Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 23615-23632 [2017-10570]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 82, Number 98 (Tuesday, May 23, 2017)]
[Notices]
[Pages 23615-23632]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-10570]


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NUCLEAR REGULATORY COMMISSION

[NRC-2017-0120]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, 
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is 
publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from April 25, 2017, to May 8, 2017. The last 
biweekly notice was published on May 9, 2017.

DATES: Comments must be filed by June 22, 2017. A request for a hearing 
must be filed by July 24, 2017.

ADDRESSES: You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0120. 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: T-8-D36M, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

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

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2017-0120, 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-2017-0120.
     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 White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2017-0120, 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

[[Page 23616]]

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 after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of the amendment unless the Commission finds an imminent 
danger to the health or safety of the public, in which case it will 
issue an appropriate order or rule under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission by July 
24, 2017. The petition must be filed in accordance with the filing 
instructions in the ``Electronic Submissions (E-Filing)'' section of 
this document, and should meet the requirements for petitions set forth 
in this section, except that under 10 CFR 2.309(h)(2) a State, local 
governmental body, or federally recognized Indian Tribe, or agency 
thereof does not need to address the standing requirements in 10 CFR

[[Page 23617]]

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

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate under 10 CFR 2.315(c), must be filed in accordance with the 
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 
46562, August 3, 2012). The E-Filing process requires participants to 
submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC's Web site at 
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 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

[[Page 23618]]

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

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station (CNS), Units 1 and 2, York County, South Carolina

    Date of amendment request: December 15, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16350A422.
    Description of amendment request: The amendments would modify 
Technical Specification (TS) 3.9.4, ``Residual Heat Removal (RHR) and 
Coolant Circulation--High Water Level,'' and TS 3.9.5, ``Residual Heat 
Removal (RHR) and Coolant Circulation--Low Water Level.'' Condition A 
of TS 3.9.4 applies when RHR requirements are not met, and includes 
four required actions. Required Action A.4 requires, within 4 hours, 
the closure of all containment penetrations providing direct access 
from containment atmosphere to outside atmosphere. The proposed changes 
revise Required Action A.4 and add new Required Actions A.5, A.6.1, and 
A.6.2 to clarify that the intent of the required actions is to 
establish containment closure. Each of these required actions will have 
a completion time of 4 hours. Condition B of TS 3.9.5 applies when no 
RHR loop is in operation, and includes three required actions. Required 
Action B.3 requires the closure of all containment penetrations 
providing direct access from containment atmosphere to outside 
atmosphere. The proposed changes are the same as the proposed changes 
to TS 3.9.4, consisting of a revision to Required Action B.3 and the 
addition of new Required Actions B.4, B.5.1, and B.5.2. These proposed 
changes are consistent with Technical Specification Task Force (TSTF) 
Traveler TSTF-197-A, Revision 2, ``Require Containment Closure When 
Shutdown Cooling Requirements Are Not Met.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the CNS TS to ensure that the 
appropriate actions are taken to establish containment closure in 
the event that Residual Heat Removal requirements are not met during 
refueling operations. Containment closure would be appropriate for 
mitigation of a loss of shutdown cooling accident, but it does not 
affect the initiation of the accident. The containment purge system 
isolation valves will be capable of being closed automatically on a 
high containment radiation signal, such that there will be no 
significant increase in the radiological consequences of a loss of 
shutdown cooling.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration to the 
plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The containment purge system isolation valves will remain 
capable of being closed automatically on a high containment 
radiation signal.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    Currently the Technical Specifications are vague and overly 
restrictive concerning the requirement for containment closure when 
shutdown cooling is lost. The proposed changes eliminate unclear 
requirements and provide a clear way to establish containment 
closure that meets the [TS] Bases description, which is to prevent 
radioactive gas from being released from the containment during a 
loss of shutdown cooling incident. The containment purge system 
isolation valves will remain capable of being closed automatically 
on a high containment radiation signal.
    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: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: January 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17025A069.
    Description of amendment request: The amendments would modify 
Technical Specification 3.1.2, ``Core Reactivity,'' to revise the 
Completion Times of Required Action A.1 and A.2 from 72 hours to 7 
days. This proposed change is consistent with Technical Specification 
Task Force (TSTF) Traveler TSTF-142-A, Revision 0, ``Increase the 
Completion Time when the Core Reactivity Balance is Not Within Limit.''
    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 extend the Completion Time to take the 
Required Actions when measured core reactivity is not within the 
specified limit of the predicted values. The Completion Time to 
respond to a difference between predicted and measured core 
reactivity is not an initiator to any accident previously evaluated. 
The radiological consequences of an accident during the proposed 
Completion Time are no different from the consequences of an 
accident during the existing Completion Time.
    Therefore, the proposed changes do not involved a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration to the 
plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The proposed changes provide additional time to investigate and 
to implement appropriate operating restrictions when

