 
emergency response organization to perform their intended functions 
to mitigate the consequences of any accident or event that will be 
credible in the permanently defueled condition.
    The probability of occurrence of previously evaluated accidents 
is not increased, because most previously analyzed accidents can no 
longer occur and the probability of the few remaining credible 
accidents are unaffected by the proposed amendment.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes reduce the scope of the FCS Emergency Plan 
and EAL scheme commensurate with the hazards associated with a 
permanently shutdown and defueled facility. The proposed changes do 
not involve installation of new equipment or modification of 
existing equipment, so that no new equipment failure modes are 
introduced. Also, the proposed changes do not result in a change to 
the way that the equipment or facility is operated resulting in new 
or different kinds of accident initiators or accident mitigation.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed changes are 
associated with the FCS Emergency Plan and EAL scheme and do not 
impact operation of the facility or its response to transients or 
accidents. The change does not affect the Technical Specifications. 
The proposed changes do not involve a change in the method of 
facility operation, and no accident analyses will be affected by the 
proposed changes. Safety analysis acceptance criteria are not 
affected by the proposed changes. The revised Emergency Plan will 
continue to provide the necessary response staff.
    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: David A. Repka, Esq., Winston & Strawn, 1700 
K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Douglas A. Broaddus.
PSEG Nuclear LLC, Docket Nos. 50-354, 50-272, and 50-311, Hope Creek 
Generating Station (HCGS) and Salem Nuclear Generating Station (SGS), 
Unit Nos. 1 and 2, Salem County, New Jersey
    Date of amendment request: February 13, 2017. A publicly-available 
version is in ADAMS under Package Accession No. ML17044A346.
    Description of amendment request: The amendments would revise the 
HCGS and SGS, Unit Nos. 1 and 2, emergency action level (EAL) schemes. 
Specifically, the licensee proposes to adopt the EAL scheme described 
in Nuclear Energy Institute (NEI) 99-01, Revision 6, ``Development of 
Emergency Action Levels for Non-Passive Reactors.'' NEI 99-01, Revision 
6, has been endorsed by the NRC.
    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 to the HCGS and SGS EALs do not impact the 
physical function of plant structures, systems or components (SSC) 
or the manner in which SSCs perform their design function. The 
proposed changes neither adversely affect accident initiators or 
precursors, nor alter design assumptions. The proposed changes do 
not alter or prevent the ability of SSCs to perform their intended 
function to mitigate the consequences of an initiating event within 
assumed acceptance limits. No operating procedures or administrative 
controls that function to prevent or mitigate accidents are affected 
by the proposed changes. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration of the 
plant (i.e., no new or different types of equipment will be 
installed or removed) or a change in the method of plant operation. 
The proposed changes will not introduce failure modes that could 
result in a new accident, and the changes do not alter assumptions 
made in the safety analysis. The proposed changes to the HCGS and 
SGS EALs are not initiators of any accidents. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is associated with the ability of the fission 
product barriers (i.e., fuel cladding, reactor coolant system 
pressure boundary, and containment structure) to limit the level of 
radiation dose to the public. The proposed changes do not impact 
operation of the plant or its response to transients or accidents. 
The changes do not affect the Technical Specifications or the 
operating license. The proposed changes do not involve a change in 
the method of plant operation, and no accident analyses will be 
affected by the proposed changes. Additionally, the proposed changes 
will not relax any criteria used to establish safety limits and will 
not relax any safety system settings. The safety analysis acceptance 
criteria are not affected by these changes. The proposed changes 
will not result in plant operation in a configuration outside the 
design basis. The proposed changes do not adversely affect systems 
that respond to safely shut down the plant and to maintain the plant 
in a safe shutdown condition. The emergency plan will continue to 
activate an emergency response commensurate with the extent of 
degradation of plant 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 15385]]

    Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: James G. Danna.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, 
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield, South 
Carolina
    Date of amendment request: February 15, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17046A660.
    Description of amendment request: The amendment request proposes to 
revise the licensing basis information to reflect changes to the 
locations of the hydrogen venting primary openings in the passive core 
cooling system (PXS) valve/accumulator rooms inside containment. 
Because this proposed change requires a departure from Tier 1 
information in the Westinghouse Electric Company's AP1000 Design 
Control Document (DCD), the licensee also requested an exemption from 
the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 
52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed revision to the hydrogen venting for the Passive 
Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and 
clarification of the venting path definition for PXS Valve/
Accumulator Room B (Room 11207) do not affect any safety-related 
equipment or function. The hydrogen ignition subsystem, including 
designed hydrogen venting features, is designed to mitigate beyond 
design basis hydrogen generation in the containment. The hydrogen 
venting changes do not involve any accident, initiating event or 
component failure; thus, the probabilities of the accidents 
previously evaluated are not affected. The modified venting 
locations and definitions will maintain the hydrogen ignition 
subsystem designed and analyzed beyond design basis function to 
maintain containment integrity. The maximum allowable containment 
leakage rate specified in the Technical Specifications is unchanged, 
and radiological material release source terms are not affected; 
thus, the radiological releases in the accident analyses are not 
affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed revision to the hydrogen venting for the Passive 
Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and 
clarification of the venting path definition for PXS Valve/
Accumulator Room B (Room 11207) will maintain the beyond design 
basis function of the hydrogen ignition subsystem. The hydrogen 
venting changes do not impact the hydrogen ignition subsystem's 
function to maintain containment integrity during beyond design 
basis accident conditions, and, thus does not introduce any new 
failure mode. The proposed changes do not create a new fault or 
sequence of events that could result in a radioactive release. The 
proposed changes would not affect any safety-related accident 
mitigating function.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed revision to the hydrogen venting for the Passive 
Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and 
clarification of the venting path definition for PXS Valve/
Accumulator Room B (Room 11207) will maintain the beyond design 
basis function of the hydrogen ignition subsystem. The proposed 
changes do not have any effect on the ability of safety-related 
structures, systems, or components to perform their beyond design 
basis functions. The proposed changes are a result of a low 
probability, severe accident scenario being evaluated. The revision 
to this scenario does not result in an increase in the plant risk 
(frequency and/or consequences). The frequency is low and there is 
no increase to the consequences because containment integrity is 
maintained and there is no containment leakage. There is no change 
to the maximum allowed containment leakage rate (0.10% of 
containment air weight per day) for the containment vessel. The 
proposed changes do not affect the ability of the hydrogen igniter 
subsystem to maintain containment integrity following a beyond 
design basis accident. The hydrogen igniter subsystem continues to 
meet the requirements for which it was designed and continues to 
meet the regulations.
    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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius, LLC, 1111 Pennsylvania NW., Washington, DC 20004-2514.
    NRC Branch Chief: Jennifer Dixon-Herrity.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, 
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, 
South Carolina
    Date of amendment request: February 16, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17047A192.
    Description of amendment request: The requested amendment proposes 
to depart from Tier 2 information in the Updated Final Safety Analysis 
Report (UFSAR) and involves changes to related plant-specific Tier 1 
information, with corresponding changes to the associated combined 
license (COL) Appendix C information, to clarify text that currently 
refers to raceways with an electrical classification (i.e., Class 1E/
non-Class 1E). This includes rewording multiple Inspections, Tests, 
Analyses, and Acceptance Criteria (ITAAC) and UFSAR material to clarify 
that any text referring to Class 1E or non-Class 1E raceways or raceway 
systems is referring to raceways or raceway systems that route Class 1E 
or non-Class 1E circuits.
    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.
    These proposed changes are for clarification and consistency. No 
structure, system, or component (SSC) or function is changed within 
this activity. There is no change to the application of regulatory 
guides or industry standards to raceways or raceway systems, nor is 
there a change to how they are designed, fabricated, procured or 
installed. Raceway systems that route Class 1E circuits will 
continue to be designated and designed as equipment Class C, safety-
related, and seismic Category I structures. The proposal to align 
the text in COL Appendix C (and plant-specific Tier 1) Section 3.3 
with the associated ITAAC is made for clarification and consistency 
to reduce misinterpretation. The proposal to reword multiple ITAAC 
in 3.3.00.07 does not change the intent of the ITAAC, nor is the 
ITAAC scope or closure method impacted.
    The proposed amendment does not affect the prevention and 
mitigation of abnormal events; e.g., accidents, anticipated 
operation occurrences, earthquakes, floods, turbine missiles, and 
fires or their safety or design analyses. This change does not 
involve containment of radioactive isotopes or any adverse effect on 
a fission product barrier. There is no impact on previously 
evaluated accidents.
    Therefore, the proposed amendment does not involve a significant 
increase in the

