Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 31089-31106 [2017-13804]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 82, Number 127 (Wednesday, July 5, 2017)]
[Notices]
[Pages 31089-31106]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-13804]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[NRC-2017-0152]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

-----------------------------------------------------------------------

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 June 3, 2017 to June 
19, 2017. The last biweekly notice was published on June 19, 2017.

DATES: Comments must be filed by August 4, 2017. A request for a 
hearing must be filed by September 5, 2017.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0152. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. 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: TWFN-8-D36M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Kay Goldstein, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-1506, email: Kay.Goldstein@nrc.gov.

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2017-0152, 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 https://www.regulations.gov and search for Docket ID NRC-2017-0152.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``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 pdr.resource@nrc.gov. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in this 
document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One

[[Page 31090]]

White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2017-0152, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

[[Page 31091]]

after issuance of the amendment. If the final determination is that the 
amendment request involves a significant hazards consideration, then 
any hearing held would take place before the issuance of the amendment 
unless the Commission finds an imminent danger to the health or safety 
of the public, in which case it will issue an appropriate order or rule 
under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission no later 
than 60 days from the date of publication of this notice. The petition 
must be filed in accordance with the filing instructions in the 
``Electronic Submissions (E-Filing)'' section of this document, and 
should meet the requirements for petitions set forth in this section, 
except that under 10 CFR 2.309(h)(2) a State, local governmental body, 
or federally recognized Indian Tribe, or agency thereof does not need 
to address the standing requirements in 10 CFR 2.309(d) if the facility 
is located within its boundaries. Alternatively, a State, local 
governmental body, Federally-recognized Indian Tribe, or agency thereof 
may participate as a non-party under 10 CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
the discretion of the presiding officer, be permitted to make a limited 
appearance pursuant to the provisions of 10 CFR 2.315(a). A person 
making a limited appearance may make an oral or written statement of 
his or her position on the issues but may not otherwise participate in 
the proceeding. A limited appearance may be made at any session of the 
hearing or at any prehearing conference, subject to the limits and 
conditions as may be imposed by the presiding officer. Details 
regarding the opportunity to make a limited appearance will be provided 
by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

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

[[Page 31092]]

pursuant to an order of the Commission or the presiding officer. If you 
do not have an NRC-issued digital ID certificate as described above, 
click cancel when the link requests certificates and you will be 
automatically directed to the NRC's electronic hearing dockets where 
you will be able to access any publicly available documents in a 
particular hearing docket. Participants are requested not to include 
personal privacy information, such as social security numbers, home 
addresses, or personal phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. For 
example, in some instances, individuals provide home addresses in order 
to demonstrate proximity to a facility or site. With respect to 
copyrighted works, except for limited excerpts that serve the purpose 
of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

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

    Date of amendment request: April 3, 2017, as supplemented by 
letters dated April 3, 2017, and May 2, 2017. Publicly-available 
versions are in ADAMS under Accession Nos. ML17093A787, ML17093A796, 
and ML17122A223, respectively.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to extend the required 
frequency of certain 18-month Surveillance Requirements (SRs) to 24 
months to accommodate a 24-month refueling cycle. In addition, the 
proposed amendment would revise certain programs in TS Section 5.5, 
``Programs and Manuals,'' to change 18-month frequencies to 24 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment changes the surveillance frequency from 
18 months to 24 months for SRs in the TSs that are normally a 
function of the refueling interval. Duke Energy Progress, LLC's 
evaluations have shown that the reliability of protective 
instrumentation and equipment will be preserved for the maximum 
allowable surveillance interval.
    The proposed change does not involve any change to the design or 
functional requirements of the associated systems. That is, the 
proposed TS change neither degrades the performance of, nor 
increases the challenges to any safety systems assumed to function 
in the plant safety analysis. The proposed change will not give rise 
to any increase in operation power level, fuel operating limits or 
effluents. The proposed change does not affect any accident 
precursors since no accidents previously evaluated relate to the 
frequency of surveillance testing and the revision to the frequency 
does not introduce any accident initiators. The proposed change does 
not impact the usefulness of the SRs in evaluating the operability 
of required systems and components or the manner in which the 
surveillances are performed.
    In addition, evaluation of the proposed TS change demonstrates 
that the availability of equipment and systems required to prevent 
or mitigate the radiological consequences of an accident is not 
significantly affected because of the availability of redundant 
systems and equipment or the high reliability of the equipment. 
Since the impact on the systems is minimal, it is concluded that the 
overall impact on the plant safety analysis is negligible.
    Furthermore, an historical review of surveillance test results 
and associated maintenance records indicates there is no evidence of 
any failure that would invalidate the above conclusions. Therefore, 
the proposed TS change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment does not require a change to the plant 
design nor the mode of plant operation. No new or different 
equipment is being installed. No installed equipment is being 
operated in a different manner. As a result, no new failure modes 
are being introduced. In addition, the proposed change does not 
impact the usefulness of the SRs in evaluating the operability of 
required systems and components or the manner in which the 
surveillances are performed. Furthermore, an historical review of 
surveillance test results and associated maintenance records 
indicates there is no evidence of any failure that would invalidate 
the above conclusions. Therefore, the implementation of the proposed 
change will not create the possibility for an accident of a new or 
different type than previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment changes the surveillance frequency from 
18 months to 24 months for SRs in the TSs that are normally a 
function of the refueling interval. SR 3.0.2 would allow a maximum 
surveillance interval of 30 months for these surveillances. Although 
the proposed change will result in an increase in the interval 
between surveillance tests, the impact on system availability is 
small based on other, more frequent testing that is performed, the 
existence of redundant systems and equipment or overall system 
reliability. There is no evidence of any time-dependent failures 
that would impact the availability of the systems. The proposed 
change does not significantly impact the condition or performance of 
structures, systems and components relied upon for accident 
mitigation. This change does not alter the existing TS allowable 
values or analytical limits. The existing operating margin between 
plant conditions and actual plant setpoints is not significantly 
reduced due to these changes. The assumptions and results in any 
safety analyses are not significantly impacted. Therefore, the 
proposed change does not involve a significant reduction in margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A, 
Charlotte, NC 28202.
    NRC Branch Chief: Undine S. Shoop.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: April 24, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17114A398.
    Description of amendment request: The amendment would revise 
Technical Specification requirements regarding steam generator tube 
inspections and reporting as described in Technical Specification Task 
Force (TSTF) Traveler TSTF-510, Revision 2, ``Revision to Steam 
Generator Program Inspection Frequencies and Tube Sample Selection,'' 
using the Consolidated Line Item Improvement Process for Arkansas 
Nuclear One, Unit No. 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

[[Page 31093]]

consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG 
tube sample selection. A steam generator tube rupture (SGTR) event 
is one of the design basis accidents that are analyzed as part of a 
plant's licensing basis. The proposed SG tube inspection frequency 
and sample selection criteria will continue to ensure that the SG 
tubes are inspected such that the probability of a[n] SGTR is not 
increased. The consequences of a[n] SGTR are bounded by the 
conservative assumptions in the design basis accident analysis. The 
proposed change will not cause the consequences of a[n] SGTR to 
exceed those assumptions.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the SG Program will not introduce any 
adverse changes to the plant design basis or postulated accidents 
resulting from potential tube degradation. The proposed change does 
not affect the design of the SGs or their method of operation. In 
addition, the proposed change does not impact any other plant system 
or component.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a[n] 
SG is maintained by ensuring the integrity of its tubes.
    SG tube integrity is a function of the design, environment, and 
the physical condition of the tube. The proposed change does not 
affect tube design or operating environment. The proposed change 
will continue to require monitoring of the physical condition of the 
SG tubes such that there will not be a reduction in the margin of 
safety compared to the current requirements.
    Therefore, it is concluded that this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, 
Entergy Services, Inc., 101 Constitution Avenue NW., Suite 200 East, 
Washington, DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.