[[Page 23619]]

measured core reactivity is not within the specified limit of the 
predicted values. The additional time will not have a significant 
effect on plant safety due to the conservatisms used in designing 
the reactor core and performing the safety analyses, and the low 
probability of an accident or transient which would approach the 
core design limits during the additional time.
    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: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Corporation, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station (MNS), Units 1 and 2, Mecklenburg County, North 
Carolina

    Date of amendment request: January 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17025A069.
    Description of amendment request: The amendments would modify 
Technical Specification (TS) 3.9.5, ``Residual Heat Removal (RHR) and 
Coolant Circulation--High Water Level,'' and TS 3.9.6, ``Residual Heat 
Removal (RHR) and Coolant Circulation--Low Water Level.'' Condition A 
of TS 3.9.5 applies when RHR requirements are not met, and includes 
four required actions. Required Action A.4 requires, within 4 hours, 
the closure of all containment penetrations providing direct access 
from containment atmosphere to outside atmosphere. The proposed changes 
revise Required Action A.4 and add new Required Actions A.5, A.6.1, and 
A.6.2 to clarify that the intent of the required actions is to 
establish containment closure. Each of these required actions will have 
a completion time of 4 hours. Condition B of TS 3.9.6 applies when no 
RHR loop is in operation, and includes three required actions. Required 
Action B.3 requires the closure of all containment penetrations 
providing direct access from containment atmosphere to outside 
atmosphere. The proposed changes are the same as the proposed changes 
to TS 3.9.5, consisting of a revision to Required Action B.3 and the 
addition of new Required Actions B.4, B.5.1, and B.5.2. These proposed 
changes are consistent with Technical Specification Task Force (TSTF) 
Traveler TSTF-197-A, Revision 2, ``Require Containment Closure When 
Shutdown Cooling Requirements Are Not Met.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the MNS TS to ensure that the 
appropriate actions are taken to establish containment closure in 
the event that Residual Heat Removal requirements are not met during 
refueling operations. Containment closure would be appropriate for 
mitigation of a loss of shutdown cooling accident, but it does not 
affect the initiation of the accident. The containment purge system 
isolation valves will be capable of being closed automatically on a 
high containment radiation signal, such that there will be no 
significant increase in the radiological consequences of a loss of 
shutdown cooling.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration to the 
plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The containment purge system isolation valves will remain 
capable of being closed automatically on a high containment 
radiation signal.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    Currently the Technical Specifications are vague and overly 
restrictive concerning the requirement for containment closure when 
shutdown cooling is lost. The proposed changes eliminate unclear 
requirements and provide a clear way to establish containment 
closure that meets the [TS] Bases description, which is to prevent 
radioactive gas from being released from the containment during a 
loss of shutdown cooling incident. The containment purge system 
isolation valves will remain capable of being closed automatically 
on a high containment radiation signal.
    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: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station (MNS), Units 1 and 2, Mecklenburg County, North 
Carolina

    Date of amendment request: January 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17025A069.
    Description of amendment request: The amendments would modify 
Technical Specification (TS) 3.6.3, ``Containment Isolation Valves,'' 
to add a Note to TS Limiting Condition for Operation (LCO) 3.6.3 
Required Actions A.2, C.2 and E.2 to allow isolation devices that are 
locked, sealed or otherwise secured to be verified by use of 
administrative means. This proposed change is consistent with Technical 
Specification Task Force (TSTF) Traveler TSTF-269-A, Revision 2, 
``Allow Administrative Means of Position Verification for Locked or 
Sealed Valves.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes modify MNS TS 3.6.3, ``Containment 
Isolation Valves''. This TS currently includes actions that require 
penetrations to be isolated and periodically verified to be 
isolated. A Note is proposed to be added to TS 3.6.3 Required 
Actions A.2, C.2, and E.2, to allow isolation devices that are 
locked, sealed, or otherwise secured to be verified by use of 
administrative means. The proposed changes do not affect any plant 
equipment, test methods, or plant operation, and is not an initiator 
of any analyzed accident sequence. The inoperable containment 
penetrations will continue to be isolated, and hence perform their 
isolation