[[Page 15386]]

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 involve a new failure mechanism or 
malfunction, which affects an SSC accident initiator, or interface 
with any SSC accident initiator or initiating sequence of events 
considered in the design and licensing bases. There is no adverse 
effect on radioisotope barriers or the release of radioactive 
materials. The proposed amendment does not adversely affect any 
accident, including the possibility of creating a new or different 
kind of accident from any accident previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different type of accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    These proposed changes are for clarification and consistency to 
reduce misinterpretation. No SSC or function is changed within this 
activity. There is no change to the application of regulatory guides 
or industry standards to raceways or raceway systems, nor is there a 
change to how they are designed, fabricated, procured or installed. 
Raceway systems that route Class 1E circuits will continue to be 
designated and designed as Equipment Class C, safety-related, and 
seismic Category I.
    The proposed changes would not affect any safety-related design 
code, function, design analysis, safety analysis input or result, or 
existing design/safety margin. No safety analysis or design basis 
acceptance limit/criterion is challenged or exceeded by the 
requested changes.
    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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC, 20004-2514.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke 
County, Georgia
    Date of amendment request: August 30, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16243A373.
    Description of amendment request: The amendment request proposes a 
change to Updated Final Safety Analysis Report in the form of 
departures from the incorporated plant-specific Design Control Document 
(DCD) Tier 2 * information and related changes to the VEGP Units 3 and 
4 Combined License (COL) Appendix C (and corresponding plant-specific 
DCD Tier 1) information.
    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, 
a design certification rule is also requested for the plant-specific 
Tier 1 material departures. The proposed change is to the thickness of 
one floor in the auxiliary building located between Column Lines I to 
J-1 and Column Lines 2 to 4 at Elevation 153'-0''. This submittal 
requests approval of the license amendment, necessary to implement 
these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below with NRC staff edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design functions of the nuclear island structures are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the nuclear island. 
The nuclear island structures are structurally designed to meet 
seismic Category I requirements as defined in Regulatory Guide 1.29.
    The change of the thickness of the floor above the [Component 
Cooling Water System (CCS)] Valve Room in the auxiliary building 
meets criteria and requirements of American Concrete Institute (ACI) 
349 and American Institute of Steel Construction (AISC) N690 and 
does not have an adverse impact on the response of the nuclear 
island structures safe shutdown earthquake ground motions or loads 
due to anticipated transient or postulated accident conditions. The 
proposed changes do not impact the support, design, or operation of 
mechanical and fluid systems. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to the predicted radioactive releases due to normal 
operation or postulated accident conditions. The plant response to 
previously evaluated accidents or external events is not adversely 
affected, nor does the change described create any new accident 
precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change is to revise the thickness of the floor 
above the CCS Valve Room in the auxiliary building. The proposed 
changes do not change the design requirements of the nuclear island 
structures. The proposed changes do not change the design function, 
support, design, or operation of mechanical and fluid systems. The 
proposed changes do not result in a new failure mechanism for the 
nuclear island structures or new accident precursors. As a result, 
the design function of the nuclear island structures is not 
adversely affected by the proposed change.
    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.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes, thus, no margin of 
safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: January 31, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17031A446.
    Description of amendment request: The requested amendment proposes 
to depart from Tier 2 information in the Updated Final Safety Analysis 
Report (UFSAR) and to change Combined License Appendix A, Technical 
Specifications (TS), to modify engineered safety features logic for 
containment vacuum relief actuation.
    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?