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

    Date of amendment request: April 24, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17114A399.
    Description of amendment request: The amendment would revise 
Technical Specification requirements regarding steam generator tube 
inspections and reporting as described in Technical Specifications Task 
Force (TSTF) Traveler TSTF-510, Revision 2, ``Revision to Steam 
Generator Program Inspection Frequencies and Tube Sample Selection,'' 
using the Consolidated Line Item Improvement Process for Arkansas 
Nuclear One, Unit No. 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the Steam Generator (SG) Program to 
modify the frequency of verification of SG tube integrity and SG 
tube sample selection. A steam generator tube rupture (SGTR) event 
is one of the design basis accidents that are analyzed as part of a 
plant's licensing basis. The proposed SG tube inspection frequency 
and sample selection criteria will continue to ensure that the SG 
tubes are inspected such that the probability of a[n] SGTR is not 
increased. The consequences of a[n] SGTR are bounded by the 
conservative assumptions in the design basis accident analysis. The 
proposed change will not cause the consequences of a[n] SGTR to 
exceed those assumptions.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the SG Program will not introduce any 
adverse changes to the plant design basis or postulated accidents 
resulting from potential tube degradation. The proposed change does 
not affect the design of the SGs or their method of operation. In 
addition, the proposed change does not impact any other plant system 
or component.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
also isolate the radioactive fission products in the primary coolant 
from the secondary system. In summary, the safety function of a[n] 
SG is maintained by ensuring the integrity of its tubes.
    SG tube integrity is a function of the design, environment, and 
the physical condition of the tube. The proposed change does not 
affect tube design or operating environment. The proposed change 
will continue to require monitoring of the physical condition of the 
SG tubes such that there will not be a reduction in the margin of 
safety compared to the current requirements.
    Therefore, it is concluded that this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, 
Entergy Services, Inc., 101 Constitution Avenue NW., Suite 200 East, 
Washington, DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana

    Date of amendment request: March 28, 2017. A publicly-available 
version is

[[Page 31094]]

in ADAMS under Accession No. ML17087A551.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.8.1.3, ``Diesel Fuel Oil,'' by 
relocating the current stored diesel fuel oil numerical volume 
requirements from the TS to the TS Bases. In addition, the proposed 
amendment would revise TS 3.8.1.1, ``A.C. [Alternating Current] 
Sources--Operating,'' and TS 3.8.1.2, ``A.C. Sources--Shutdown,'' to 
relocate the specific numerical value for feed tank fuel oil volume to 
the TS Bases and replace it with the feed tank time requirement. The 
proposed changes are consistent with Technical Specifications Task 
Force (TSTF) Traveler TSTF-501, Revision 1, ``Relocate Fuel Oil and 
Lube Oil Volume Values to Licensee 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 revise [TS] 3.8.1.3 (Diesel Fuel Oil) by 
removing the current stored diesel fuel oil numerical volume 
requirements from the TS and replacing them with diesel operating 
time requirements. The specific volume of fuel oil equivalent to a 7 
and 6 day supply is calculated using the NRC approved methodology 
described in Regulatory Guide 1.137, Revision 1, ``Fuel-Oil Systems 
for Standby Diesel Generators'' and [American Nuclear Standards 
Institute (ANSI)] N195-1976, ``Fuel Oil Systems for Standby Diesel-
Generators'' using the time dependent load method as approved in 
Waterford 3 License Amendment 157. Because the requirement to 
maintain a 7 day supply of diesel fuel oil is not changed and is 
consistent with the assumptions in the accident analyses, and the 
actions taken when the volume of fuel oil is less than a 6 day 
supply have not changed, neither the probability nor the 
consequences of any accident previously evaluated will be affected.
    The proposed change also removes the TS 3.8.1.1 and TS 3.8.1.2 
diesel feed tank fuel oil numerical volume requirements and replaces 
them with the diesel one hour diesel generator operation 
requirement. The specific volume and time is not changed and is 
consistent with the existing plant design basis to support a diesel 
generator under accident load 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 change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The change does not involve a physical alteration of the plant 
(i.e., no new or different type of equipment will be installed) or a 
change in the methods governing normal plant operation. The change 
does not alter assumptions made in the safety analysis but ensures 
that the diesel generator operates as assumed in the accident 
analysis. The proposed change is consistent with the safety analysis 
assumptions. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes revise [TS] 3.8.1.3 (Diesel Fuel Oil) by 
removing the current stored diesel fuel oil numerical volume 
requirements from the TS and replacing them with diesel operating 
time requirements. As the bases for the existing limits on diesel 
fuel oil are not changed, no change is made to the accident analysis 
assumptions and no margin of safety is reduced as part of this 
change.
    The proposed change also removes the TS 3.8.1.1 and TS 3.8.1.2 
diesel feed tank fuel oil numerical volume requirements and replaces 
them with the diesel one hour diesel generator operation 
requirement. As the basis for the existing limits on diesel fuel oil 
are not changed, no change is made to the accident analysis 
assumptions and no margin of safety is reduced as part of this 
change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Anna Vinson Jones, Senior Counsel, 
Entergy Services, Inc., 101 Constitution Avenue NW., Suite 200 East, 
Washington, DC 20001.
    NRC Branch Chief: Robert J. Pascarelli.

Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket No. 50-277, 
Peach Bottom Atomic Power Station (PBAPS), Unit 2, York and Lancaster 
Counties, Pennsylvania

    Date of amendment request: May 19, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17139D357.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TSs) to decrease the number of safety relief 
valves and safety valves required to be operable when operating at a 
power level less than or equal to 3358 megawatts thermal (MWt). This 
change would be in effect for the current PBAPS, Unit 2, Cycle 22 that 
is scheduled to end in October 2018.
    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 change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would revise TS Section 3.4.3 to decrease 
the required number of Safety Relief Valves (SRVs) and Safety Valves 
(SVs) from a total of 13 to 12, under reduced reactor thermal power 
operation of 3358 MWt (approximately 85% of Current Licensed Thermal 
Power (CLTP)). A compensatory reduction in maximum allowed reactor 
power to 3358 MWt has been determined to conservatively offset the 
impact/effects of operation with an additional (up to 2) SRVs/SVs 
Out-of-Service. The Reactor Pressure Vessel (RPV) overpressure 
protection capability of the 12 operable SRVs and SVs is adequate at 
the lower power level to ensure the ASME [American Society of 
Mechanical Engineers] code allowable peak pressure limits are not 
exceeded. With the maximum thermal power limitation condition, the 
proposed change has no adverse effect on plant operation, or the 
availability or operation of any accident mitigation equipment. The 
plant response to the design basis accidents, Anticipated 
Operational Occurrence (AOO) events and Special Events remains 
bounded by existing analyses. The proposed change does not require 
any new or unusual operator actions. The proposed change does not 
introduce any new failure modes that could result in a new or 
different accident. The SRVs and SVs are not being modified or 
operated differently and will continue to operate to meet the design 
basis requirements for RPV overpressure protection. The proposed 
change does not alter the manner in which the RPV overpressure 
protection system is operated and functions and thus, there is no 
significant impact on reactor operation. There is no change being 
made to safety limits or limiting safety system settings that would 
adversely affect plant safety as a result of the proposed change.
    For PBAPS, the limiting overpressure AOO event is the main steam 
isolation valve closure with scram on high flux (MSIVF). The PBAPS 
ATWS [anticipated transients without scram] Special Event evaluation 
considered the limiting cases for RPV overpressure and is analyzed 
under two cases: (1) Main Steam Isolation Valve Closure (MSIVC) and 
(2) Pressure Regulator Failure Open (PRFO). These events were 
analyzed under the proposed conditions and it was confirmed that the 
existing analyses remain bounding for the condition of adding a

[[Page 31095]]

second SRV/SV Out-of-Service with a limited maximum operating power 
level of 3358 MWt.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change would revise TS Section 3.4.3 to decrease 
the required number of SRVs and SVs from a total of 13 to 12, under 
reduced reactor thermal power operation of 3358 MWt (approximately 
85% of CLTP). A compensatory reduction in maximum allowed reactor 
power to 3358 MWt has been determined to conservatively offset the 
impact/effects of operation with an additional (up to 2) SRVs/SVs 
Out-of-Service. The RPV overpressure protection capability of the 12 
operable SRVs and SVs is adequate at the lower power level to ensure 
the ASME code allowable peak pressure limits are not exceeded. The 
SRVs and SVs are not being modified or operated differently and will 
continue to operate to meet the design basis requirements for RPV 
overpressure protection. The proposed change does not introduce any 
new failure modes that could result in a new or different accident. 
The proposed reactor thermal power restriction of 3358 MWt is within 
the existing normal operating domain and no new or special operating 
actions are necessary to operate at the intermediate power level. 
The proposed change does not alter the manner in which the RPV 
overpressure protection system is operated and functions and thus, 
there is no new failure mechanisms for the overpressure protection 
system. The plant response to the design basis accidents, AOO events 
and Special Events remains bounded by existing analyses. [These] 
events were analyzed under the proposed conditions and it was 
confirmed that the existing analyses remain bounding for the 
condition of adding a second SRV/SV Out-of-Service with a limited 
maximum operating power level of 3358 MWt.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established though the design of the 
plant structures, systems and components, the parameters within 
which the plant is operated, and the establishment of setpoints for 
the actuation of equipment relied upon to respond to an event. The 
proposed change does not change the setpoints at which the 
protective actions are initiated. The proposed change would revise 
TS Section 3.4.3 to decrease the required number of SRVs and SVs 
under reduced reactor thermal power operation of 3358 MWt 
(approximately 85% of CLTP). A compensatory reduction in maximum 
allowed reactor power to 3358 MWt has been determined to 
conservatively offset the impact/effects of operation with an 
additional (up to 2) SRVs/SVs Out-of-Service. The RPV overpressure 
protection capability of the 12 operable SRVs and SVs is adequate at 
the lower power level to ensure the ASME code allowable peak 
pressure limits are not exceeded. The plant response to the design 
basis accidents, AOO events and Special Events remains bounded by 
existing analyses. These events were analyzed under the proposed 
conditions and it was confirmed that the existing analyses remain 
bounding for the condition of adding a second SRV/SV Out-of-Service 
with a limited maximum operating power level of 3358 MWt.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL 
60555.
    NRC Branch Chief: James G. Danna.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station (CPS), Unit No.1, DeWitt County, Illinois

    Date of amendment request: May 4, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17124A121.
    Description of amendment request: The proposed change would delete 
a surveillance requirement (SR) Note associated with technical 
specification (TS) 3.5.1, ``ECCS [emergency core cooling system]--
Operating,'' TS 3.5.2, ``ECCS--Shutdown,'' and TS 3.6.1.7, ``Residual 
Heat Removal (RHR) Containment Spray System,'' to more appropriately 
reflect the RHR system design, and ensure the RHR system operation is 
consistent with the TS limiting condition for operation (LCO) 
requirements. In addition, the proposed amendment would insert a Note 
in the LCO for TSs 3.5.1, 3.5.2, 3.6.1.7, 3.6.1.9, ``Feedwater Leakage 
Control System,'' and 3.6.2.3, ``Residual Heat Removal (RHR) 
Suppression Pool Cooling,'' to clarify that one of the required 
subsystems in each of the affected TS sections may be inoperable during 
alignment and operation of the RHR system for shutdown cooling (SDC) 
with the reactor steam dome pressure less than the RHR cut in 
permissive value.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    No physical changes to the facility will occur as a result of 
this proposed amendment. The proposed changes will not alter the 
physical design. The current TS (CTS) Note in SR 3.5.1.4, SR 
3.5.2.4, and 3.6.1.7 could make CPS susceptible to potential water 
hammer in the RHR system while operating in the SDC mode of RHR in 
MODE 3 when swapping from the SDC to LPCI [low-pressure coolant 
injection] and RHR containment spray modes of RHR. Deletion of the 
Note from SR 3.5.1.2, SR 3.5.2.4, and SR 3.6.1.7.1 will eliminate 
the risk for cavitation of the pump and voiding in the suction 
piping, thereby avoiding the potential to damage the RHR system, 
including water hammer. The addition of proposed TS note to LCO 
3.5.1, LCO 3.5.2, LCO 3.6.1.7, LCO 3.6.1.9, and LCO 3.6.2.3 will re-
establish consistency of the CPS RHR system design with the original 
TS requirements.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not alter the physical design, safety 
limits, or safety analysis assumptions associated with the operation 
of the plant. Accordingly, the change does not introduce any new 
accident initiators, nor does it reduce or adversely affect the 
capabilities of any plant structure, system, or component to perform 
their safety function. Deletion of the Note from SR 3.5.1.2, SR 
3.5.2.4 and SR 3.6.1.7.1 is appropriate because current TSs could 
put the plant at risk for potential cavitation of the pump and 
voiding in the suction piping, resulting in potential to damage the 
RHR system, including water hammer. The addition of proposed TS note 
to LCO 3.5.1, LCO 3.5.2, LCO 3.6.1.7, LCO 3.6.1.9, and LCO 3.6.2.3 
will re-establish consistency of the CPS RHR system design with the 
original TS requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change conforms to NRC regulatory guidance 
regarding the content of plant Technical Specifications. The 
proposed change does not alter the physical design, safety limits, 
or safety analysis assumptions associated with the operation of the 
plant.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 31096]]

review it appears the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: David J. Wrona.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station, Unit No.1, DeWitt County, Illinois