[[Page 23620]]

function. Operation in accordance with the proposed TSs will ensure 
that all analyzed accidents will continue to be mitigated as 
previously analyzed.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration to the 
plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The proposed changes will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. Affected containment penetrations will continue to be 
isolated as required by the existing TS.
    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: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: January 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17025A069.
    Description of amendment request: The amendments would modify 
Technical Specification (TS) 3.8.1, ``AC [Alternating Current] 
Sources--Operating,'' to allow greater flexibility in performing 
Surveillance Requirements (SRs) by modifying Mode restriction notes in 
TS SRs 3.8.1.8, 3.8.1.11, 3.8.1.16, 3.8.1.17, and 3.8.1.19. This 
proposed change is consistent with Technical Specification Task Force 
(TSTF) Traveler TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode 
Restriction Notes.''
    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 modify Mode restriction Notes in TS SRs 
3.8.1.8, 3.8.1.11, 3.8.1.16, 3.8.1.17, and 3.8.1.19 to allow 
performance of the Surveillance in whole or in part to reestablish 
Diesel Generator (DG) Operability, and to allow the crediting of 
unplanned events that satisfy the Surveillance(s) [Requirements]. 
The emergency diesel generators and their associated emergency loads 
are accident mitigating features, and are not an initiator of any 
accident previously evaluated. As a result, the probability of any 
accident previously evaluated is not significantly increased. To 
manage any increase in risk, the proposed changes require an 
assessment to verify that plant safety will be maintained or 
enhanced by performance of the Surveillance in the current 
prohibited Modes. The radiological consequences of an accident 
previously evaluated during the period that the DG is being tested 
to reestablish operability are no different from the radiological 
consequences of an accident previously evaluated while the DG is 
inoperable. As a result, the consequences of any accident previously 
evaluated are not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration to the 
plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The purpose of Surveillances is to verify that equipment is 
capable of performing its assumed safety function. The proposed 
changes will only allow the performance of the Surveillances to 
reestablish operability, and the proposed changes may not be used to 
remove a DG from service. In addition, the proposed changes will 
potentially shorten the time that a DG is unavailable because 
testing to reestablish operability can be performed without a plant 
shutdown. The proposed changes also require an assessment to verify 
that plant safety will be maintained or enhanced by performance of 
the Surveillance in the current prohibited Modes.
    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: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: January 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17025A069.
    Description of amendment request: The amendments would modify 
Technical Specification (TS) 3.4.12, ``Low Temperature Overpressure 
Protection (LTOP) System,'' to increase the time allowed for swapping 
charging pumps to 1 hour. Additionally, an existing note in the 
Applicability section of TS 3.4.12 is being reworded and relocated to 
the Limiting Condition for Operation section of TS 3.4.12 as Note 2. 
These proposed changes are consistent with Technical Specification Task 
Force (TSTF) Traveler TSTF-285-A, Revision 1, ``Charging Pump Swap LTOP 
Allowance.''
    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 increase the time allowed for swapping 
charging pumps from 15 minutes to one hour, and make several other 
associated administrative changes and clarifications to the TS. 
These changes do not affect event initiators or precursors. Thus, 
the proposed changes do not involve a

[[Page 23621]]

significant increase in the probability of an accident previously 
evaluated. In addition, the proposed changes do not alter any 
assumptions previously made in the radiological consequence 
evaluations nor affect mitigation of the radiological consequences 
of an accident described in the Updated Final Safety Analysis Report 
(UFSAR). As such, the consequences of accidents previously evaluated 
in the UFSAR will not be increased and no additional radiological 
source terms are generated. Therefore, there will be no reduction in 
the capability of those structures, systems, and components (SSCs) 
in limiting the radiological consequences of previously evaluated 
accidents, and reasonable assurance that there is no undue risk to 
the health and safety of the public will continue to be provided. 
Thus, the proposed changes do not involve a significant increase in 
the 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve physical changes to analyzed 
SSCs or changes to the modes of plant operation defined in the 
technical specification. The proposed changes do not involve the 
addition or modification of plant equipment (no new or different 
type of equipment will be installed) nor do they alter the design or 
operation of any plant systems. No new accident scenarios, accident 
or transient initiators or precursors, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
changes. The proposed changes do not cause the malfunction of 
safety-related equipment assumed to be operable in accident 
analyses. No new or different mode of failure has been created and 
no new or different equipment performance requirements are imposed 
for accident mitigation. As such, the proposed changes have no 
effect on previously evaluated accidents.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The proposed changes do not adversely affect any current plant 
safety margins or the reliability of the equipment assumed in the 
safety analysis. Therefore, there are no changes being made to any 
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: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: January 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17025A069.
    Description of amendment request: The amendments would modify 
Technical Specification (TS) 3.1.8, ``PHYSICS TESTS Exceptions,'' to 
allow the numbers of channels required by the Limiting Condition for 
Operation (LCO) section of TS 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' to be reduced from ``4'' to ``3'' to allow one 
nuclear instrumentation channel to be used as an input to the 
reactivity computer for physics testing without placing the nuclear 
instrumentation channel in a tripped condition. This proposed change is 
consistent with Technical Specification Task Force (TSTF) Traveler 
TSTF-315-A, Revision 0, ``Reduce Plant Trips Due to Spurious Signals to 
the Nuclear Instrumentation System (NIS) During Physics 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 consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise TS 3.1.8, ``PHYSICS TESTS 
Exceptions,'' to allow the number of channels required by LCO 3.3.1, 
``RTS Instrumentation,'' to be reduced from ``4'' to ``3'', to allow 
one nuclear instrumentation channel to be used as an input to the 
reactivity computer for physics testing without placing the nuclear 
instrumentation channel in a tripped condition. A reduction in the 
number of required nuclear instrumentation channels is not an 
initiator to any accident previously evaluated. With the nuclear 
instrumentation channel placed in bypass instead of in trip, reactor 
protection is still provided by the nuclear instrumentation system 
operating in a two-out-of-three channel logic. As a result, the 
ability to mitigate any accident previously evaluated is not 
significantly affected. The proposed changes will not affect the 
source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of any 
accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration to the 
plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The proposed changes reduce the probability of a spurious 
reactor trip during physics testing. The reactor trip system 
continues to be capable of protecting the reactor utilizing the 
power range neutron flux trips operating in a two-out-of-three trip 
logic. As a result, the reactor is protected and the probability of 
a spurious reactor trip is significantly reduced.
    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: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: January 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17025A069.
    Description of amendment request: The amendments would modify 
Technical Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW) 
System,'' to expand the TS 3.7.5 Limiting