[[Page 15387]]

    Response: No.
    The proposed changes to the UFSAR and TS will include the 
Containment Pressure--Low automatic reset function for the 
containment vacuum relief valves manual initiation logic, such that 
the containment vacuum relief manual actuation will be automatically 
reset when the containment pressure rises above the Containment 
Pressure--Low setpoint. This reset allows a containment isolation 
signal to close the valves when necessary. The Containment 
Pressure--Low signal is an interlock for the containment vacuum 
relief manual actuation such that the valves cannot be opened unless 
the Containment Pressure--Low setpoint has been reached in any two-
out-of-four divisions. The modified logic will ensure that the 
automatic initiation of containment isolation is made available 
following manual initiation of containment vacuum relief actuation. 
The analyzed design and function of the Engineered Safety Features 
Actuation System and its actuated components is not affected. The 
proposed changes do not adversely affect any safety-related 
equipment and does not involve any accident, initiating event, or 
component failure, thus the probabilities of accidents previously 
evaluated are not affected. The proposed changes do not adversely 
interface with or adversely affect any system containing 
radioactivity or affect any radiological material release source 
term; thus the radiological releases in an accident are not 
affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The changes to the UFSAR and TS to include the Containment 
Pressure--Low manual actuation interlock and automatic reset 
function for the containment vacuum relief valves manual initiation 
logic will maintain the Engineered Safety Features Actuation System 
and Plant Safety and Monitoring System in accordance with the design 
objectives as licensed. The design of the Class 1E Containment 
Pressure--Low manual actuation interlock and automatic reset 
function is required to meet the licensing basis for the Engineered 
Safety Features Actuation System and Plant Safety and Monitoring 
System. The changes to the manual initiation logic do not adversely 
affect the function of any safety-related structure, system, or 
component, and thus does not introduce a new failure mode. The 
changes to the containment vacuum relief valves manual initiation 
logic do not adversely interface with any safety-related equipment 
or any equipment associated with radioactive material and, thus, do 
not create a new fault or sequence of events that could result in a 
new or different kind of accident.
    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 changes to the UFSAR and TS to include the Containment 
Pressure--Low automatic reset function for the containment vacuum 
relief valves manual initiation logic will maintain the Engineered 
Safety Features Actuation System and Plant Safety and Monitoring 
System in accordance with the design objectives as licensed. The 
changes to the manual initiation logic do not adversely interface 
with any safety-related equipment or adversely affect any safety-
related function. The changes to the containment vacuum relief 
manual initiation logic continue to comply with existing design 
codes and regulatory criteria, and do not involve a significant 
reduction in the margin of safety.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: March 2, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17061A747.
    Description of amendment request: The requested amendment consist 
of changes to Inspections, Tests, Analyses, and Acceptance Criteria 
(ITAAC) in combined license (COL) Appendix C, with corresponding 
changes to the associated plant-specific Tier 1 information, to 
consolidate a number of ITAAC to improve efficiency of the ITAAC 
completion and closure process.
    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 
Design Control Document 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 non-technical change to COL Appendix C will 
consolidate, relocate and subsume redundant ITAAC in order to 
improve and create a more efficient process for the ITAAC Closure 
Notification submittals. No structure, system, or component (SSC) 
design or function is affected. No design or safety analysis is 
affected. The proposed changes do not affect any accident initiating 
event or component failure, thus the probabilities of the accidents 
previously evaluated are not affected. No function used to mitigate 
a radioactive material release and no radioactive material release 
source term is involved, thus the radiological releases in the 
accident analyses are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to COL Appendix C does not affect the design 
or function of any SSC, but will consolidate, relocate and subsume 
redundant ITAAC in order to improve efficiency of the ITAAC 
completion and closure process. The proposed changes would not 
introduce a new failure mode, fault or sequence of events that could 
result in a radioactive material release.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to COL Appendix C to consolidate, relocate 
and subsume redundant ITAAC in order to improve efficiency of the 
ITAAC completion and closure process is considered non-technical and 
would not affect any design parameter, function or analysis. There 
would be no change to an existing design basis, design function, 
regulatory criterion, or analysis. No safety analysis or design 
basis acceptance limit/criterion is involved.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.