    Date of amendment request: May 1, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17121A517.
    Description of amendment request: The proposed change replaces 
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 reactor vessel water level to be greater than the top of 
active irradiated fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

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

    The NRC staff has reviewed the licensee's analysis and, based on 
this review it appears the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: David J. Wrona.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station (LGS), Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of amendment request: April 24, 2017. A publicly available 
version is in ADAMS under Accession No. ML17115A087.
    Description of amendment request: The amendments would revise the 
LGS, Units 1 and 2, Technical Specifications (TSs) to a set of Improved 
Technical Specifications (ITS) based on NUREG-1433, Revision 4, 
``Standard Technical Specifications--General Electric Plants, BWR/4,'' 
published April 2012. Specifically, the amendments would relocate TS 
Section 3.3.7.12, ``Offgas Gas Monitoring Instrumentation''; TS 
3.11.2.5, ``Explosive Gas Mixture''; and Surveillance Requirement (SR) 
4.11.2.6.1, which requires continuously monitoring the main condenser 
gaseous effluent to the LGS Offsite Dose Calculation Manual or to the 
LGS Technical Requirements Manual. In

[[Page 31097]]

addition, associated with the relocation of the main condenser offgas 
noble gas activity monitor, (1) SR 4.11.2.6.2.b will be changed to 
account for the relocated instrument's requirements, and (2) associated 
with the relocation of the explosive gas mixture instrumentation and 
gaseous effluent TS sections, a new TS Program Section, 6.8.4.l, 
``Explosive Gas Monitoring Program,'' will be added to TS Section 6.8, 
``Procedures and Programs.''
    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 relocate certain operability and 
surveillance requirements for the Main Condenser Offgas Monitoring 
Instrumentation and Gaseous Effluents limits from the Limerick 
Generating Station (LGS) Technical Specifications (TS) to a 
licensee-controlled document under the control of 10 CFR 50.59 or 
under the control of regulatory requirements applicable to the 
licensee-controlled document. A new TS Administrative Program is 
proposed to be added to ensure the limit for Main Condenser Offgas 
hydrogen concentration is maintained.
    The proposed changes do not alter the physical design of any 
plant structure, system, or component; therefore, the proposed 
changes have no adverse effect on plant operation, or the 
availability or operation of any accident mitigation equipment. The 
plant response to the design basis accidents does not change. 
Operation or failure of the Main Condenser Offgas Radioactivity and 
Hydrogen Monitors capability are not assumed to be an initiator of 
any analyzed event in the Updated Final Safety Analysis Report 
(UFSAR) and cannot cause an accident. Whether the requirements for 
the Main Condenser Offgas Radioactivity and Hydrogen Monitor 
capability are located in TS or another licensee-controlled document 
has no effect on the probability or consequences of any accident 
previously evaluated.
    The proposed changes conform to NRC regulatory requirements 
regarding the content of plant TS as identified in 10 CFR 50.36, and 
also the guidance as approved by the NRC in NUREG-1433, ``Standard 
Technical Specifications--General Electric BWR/4 Plants.''
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any 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 relocate certain operability and 
surveillance requirements for the Main Condenser Offgas Monitoring 
Instrumentation and Gaseous Effluents limits from the LGS TS to a 
licensee-controlled document under the control of 10 CFR 50.59 or 
under the control of regulatory requirements applicable to the 
licensee-controlled document. A new TS Administrative Program is 
proposed to be added to ensure the limit for Main Condenser Offgas 
hydrogen concentration is maintained.
    The proposed changes do not alter the plant configuration (no 
new or different type of equipment is being installed) or require 
any new or unusual operator actions. The proposed changes do not 
alter the safety limits or safety analysis assumptions associated 
with the operation of the plant. The proposed changes do not 
introduce any new failure modes that could result in a new accident. 
The proposed changes do not reduce or adversely affect the 
capabilities of any plant structure, system, or component in the 
performance of their safety function. Also, the response of the 
plant and the operators following the design basis accidents is 
unaffected by the proposed changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes relocate certain operability and 
surveillance requirements for the Main Condenser Offgas Monitoring 
Instrumentation and Gaseous Effluents limits from the LGS TS to a 
licensee-controlled document under the control of 10 CFR 50.59 or 
under the control of regulatory requirements applicable to the 
licensee-controlled document. A new TS Administrative Program is 
proposed to be added to ensure the limit for the Main Condenser 
Offgas hydrogen concentration is maintained. The relocated TS 
requirements do not meet any of the 10 CFR 50.36c(2)(ii) criteria on 
items for which a TS must be established.
    The proposed changes have no adverse effect on plant operation, 
or the availability or operation of any accident mitigation 
equipment. The plant response to the design basis accidents does not 
change. The proposed changes do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to 
operate in the safety analyses. There is no change being made to 
safety analysis assumptions, safety limits or limiting safety system 
settings that would adversely affect plant safety as a result of the 
proposed changes.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: James G. Danna.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant (PNPP), Unit No. 1, Lake County, Ohio

    Date of amendment request: April 26, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17116A575.
    Description of amendment request: The proposed amendment would 
revise the PNPP Environmental Protection Plan (nonradiological) to 
clarify and enhance wording, to remove duplicative or outdated program 
information, and to relieve the burden of submitting unnecessary or 
duplicative information to 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 amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment involves changes to the Environmental 
Protection Plan (EPP), which provides for protection of 
nonradiological environmental values during operation of the nuclear 
facility. The proposed amendment does not change the objectives of 
the EPP, does not change the way the plant is maintained or 
operated, and does not affect any accident mitigating feature or 
increase the likelihood of malfunction for plant structures, systems 
and components.
    The proposed amendment will not change any of the analyses 
associated with the PNPP Updated Safety Analysis Report Chapter 15 
accidents because plant operation, plant structures, systems, 
components, accident initiators, and accident mitigation functions 
remain unchanged.
    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 amendment involves changes to the EPP, which 
provides for protection of nonradiological environmental values 
during operation of the nuclear facility. The proposed amendment 
does not involve a physical alteration of the plant. No new or 
different type of equipment will be installed, and there are no 
physical modifications to existing installed equipment associated 
with the proposed changes. The

[[Page 31098]]

proposed amendment does not change the way the plant is operated or 
maintained and does not create a credible failure mechanism, 
malfunction or accident initiator not already considered in the 
design and licensing basis.
    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.
    Safety margins are applied to design and licensing basis 
functions and to the controlling values of parameters to account for 
various uncertainties and to avoid exceeding regulatory or licensing 
limits. The proposed amendment involves changes to the EPP, which 
provides for protection of nonradiological environmental values 
during operation of the nuclear facility. The proposed amendment 
does not involve a physical change to the plant, does not change 
methods of plant operation within prescribed limits, or affect 
design and licensing basis functions or controlling values of 
parameters for plant systems, structures, and components.
    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 W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: David J. Wrona.

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: May 2, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17144A294.
    Description of amendment request: The amendments would revise the 
St. Lucie Plant Unit Nos. 1 and 2 Renewed Facility Operating Licenses, 
Nos. DPR-67 and NPF-16, respectively, fire protection license 
conditions. The revisions would incorporate new references into these 
license conditions that propose and approve a revision to plant 
modifications previously approved in the March 31, 2016, NRC issuance 
of amendments regarding transition to a risk-informed, performance-
based fire protection program in accordance with 10 CFR 50.48(c), dated 
March 21, 2016 (ADAMS Accession No. ML15344A346) (known as the National 
Fire Protection Association Standard 805 (NFPA 805)).
    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 are clarifications to methods applied to 
ensure compliance with NFPA 30, section 2348. The revised methods 
comply with NFPA 30, section 2348. This LAR [license amendment 
request] is essentially an administrative change to revise the 
letter referenced by the Fire Protection Transition License 
Conditions. The actual design changes and any related procedural 
changes are being managed separately from this LAR per 10 CFR 50.59.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed changes do not 
adversely affect the ability of structures, systems and components 
(SSCs) to perform their intended safety function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change does not increase the probability or 
consequence of an accident.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are clarifications to methods applied to 
ensure compliance with NFPA 30, section 2348. The revised methods of 
compliance align with NFPA 30, section 2348, and will not result in 
new or different kinds of accidents. This LAR is essentially an 
administrative change to revise the letter referenced by the Fire 
Protection Transition License Conditions. The actual design changes 
and any related procedural changes are being managed separately from 
this LAR per 10 CFR 50.59.
    The requirements in NFPA 30 address only fire protection. The 
impacts of fire effects on the plant have been evaluated. The 
proposed amendment does not involve new failure mechanisms or 
malfunctions that could initiate a new or different kind of accident 
beyond those already analyzed in the Unit 1 and Unit 2 UFSARs 
[updated final safety analysis reports].
    Therefore, this change does not create the possibility of a new 
or different kind of accident from an accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Operation of Plant St. Lucie (PSL) in accordance with the 
proposed amendment does not involve a reduction in the margin of 
safety. The proposed amendment does not alter the manner in which 
safety limits, limiting safety system settings or limiting 
conditions for operation are determined. The safety analysis 
acceptance criteria are not affected by this change. The proposed 
amendment does not adversely affect existing plant safety margins or 
the reliability of equipment assumed to mitigate accidents in the 
UFSAR. The proposed amendment does not adversely affect the ability 
of SSCs to perform their design function. SSCs required to safely 
shut down the reactor and to maintain it in a safe shutdown 
condition remain capable of performing their design function.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Undine S. Shoop.

Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant (CNP), Units Nos. 1 and 2, Berrien County, 
Michigan

    Date of amendment request: May 23, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17146A073.
    Description of amendment request: The proposed changes update the 
emergency action levels (EALs) used at CNP, Unit Nos. 1 and 2, from the 
current scheme based on Nuclear Management and Resources Council 
(NUMARC) and National Environmental Studies Project (NESP) NUMARC/NESP-
007, ``Methodology for Development of Emergency Action Levels'' dated 
January 1992, to a scheme based on Nuclear Energy Institute 99-01, 
Revision 6, ``Development of Emergency Action Levels for Non-Passive 
Reactors.''
    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.

[[Page 31099]]

    The proposed changes to the CNP EALs do not impact the physical 
function of plant structures, systems, or components (SSC) or the 
manner in which SSCs perform their design function. EALs are used as 
criteria for determining the need for notification and participation 
of local and State agencies, and for determining when and what type 
of protective measures should be considered within and outside the 
site boundary to protect health and safety. 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the CNP EALs do not involve any physical 
changes to plant systems or equipment. The proposed changes do not 
involve the addition of any new equipment. EALs are based on plant 
conditions, so the proposed changes will not alter the design 
configuration or the method of plant operation. The proposed changes 
will not introduce failure modes that could result in a new or 
different type of accident, and the change does not alter 
assumptions made in the safety analysis. The proposed changes to the 
CNP Emergency Plan are not initiators of any accidents.
    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 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 to the CNP EALs 
do not impact operation of the plant or its response to transient 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 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.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed changes 
involve references to available plant indications to assess 
conditions for determination of entry into an emergency action 
level. There is no change to these established safety margins as a 
result of this change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: David J. Wrona.

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: May 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17135A225.
    Description of amendment request: The requested amendment proposes 
to depart from combined license (COL) Appendix C information (with 
corresponding changes to the associated plant-specific Tier 1 
information) and involves associated Tier 2 information in the Updated 
Final Safety Analysis Report (UFSAR). Specifically, proposed changes 
clarify that there is more than one turbine building main sump and adds 
a second sump pump for each of the two turbine building main sumps into 
UFSAR Tier 2 and COL Appendix C (and associated plant-specific 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, 
design certification rule is also requested for the plant-specific 
Design Control Document Tier 1 departures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The activity adds a second pump to each of the turbine building 
main sumps, and identifies that there is more than one turbine 
building sump. The reason for the additional pumps is to account for 
an increase in volume due to the changes to the [condensate 
polishing system (CPS)] rinse effluent flowpath from [component 
cooling water system (CCW)] CCW to [waste water system (WWS)] WWS 
via the Turbine Building sumps. The extra sump pumps will prevent 
potential overflowing and flooding of the sumps during CPS rinse 
operations. The CPS serves no safety-related function. By directing 
the effluent to the turbine building sumps it is subject to 
radiation monitoring. Under normal operating conditions, there are 
no significant amounts of radioactive contamination within the CPS. 
However, radioactive contamination of the CPS can occur as a result 
of a primary to secondary leakage in the steam generator should a 
steam generator tube leak develop while the CPS is in operation and 
radioactive condensate is processed by the CPS. Radiation monitors 
associated with the steam generator blowdown, steam generator, and 
turbine island vents, drains and relief systems provide the means to 
determine if the secondary side is radioactively contaminated. The 
main turbine building sumps and sump pumps are not safety-related 
components and do not interface with any systems, structures, or 
components (SSC) accident initiator or initiating sequence of 
events; thus, the probability of accidents evaluated within the 
plant-specific UFSAR are not affected. The proposed changes do not 
involve a change to the predicted radiological releases due to 
accident conditions, thus the consequences of accidents evaluated in 
the UFSAR are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the non-safety waste water system (WWS) 
do not affect any safety-related equipment, nor does it add any new 
interface to safety-related SSCs. No system or design function or 
equipment qualification is affected by this change. The changes do 
not introduce a new failure mode, malfunction, or sequence of events 
that could affect safety or safety-related equipment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The WWS is a nonsafety-related system that does not interface 
with any safety-related equipment. The proposed changes to identify 
that there is more than one turbine building sump and to add two 
turbine building sump pumps do not affect any design code,

[[Page 31100]]

function, design analysis, safety analysis input or result, or 
design/safety margin. No safety analysis or design basis acceptance 
limit/criterion is challenged or exceeded by the proposed change.
    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.