[[Page 15388]]

Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: February 22, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17053A425.
    Description of amendment request: The amendment request proposes to 
revise the licensing basis information to reflect changes to the 
locations of the hydrogen venting primary openings in the passive core 
cooling system (PXS) valve/accumulator rooms inside containment. 
Because, this proposed change requires a departure from Tier 1 
information in the Westinghouse Electric Company's AP1000 Design 
Control Document (DCD), the licensee also requested an exemption from 
the requirements of the Generic DCD Tier 1 in accordance with 10 CFR 
52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed revision to the hydrogen venting for the Passive 
Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and 
clarification of the venting path definition for PXS Valve/
Accumulator Room B (Room 11207) do not affect any safety-related 
equipment or function. The hydrogen ignition subsystem, including 
designed hydrogen venting features, is designed to mitigate beyond 
design basis hydrogen generation in the containment. The hydrogen 
venting changes do not involve any accident, initiating event or 
component failure; thus, the probabilities of the accidents 
previously evaluated are not affected. The modified venting 
locations and definitions will maintain the hydrogen ignition 
subsystem designed and analyzed beyond design basis function to 
maintain containment integrity. The maximum allowable containment 
leakage rate specified in the Technical Specifications is unchanged, 
and radiological material release source terms are not affected; 
thus, the radiological releases in the accident analyses are not 
affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed revision to the hydrogen venting for the PXS Valve/
Accumulator Room A (Room 11206) and clarification of the venting 
path definition for PXS Valve/Accumulator Room B (Room 11207) will 
maintain the beyond design basis function of the hydrogen ignition 
subsystem. The hydrogen venting changes do not impact the hydrogen 
ignition subsystem's function to maintain containment integrity 
during beyond design basis accident conditions, and, thus does not 
introduce any new failure mode. The proposed changes do not create a 
new fault or sequence of events that could result in a radioactive 
release. The proposed changes would not affect any safety-related 
accident mitigating function.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed revision to the hydrogen venting for the Passive 
Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and 
clarification of the venting path definition for PXS Valve/
Accumulator Room B (Room 11207) will maintain the beyond design 
basis function of the hydrogen ignition subsystem. The proposed 
changes do not have any effect on the ability of safety-related 
structures, systems, or components to perform their beyond design 
basis functions. The proposed changes are a result of a low 
probability, severe accident scenario being evaluated. The revision 
to this scenario does not result in an increase in the plant risk 
(frequency and/or consequences). The frequency is low and there is 
no increase to the consequences because containment integrity is 
maintained and there is no containment leakage. There is no change 
to the maximum allowed containment leakage rate (0.10% of 
containment air weight per day) for the containment vessel. The 
proposed changes do not affect the ability of the hydrogen igniter 
subsystem to maintain containment integrity following a beyond 
design basis accident. The hydrogen igniter subsystem continues to 
meet the requirements for which it was designed and continues to 
meet the regulations.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Tennessee Valley Authority (TVA), Docket No. 50-391, Watts Bar Nuclear 
Plant (WBN), Unit 2, Rhea County, Tennessee
    Date of amendment request: December 21, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16356A673.
    Description of amendment request: The amendment would revise the 
containment ice mass limits in WBN, Unit 2, Technical Specification 
(TS) Surveillance Requirements (SRs) 3.6.11.2 and 3.6.11.3 to be 
identical to the ice mass limits in the WBN, Unit 1, TS SRs 3.6.11.2 
and 3.6.11.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The primary purpose of the ice bed is to provide a large heat 
sink to limit peak containment pressure in the event of a release of 
energy from a design basis LOCA [loss-of-coolant accident] or high 
energy line break (HELB) in containment. The LOCA requires the 
greatest amount of ice compared to other accident scenarios; 
therefore, the reduction in ice weight is based on the LOCA 
analysis. The amount of ice in the bed has no impact on the 
initiation of an accident, but rather on the mitigation of the 
accident. The containment integrity analysis shows that the proposed 
reduced ice weight is sufficient to maintain the peak containment 
pressure below the containment design pressure, and that the 
containment heat removal systems function to rapidly reduce the 
containment pressure and temperature in the event of a LOCA. 
Therefore, containment integrity is maintained and the consequences 
of an accident previously evaluated in the WBN dual-unit Updated 
Final Safety Analysis Report (UFSAR) are not significantly 
increased. 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 ice condenser serves to limit the peak pressure inside 
containment following a LOCA. TVA has evaluated the revised 
containment pressure analysis and determined that sufficient ice 
would be present to maintain the peak containment pressure below the 
containment design pressure. Therefore, the reduced ice weight does 
not create the possibility of an accident that is different than any 
already evaluated in the WBN dual-unit UFSAR. No new accident 
scenarios, failure mechanisms, or limiting single failures are 
introduced as a result of this proposed change.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed TS ice weight SR limit is based on the conservatism 
of the WBN Unit