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: May 16, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17137A107.
    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 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: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Jennifer Dixon-Herrity.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment request: May 16, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17142A315.
    Description of amendment request: The proposed amendment would 
revise the Facility Operating Licenses for the San Onofre Nuclear 
Generating Station (SONGS), Units 2 and 3, to reflect deletion of the 
Cyber Security Plan from License Condition 2.E. This will allow 
Southern California Edison (SCE) to terminate the SONGS Cyber Security 
Plan and associated activities at the site. These changes will more 
fully reflect the permanently shutdown and defueled status of the 
facility, as well as the reduced scope of potential radiological 
accidents and security concerns that exist during the decommissioning 
process.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to remove the San Onofre Nuclear Generating 
Station (SONGS) Cyber Security Plan requirement does not alter 
accident analysis assumptions, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected. The proposed 
change does not require any plant modifications which affect the 
performance capability of the structures, systems, and components 
(SSCs) relied upon to mitigate the consequences of postulated 
accidents, and has no impact on the probability or consequences of 
an accident previously evaluated.
    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 change to remove the SONGS Cyber Security Plan 
requirement does not alter accident analysis assumptions, add any 
initiators, or affect the function of plant systems or the manner in 
which systems are operated, maintained, modified, tested, or 
inspected. The proposed change does not require any plant 
modifications which affect the performance capability of the SSCs 
relied upon to mitigate the consequences of postulated accidents, 
and does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
for operation,

[[Page 31101]]

limiting safety system settings, and safety limits specified in the 
technical specifications. The proposed change to the SONGS Cyber 
Security Plan does not change these established safety margins. 
Therefore the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Walker A. Matthews, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Branch Chief: Bruce Watson, CHP.

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

    Date of amendment request: May 5, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17125A331.
    Description of amendment request: The amendment request proposes to 
depart from plant-specific Tier 1 emergency planning inspection, test, 
analysis, and acceptance criteria (ITAAC) information and associated 
combined license (COL) Appendix C information. The proposed changes do 
not involve changes to the approved emergency plan or the plant-
specific Tier 2 Design Control Document (DCD). Specifically, the 
requested amendment proposes to revise plant-specific emergency 
planning inspections (ITAAC) in Appendix C of the VEGP Units 3 and 4 
COLs. Also, proposed changes to COL Appendix C information also include 
changes to the list of acronyms and abbreviations. Because, this 
proposed change requires a departure from Tier 1 information in the 
Westinghouse Electric Company's AP1000 Design 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 VEGP 3 and 4 emergency planning inspections, tests, 
analyses, and acceptance criteria (ITAAC) provide assurance that the 
facility has been constructed and will be operated in conformity 
with the license, the provisions of the Act, and the Commission's 
rules and regulations. The proposed changes do not affect the design 
of a system, structure, or component (SSC) use to meet the design 
bases of the nuclear plant. Nor do the changes affect the 
construction or operation of the nuclear plant itself, so there is 
no change to the probability or consequences of an accident 
previously evaluated. Changing the VEGP 3 and 4 emergency planning 
ITAAC and COL, Appendix C, list of acronyms and abbreviations do not 
affect prevention and mitigation of abnormal events (e.g., 
accidents, anticipated operational occurrences, earthquakes, floods, 
or turbine missiles) or their safety or design analyses. No safety-
related structure, system, component (SSC) or function is adversely 
affected. The changes neither involve nor interface with any SSC 
accident initiator or initiating sequence of events, so the 
probabilities of the accidents evaluated in the Updated Final Safety 
Analysis Report (UFSAR) are not affected. Because the changes do not 
involve any safety-related SSC or function used to mitigate an 
accident, the consequences of the accidents evaluated in the UFSAR 
are not affected.
    Therefore, the requested 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 VEGP 3 and 4 emergency planning ITAAC provide assurance that 
the facility has been constructed and will be operated in conformity 
with the license, the provisions of the Act, and the Commissioner's 
rules and regulations. The changes do not affect the design of an 
SSC used to meet the design bases of the nuclear plant. Nor do the 
changes affect the construction or operation of the nuclear plant. 
Consequently, there is no new or different kind of accident from any 
accident previously evaluated. The changes do not affect safety-
related equipment, nor do they affect equipment that, if it failed, 
could initiate an accident or a failure of a fission product 
barrier. In addition, the changes do not result in a new failure 
mode, malfunction, or sequence of events that could affect safety or 
safety-related equipment.
    No analysis is adversely affected. No system or design function 
or equipment qualification is adversely affected by the changes. 
This activity will not allow for a new fission product release path, 
nor will it result in a new fission product barrier failure mode, 
nor create a new sequence of events that would result in significant 
fuel cladding failures.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    2. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The VEGP 3 and 4 emergency planning ITAAC provide assurance that 
the facility has been constructed and will be operated in conformity 
with the license, the provisions of the Act, and the Commissioner's 
rules and regulations. The changes do not affect the assessments or 
the plant itself. The changes do not adversely affect the safety-
related equipment or fission product barriers. No safety analysis or 
design basis acceptance limit or criterion is challenged or exceeded 
by the proposed change.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: May 19, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17139D394.
    Description of amendment request: The requested amendment proposes 
to depart from combined license (COL) Appendix C information (with 
corresponding changes to the associated plant-specific Tier 1 
information) and involves associated Tier 2 information in the Updated 
Final Safety Analysis Report (UFSAR). Specifically, proposed changes 
clarify that there is more than one turbine building main sump and adds 
a second sump pump for each of the two turbine building main sumps into 
the UFSAR Tier 2 and COL Appendix C (and associated plant-specific 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, 
design certification rule is also requested for the plant-specific 
Design Control Document Tier 1 departures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or

[[Page 31102]]

consequences of an accident previously evaluated?
    Response: No.
    The activity adds a second pump to each of the turbine building 
main sumps, and identifies that there is more than one turbine 
building sump. The reason for the additional pumps is to account for 
an increase in volume due to the changes to the condensate polishing 
system (CPS) rinse effluent flowpath from CPS to waste water system 
(WWS) via the turbine building sumps. The extra sump pumps will 
prevent potential overflowing and flooding of the sumps during CPS 
rinse operations. The CPS serves no safety-related function. By 
directing the effluent to the turbine building sumps it is subject 
to radiation monitoring. Under normal operating conditions, there 
are is no significant amount of radioactive contamination within the 
CPS. However, radioactive contamination of the CPS can occur as a 
result of a primary-to-secondary leakage in the steam generator 
should a steam generator tube leak develop while the CPS is in 
operation and radioactive condensate is processed by the CPS. 
Radiation monitors associated with the steam generator blowdown, 
steam generator, and turbine island vents, drains and relief systems 
provide the means to determine if the secondary side is 
radioactively contaminated. The main turbine building sumps and sump 
pumps are not safety-related components and do not interface with 
any systems, structures, or components (SSC) accident initiator or 
initiating sequence of events; thus, the probability of accidents 
evaluated within the plant-specific UFSAR are not affected. The 
proposed changes do not involve a change to the predicted 
radioactive releases due to accident conditions, thus the 
consequences of accidents evaluated in the UFSAR are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the nonsafety-related WWS do not affect 
any safety-related equipment, nor do they add any new interface to 
safety-related SSCs. No system or design function or equipment 
qualification is affected by this change. The changes do not 
introduce a new failure mode, malfunction, or sequence of events 
that could affect safety or safety-related equipment. Therefore, the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The WWS is a nonsafety-related system that does not interface 
with any safety-related equipment. The proposed changes to identify 
that there is more than one turbine building sump and to add two 
turbine building sump pumps do not affect any design code, function, 
design analysis, safety analysis input or result, or design/safety 
margin. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed change.
    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.

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

    Date of amendment request: March 13, 2017. A publicly available 
version is in ADAMS under Accession No. ML17073A018.
    Description of amendment request: The amendments would modify the 
Surveillance Requirement (SR) 3.8.1.17 of the Technical Specification 
(TS) 3.8.1, ``AC [Alternating Current] Sources--Operating,'' to delete 
the note to allow the performance of the SR in Modes 1 through 4 when 
the associated load is out of service for maintenance or testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The proposal does not alter the function of any structure, 
system or component functions, does not modify the manner in which 
the plant is operated, and does not alter equipment out-of-service 
time. This request does not degrade the ability of the emergency 
diesel generator or equipment downstream of the load sequencers to 
perform their intended function.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve any physical changes to 
plant safety related structure, system or component or alter the 
modes of plant operation in a manner that is outside the bounds of 
the current emergency diesel generator system design analyses. The 
proposed change to revise the note modifying SR 3.8.1.17 to allow 
the performance of the SR in Modes 1 through 4 when the associated 
equipment is out of service for maintenance or testing does not 
create the possibility for an accident or malfunction of a different 
type than any evaluated previously in SQN's Updated Final Safety 
Analysis Report. The proposal does not alter the way any structure, 
system or component function and does not modify the manner in which 
the plant is operated. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to TS 3.8.1, ``AC Sources--Operating'' to 
revise the note modifying SR 3.8.1.17 to allow the performance of 
the SR in Modes 1 through 4 when the associated equipment is out of 
service for maintenance or testing does not reduce the margin of 
safety because the test methodologies are not being changed and LCO 
[limiting condition for operation] allowed outage times are not 
being changed. The results of accident analyses remain unchanged by 
this request. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Branch Chief: Undine S. Shoop.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant 
(WBN), Unit 1, Rhea County, Tennessee

    Date of amendment request: March 31, 2017. A publicly available 
version is in ADAMS under Accession No. ML17093A854.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 5.7.2.14, ``Ventilation Filter Testing 
Program (VFTP),'' to delete references to the reactor building (RB) 
purge filters. A previous amendment deleted the reactor building purge 
air cleanup system from the TSs based on partial implementation of the 
alternate source term methodology; however, references to the RB purge 
filters were not removed from TS 5.7.2.14 at that time due to an 
administrative oversight. The proposed change corrects the 
administrative

[[Page 31103]]

oversight by deleting references to the RB purge filters in TS 
5.7.2.14.
    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 WBN TS 5.7.2.1.14 is administrative in 
nature. Nuclear Regulatory Commission (NRC) Amendment Number 92 
(ML13141A564) deleted TS 3.9.8, ``Reactor Building Purge Air Cleanup 
Units,'' based on implementation of the alternate source term (AST) 
methodology because no credit is taken for the operation of reactor 
building air cleanup units for the dose analysis during a fuel 
handling accident (FHA). However, TVA neglected to remove the 
references to the RB purge filters in TS 5.7.2.14. The proposed 
change corrects this oversight by deleting the references to the RB 
purge filters in TS 5.7.2.14a. through d.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes would not require any new or different 
accidents to be postulated and subsequently evaluated because no 
changes are being made to the plant that would introduce any new 
accident causal mechanisms. This license amendment request does not 
impact any plant systems that are potential accident initiators, nor 
does it have any significantly adverse impact on any accident 
mitigating systems. No new or different accident scenarios, 
transient precursors, failure mechanisms, or limiting single 
failures will be introduced as a result of these changes.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change does not alter the permanent plant design, 
including instrument setpoints, nor does it change the assumptions 
contained in the safety analyses. Margin of safety is related to the 
ability of the fission product barriers to perform their design 
functions during and following accident conditions. These barriers 
include the fuel cladding, the reactor coolant system, and the 
containment system. The performance of these barriers will not be 
significantly degraded by the proposed changes.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Branch Chief: Undine S. Shoop.

Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant, 
Unit 2, Rhea County, Tennessee

    Date of amendment request: March 28, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17093A608.
    Description of amendment request: The amendment would revise the 
Facility Operating License (OL) to extend the completion date for 
Condition 2.C.(5) regarding the reporting of actions taken to resolve 
issues identified in Nuclear Regulatory Commission Bulletin 2012-01, 
``Design Vulnerability in Electric Power System,'' dated July 27, 2012 
(ADAMS Accession No. ML12074A115).
    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 revise the completion date for OL 
Condition 2.C(5) for WBN Unit 2 regarding the reporting of actions 
taken to resolve issues identified in NRC Bulletin 2012-01 from 
December 31, 2017 to December 31, 2018 do not affect the structures, 
systems, or components (SSCs) of the plant, affect plant operations, 
or any design function or any analysis that verifies the capability 
of an SSC to perform a design function. No change is being made to 
any of the previously evaluated accidents in the WBN Updated Final 
Safety Analysis Report (UFSAR).
    The proposed changes do not (1) require physical changes to 
plant SSCs; (2) prevent the safety function of any safety-related 
system, structure, or component during a design basis event; (3) 
alter, degrade, or prevent action described or assumed in any 
accident described in the WBN UFSAR from being performed because the 
safety-related SSCs are not modified; (4) alter any assumptions 
previously made in evaluating radiological consequences; or (5) 
affect the integrity of any fission product barrier.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not introduce any new accident causal 
mechanisms, because no physical changes are being made to the plant, 
nor do they affect any plant systems that are potential accident 
initiators.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed changes will have no effect 
on the availability, operability, or performance of safety-related 
systems and components. The proposed change will not adversely 
affect the operation of plant equipment or the function of equipment 
assumed in the accident analysis.
    The proposed amendment does not involve changes to any safety 
analyses assumptions, safety limits, or limiting safety system 
settings. The changes do not adversely affect plant-operating 
margins or the reliability of equipment credited in the safety 
analyses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Branch Chief: Undine S. Shoop.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as

[[Page 31104]]

applicable, proposed no significant hazards consideration 
determination, and opportunity for a hearing in connection with these 
actions, was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.