[[Page 15389]]

1 WCOBRA/TRAC LOCA M&E [mass and energy] methodology in comparison 
to the WBN Unit 2 operating conditions. The WBN Unit 1 WCOBRA/TRAC 
LOCA M&E methodology is modeled on the WBN Unit 1 RSGs [replacement 
steam generators], which have a greater mass, volume, and stored 
metal energy than the WBN Unit 2 original model D3 SGs [steam 
generators]. Additionally, the containment pressure calculations in 
Section 6.2.1.3.3 of the WBN Unit 1 portion of the WBN dual-unit 
UFSAR state that the analytical limit for the mass of ice assumed in 
the WBN Unit 1 ice condenser, in order to limit the maximum 
containment peak pressure from a LOCA to below the containment 
design pressure, is 2,260,000 lb. The proposed revised TS SR ice 
mass limit of 2,404,500 lb [pound] includes additional ice mass to 
conservatively bound ice bed sublimation effects. Based on TVA's 
evaluation and the revised containment analysis, TVA considers the 
reduction of the ice mass limit to be acceptable for satisfying the 
safety function of the ice condenser for the current SR interval. 
Therefore, the proposed 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Branch Chief: Benjamin G. Beasley.

III. 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 No. 50-391 Watts Bar Nuclear Plant, 
Unit 2, Rhea County, Tennessee
    Date of amendment request: November 14, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16320A161.
    Brief description of amendment request: The proposed amendment 
would revise the Watts Bar Nuclear Plant, Unit 2, Cyber Security Plan 
Implementation Schedule for Milestone 8 and would revise the associated 
license condition in the Facility Operating License.
    Date of publication of individual notice in Federal Register: 
January 5, 2017 (82 FR 1370).
    Expiration date of individual notice: February 6, 2017 (public 
comments); March 6, 2017 (hearing requests).

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan
    Date of amendment request: March 22, 2016, as supplemented by 
letter dated August 11, 2016.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 5.5.12, ``Primary Containment Leakage Rate Testing 
Program,'' for the permanent extension of the Type A test interval up 
to one test in 15 years, as stipulated in Nuclear Energy Institute 
(NEI) 94-01, Revision 2-A, ``Industry Guideline for Implementing 
Performance-Based Option of 10 CFR part 50, Appendix J,'' October 2008 
(ADAMS Accession No. ML100620847). The license amendment request also 
proposes to increase the containment isolation valves leakage test 
intervals (i.e., Type C tests) from their current 60 months to 75 
months by replacing TS 5.5.12.a. reference to Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Test Program'' (ADAMS Accession 
No. ML003740058), with a reference to NEI 94-01, Revision 3-A (ADAMS 
Accession No. ML12221A202), and the conditions and limitations 
specified in NEI 94-01, Revision 2-A, to implement the performance-
based leakage testing program in accordance with title 10 of the Code 
of Federal Regulations part 50, Appendix J, Option B. The amendment 
also deletes from TS 5.5.12, text that authorized a one-time extension 
of the Type A test interval to 2007 and revised paragraph 2.D of the 
renewed facility operating license to reflect removal of a reference to 
an exemption from 10 CFR part 50, Appendix J, requirements for testing 
of containment air locks.
    Date of issuance: March 9, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 205. A publicly-available version is in ADAMS under 
Accession No. ML16351A460; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-43: Amendment revised 
the renewed facility operating license and TSs.
    Date of initial notice in Federal Register: June 7, 2016 (81 FR 
36616). The August 11, 2016 supplement provided additional information 
that clarified the application, did not expand the scope of the 
application as originally