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

    Date of amendment request: November 18, 2016.
    Brief description of amendments: The amendments adopted the 
approved Technical Specification Task Force (TSTF) Improved Standard 
Technical Specifications Change Traveler TSTF-535, revising the 
Technical Specification definition of Shutdown Margin (SDM) to require 
calculation of the SDM at a reactor moderator temperature of 68 degrees 
Fahrenheit, or a higher temperature that represents the most reactive 
state throughout the operating cycle.
    Date of issuance: June 7, 2017.
    Effective date: As of date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 277 and 305. A publicly-available version is in 
ADAMS under Accession No. ML17088A396; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: January 17, 2017 (82 FR 
4929).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 7, 2017.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear 
Plant, Van Buren County, Michigan

    Date of amendment request: November 9, 2016.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.5.10, ``Ventilation Filter Testing Program,'' to 
correct and modify the description of the control room ventilation and 
fuel handling area ventilation systems. In addition, the amendment 
corrects an editorial omission in TS Limiting Condition for Operation 
3.0.9.
    Date of issuance: June 8, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 263. A publicly-available version is in ADAMS under 
Accession No. ML17121A510; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-20: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: February 14, 2017 (82 
FR 10596).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 8, 2017.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: October 26, 2016.
    Brief description of amendment: The amendment changed the Technical 
Specifications (TS) to revise requirements for unavailable barriers by 
adding new Limiting Condition for Operation (LCO) 3.0.9. This LCO 
establishes conditions under which systems would remain operable when 
required physical barriers are not capable of providing their related 
support function. This amendment is consistent with NRC-approved 
Technical Specification Task Force (TSTF) Improved Standard Technical 
Specifications Change Traveler, TSTF-427, Revision 2, ``Allowance for 
Non Technical Specification Barrier Degradation on Supported System 
OPERABILITLY.'' The Notice of Availability of this TS improvement and 
the model application was published in the Federal Register on October 
3, 2006 (71 FR 58444), as part of the consolidated line item 
improvement process.
    Date of issuance: June 7, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No: 212. A publicly-available version is in ADAMS under 
Accession No. ML17116A032; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-29: The amendment revised the 
Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 20, 2016 (81 
FR 92866).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 7, 2017.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: November 1, 2016.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 2.1.1, ``Reactor Core Safety Limits,'' to reduce the 
reactor steam dome pressure value specified in TS 2.1.1.1 and TS 
2.1.1.2 from 785 pounds per square inch gauge (psig) to 686 psig.
    Date of issuance: June 19, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 176. A publicly-available version is in ADAMS under 
Accession No. ML17139C372; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-58: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: December 20, 2016 (81 
FR 92868).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 19, 2017.
    No significant hazards consideration comments received: No.

[[Page 31105]]

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant (CNP), Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of amendment request: October 18, 2016, as supplemented by 
letter dated February 27, 2017.
    Brief description of amendments: The amendments revised the CNP, 
Unit Nos. 1 and 2, Technical Specification 5.5.14, ``Containment 
Leakage Rate Testing Program,'' to clarify the containment leakage rate 
testing pressure criteria.
    Date of issuance: June 7, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 336 for Unit No. 1 and 318 for Unit No. 2. A 
publicly-available version is in ADAMS under Accession No. ML17131A277; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-58 and DPR-74: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 2016 (81 FR 
87972). The supplemental letter dated February 27, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 7, 2017.
    No significant hazards consideration comments received: No.

Northern States Power Company--Minnesota (NSPM), Docket No. 50-263, 
Monticello Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: July 28, 2016.
    Brief description of amendment: The amendment adopts TSTF-545, 
Revision 3, ``TS [technical specification] Inservice Testing Program 
Removal & Clarify SR [surveillance requirements] Usage Rule Application 
to Section 5.5 Testing.''
    Date of issuance: June 16, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 194. A publicly-available version is in ADAMS under 
Accession No. ML17123A321; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-22: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 11, 2016 (81 FR 
70181).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 2017.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company and South Carolina Public Service 
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear 
Station, Units 2 and 3, Fairfield County, South Carolina

    Date of amendment request: October 9, 2015, as supplemented on 
December 1, 2015, August 11, 2016, and December 21, 2016.
    Description of amendment: This amendment revises License Condition 
(LC) 2.D(12)(c)1. related to initial Emergency Action Levels (EALs). 
The LC will require the licensee to submit a fully-developed set of 
EALs before initial fuel load in accordance with the criteria defined 
in this license amendment.
    Date of issuance: April 10, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment Nos.: 68 (Unit 2) and 68 (Unit 3). A publicly-available 
version is in ADAMS under Accession Package No. ML16214A135; documents 
related to this amendment are listed in the Safety Evaluation enclosed 
with the amendment.
    Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendment 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: January 19, 2016 (81 FR 
2919). The supplemental letters dated December 1, 2015, August 11, 
2016, and December 21, 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 amendment is contained 
in the Safety Evaluation dated April 10, 2017.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, 
Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield, 
South Carolina

    Date of amendment request: January 20, 2017, and supplemented by 
letter dated March 8, 2017.
    Description of amendment: The amendment consists of changes to the 
VCSNS Units 2 and 3 Updated Final Safety Analysis Report (UFSAR) in the 
form of departures from the incorporated plant specific Design Control 
Document Tier 2 information. Specifically, the amendment consists of 
changes to the UFSAR to provide clarification of the interface criteria 
for nonsafety-related instrumentation that monitors safety-related 
fluid systems.
    Date of issuance: May 31, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 74. A publicly-available version is in ADAMS under 
Accession No. ML17130A903; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendment 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: February 28, 2017 (82 
FR 12130). The supplemental letter dated March 8, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application request as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in the Safety Evaluation dated May 31, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 15, 2016, as supplemented by 
letters dated August 19, 2016, August 26, 2016, September 13, 2016, 
December 16, 2016, and March 17, 2017.
    Description of amendment: The amendment authorizes changes to the 
VEGP Units 3 and 4 Updated Final Safety Analysis Report (UFSAR) in the 
form of departures from the incorporated plant-specific Design Control 
Document Tier 2 information and involves related changes to the 
associated plant-specific Tier 2* information. Specifically, the 
departures

[[Page 31106]]

consist of changes to UFSAR text and tables, and information 
incorporated by reference into the UFSAR related to updates to WCAP-
16096, ``Software Program Manual for Common Q\TM\ Systems,'' and WCAP-
16097, ``Common Qualified Platform Topical Report.''
    Date of issuance: June 8, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 79 (Unit 3) and 78 (Unit 4). A publicly-available 
version is in ADAMS under Accession No. ML17104A109; documents related 
to this amendment are listed in the Safety Evaluation enclosed with the 
amendment.
    Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment 
revised the Facility Combined License.
    Date of initial notice in Federal Register: April 12, 2016 (81 FR 
21602). The supplemental letters dated August 19, 2016, August 26, 
2016, September 13, 2016, December 16, 2016, and March 17, 2017, 
provided additional information that clarified the application, did not 
expand the scope of the application request as noticed on February 15, 
2016, and did not change the staff's proposed no significant hazards 
consideration determination as published in the Federal Register on 
April 12, 2016.
    The Commission's related evaluation of the amendment is contained 
in the Safety Evaluation dated June 8, 2017.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 23rd day of June 2017.

    For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2017-13804 Filed 7-3-17; 8:45 am]
 BILLING CODE 7590-01-P