[[Page 15390]]

noticed, and did not change the staff's original proposed no 
significant hazard consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 2017.
    No significant hazards consideration comments received: No.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick 
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina
    Date of amendment request: September 26, 2016.
    Brief description of amendments: The amendments revised Technical 
Specification Section 2.1.1.2 to change the minimum critical power 
ratio safety limit.
    Date of issuance: March 10, 2017.
    Effective date: As of date of issuance and shall be implemented for 
Unit 1 prior to start-up from the 2018 refueling outage (March 2018) 
and for Unit 2 prior to start-up from the 2017 refueling outage.
    Amendment Nos.: 272 (Unit 1) and 300 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17059D146; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-71 and DPR-62: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: December 20, 2016 (81 
FR 92866).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 10, 2017.
    No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
    Date of application for amendments: May 5, 2016, as supplemented by 
letter dated June 16, 2016.
    Brief description of amendments: The amendments would modify the 
McGuire Nuclear Station, Units 1 and 2, Technical Specifications (TS) 
by removing footnote (c) from TS Table 3.3.2-1, ``Engineered Safety 
Feature Actuation System Instrumentation,'' which is no longer 
applicable, and by removing an expired footnote from TS 3.8.1, ``AC 
Sources--Operating.''
    Date of issuance: March 8, 2017.
    Effective date: As of its date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 293 and 272. A publicly-available version is in 
ADAMS under Accession No. ML17003A019; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the licenses and technical specifications.
    Date of initial notice in Federal Register: July 5, 2016 (81 FR 
43649). The supplemental letter dated June 16, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 8, 2017.
    No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of application for amendment: May 10, 2016, as supplemented by 
letters dated May 18, 2016, and January 31, 2017.
    Brief description of amendment: The amendment revised the safety 
function lift and lower setpoint tolerances of the safety/relief valves 
that are listed in Surveillance Requirements 3.4.3.1 and 3.4.4.1 of the 
Technical Specifications.
    Date of issuance: March 9, 2017.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 240. A publicly-available version is in ADAMS under 
Accession No. ML17052A125; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-21: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: July 19, 2016 (81 FR 
46961). The supplemental letter January 31, 2017, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 2017.
    No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
1, Pope County, Arkansas
    Date of amendment request: March 25, 2016.
    Brief description of amendment: The amendment deleted Technical 
Specification (TS) 5.5.8, ``Inservice Testing Program.'' A new defined 
term, ``Inservice Testing Program,'' is added to TS Section 1.1, 
``Definitions.'' Also, existing uses of the term ``Inservice Testing 
Program'' in the TSs are capitalized throughout to indicate that it is 
now a defined term. The NRC staff has concluded that the amendment is 
consistent with Technical Specifications Task Force Traveler TSTF-545, 
Revision 3, which was made available to the TSTF via NRC letter dated 
December 11, 2015 (ADAMS Accession No. ML15317A071).
    Date of issuance: March 10, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 257. A publicly-available version is in ADAMS under 
Accession No. ML16165A423; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-51: The amendment 
revised the Renewed Facility Operating License and Technical 
Specifications.
    Date of initial notice in Federal Register: June 7, 2016 (81 FR 
36619).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 10, 2017.
    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
    Date of amendment request: April 4, 2016.
    Brief description of amendments: The amendments revised the 
technical specification (TS) requirements for the high pressure coolant 
injection (HPCI) and reactor core isolation cooling (RCIC) system 
actuation instrumentation. Specifically, the amendments add a footnote 
to the TSs indicating that the injection functions of drywell pressure-
high (HPCI only) and manual initiation (HPCI and RCIC) are not required 
to be operable under low reactor pressure conditions.
    Date of issuance: February 28, 2017.

[[Page 15391]]

    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 224 (Unit 1) and 185 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16356A272; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-39 and NPF-85: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: June 7, 2016 (81 FR 
36620).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 28, 2017.
    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-220, Nine Mile Point 
Nuclear Station, Unit 1, (NMP1), Oswego County, New York
    Date of amendment request: January 3, 2017.
    Brief description of amendment: The amendment revised the NMP1 
licensing basis related to alternative source term analysis in the 
updated final safety analysis report (UFSAR) to allow the use of the 
release fractions listed in Tables 1 and 3 of NRC Regulatory Guide 
1.183, ``Alternative Radiological Source Terms for Evaluating Design 
Basis Accidents at Nuclear Power Reactors,'' July 2000 (ADAMS Accession 
No. ML003716792), for partial length fuel rods (PLRs) that are 
operating above the peak burnup limit for the remainder of the current 
operating cycle. In addition, the proposed change revised the NMP1 
licensing basis to allow movement of irradiated fuel bundles containing 
PLRs that have been in operation above 62,000 megawatt days per metric 
tons of uranium (MWD/MTU).
    Date of issuance: March 9, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 226. A publicly-available version is in ADAMS under 
Accession No. ML17055A451; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-63: Amendment revised 
the licensing basis related to alternative source term analysis in the 
UFSAR.
    Date of initial notice in Federal Register: January 31, 2017 (82 FR 
8871).
    The Commission's related evaluation of the amendment and final no 
significant hazards consideration determination are contained in a 
Safety Evaluation dated March 9, 2017.
    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey
    Date amendment request: May 17, 2016, as supplemented by letters 
dated November 2, 2016, and March 1, 2017.
    Brief description of amendment: The amendment revised and removed 
certain requirements from the Section 6, ``Administrative Controls,'' 
portions of the Oyster Creek Nuclear Generating Station Technical 
Specifications (TSs) that are not applicable to the facility in a 
permanently defueled condition. In addition, the amendment added 
definitions to TS Section 1, ``Definitions.'' Also, the amendment made 
additions to, deletions from, and conforming administrative changes to 
the TSs.
    Date of issuance: March 7, 2017.
    Effective date: Effective upon the licensee's submittal of the 
certifications required by 10 CFR 50.82(a)(1)(i) and 50.82(a)(1)(ii), 
and shall be implemented within 60 days of the effective date of the 
amendment, but may not exceed March 29, 2020.
    Amendment No.: 290. A publicly-available version is in ADAMS under 
Accession No. ML16235A413; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-16: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: July 19, 2016 (81 FR 
46963). On July 19, 2016, the NRC staff published a proposed no 
significant hazards consideration (NSHC) determination regarding the 
amendment request in the Federal Register (81 FR 46963). Subsequently, 
by letter dated November 2, 2016, the licensee provided additional 
information that expanded the scope of the amendment request as 
originally noticed in the Federal Register. Accordingly, the NRC staff 
published a second proposed NSHC determination regarding the amendment 
request in the Federal Register on November 22, 2016 (81 FR 83876), 
which superseded the original Federal Register notice in its entirety. 
The supplemental letter dated March 1, 2017, provided additional 
information that clarified the application, did not expand the scope of 
the application as noticed, and did not change the NRC staff's second 
proposed NSHC determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2017.
    No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 50-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: June 16, 2016.
    Brief description of amendments: The amendments changed Combined 
License Nos. NPF-91 and NPF-92 for the Vogtle Electric Generating Plant 
Units 3 and 4. The amendments authorized changes to the Updated Final 
Safety Analysis Report (UFSAR) in the form of departures from the 
incorporated plant-specific Design Control Document Tier 2 information. 
Specifically, the changes to the Technical Specifications (TS) and 
information in the UFSAR revised the AP1000 protection and safety 
monitoring system functional logic to comply with the requirements on 
operating bypasses in Clause 6.6, ``Operating Bypasses'' of the 
Institute of Electrical and Electronics Engineers (IEEE) Std. 603-1991, 
``IEEE Standard Criteria for Safety Systems for Nuclear Power 
Generating Stations.''
    Date of issuance: February 24, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 71/70. A publicly-available version is in ADAMS 
under Accession No. ML16320A097; documents related to these amendments 
are listed in the Safety Evaluation enclosed with the amendments.
    Facility Operating License Nos. NPF-91 and NPF-92: Amendment 
revised the Facility Combined License and TS.
    Date of initial notice in Federal Register: August 16, 2016 (81 FR 
54610).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 24, 2017.
    No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1 (VCSNS), Fairfield County, South Carolina
    Date of amendment request: June 30, 2016, as supplemented by letter 
dated August 4, 2016.
    Brief description of amendment: This amendment revised the date of 
the

[[Page 15392]]

Cyber Security Plan implementation schedule for Milestone 8. Milestone 
8 requires full implementation of the VCSNS Cyber Security Plan.
    Date of issuance: March 9, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 208. A publicly-available version is in ADAMS under 
Accession No. ML17011A050; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-12: Amendment revised 
the Renewed Facility Operating License.
    Date of initial notice in Federal Register: October 4, 2016 (81 FR 
68472).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 2017.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 16th day of March 2017.

    For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2017-05990 Filed 3-27-17; 8:45 am]
 BILLING CODE 7590-01-P