[[Page 23622]]

Condition for Operation, Condition A to include the situation when one 
turbine driven AFW pump is operable in MODE 3, immediately following a 
refueling outage (if MODE 2 has not been entered), with a 7-day 
Completion Time. This proposed change is consistent with Technical 
Specification Task Force (TSTF) Traveler TSTF-340-A, Revision 3, 
``Allow 7 Day Completion Time for a Turbine-Driven AFW Pump 
Inoperable.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise TS 3.7.5, ``Auxiliary Feedwater 
(AFW) System,'' to allow a 7 day Completion Time to restore an 
inoperable AFW turbine-driven pump in MODE 3 immediately following a 
refueling outage, if MODE 2 has not been entered. An inoperable AFW 
turbine-driven pump is not an initiator of any accident previously 
evaluated. The ability of the plant to mitigate an accident is no 
different while in the extended Completion Time than during the 
existing Completion Time. The proposed changes will not affect the 
source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of any 
accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration to the 
plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The proposed changes revise TS 3.7.5, ``Auxiliary Feedwater 
(AFW) System,'' to allow a 7 day Completion Time to restore an 
inoperable turbine-driven AFW pump in Mode 3, immediately following 
a refueling outage, if Mode 2 has not been entered. In Mode 3 
immediately following a refueling outage, core decay heat is low and 
the need for AFW is also diminished. The two operable motor driven 
AFW pumps are available and there are alternate means of decay heat 
removal if needed. As a result, the risk presented by the extended 
Completion Time is minimal.
    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: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station (MNS), Units 1 and 2, Mecklenburg County, North 
Carolina

    Date of amendment request: January 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17025A069.
    Description of amendment request: The amendments would modify 
Technical Specification (TS) 3.4.10, ``Pressurizer Safety Valves,'' TS 
3.4.12, ``Low Temperature Overpressure Protection (LTOP) System,'' and 
TS 3.7.4, ``Steam Generator Power Operated Relief Valves (SG PORVs),'' 
to revise the Completion Times for Limiting Condition for Operation 
(LCO) 3.4.10 Required Action B.2, and LCO 3.7.4 Required Action C.2 
from 12 to 24 hours and LCO 3.4.12 Required Action G.1 from 8 to 12 
hours. The proposed changes are consistent with Technical Specification 
Task Force (TSTF) Traveler TSTF-352-A, Revision 1, ``Provide Consistent 
Completion Time to Reach MODE 4.''
    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 allow a more reasonable time to plan and 
execute required actions, and will 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 will not 
alter or prevent the ability of structures, systems, and components 
(SSCs) from performing their intended functions to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed changes do not physically alter safety-related 
systems nor affect the way in which safety-related systems perform 
their functions. All accident analysis acceptance criteria will 
continue to be met with the proposed changes. The proposed changes 
will not affect the source term, containment isolation, or 
radiological release assumptions used in evaluating the radiological 
consequences of an accident previously evaluated. The proposed 
changes will not alter any assumptions or change any mitigation 
actions in the radiological consequence evaluations in the MNS 
Updated Final Safety Analysis Report (UFSAR). The applicable 
radiological dose acceptance criteria will continue to be met.
    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.
    There are no proposed design changes nor are there any changes 
in the method by which any safety-related plant SSC performs its 
safety function. The proposed changes will not affect the normal 
method of plant operation or change any operating parameters. No 
equipment performance requirements will be affected. The proposed 
changes will not alter any assumptions made in the safety analyses.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of this amendment. There will be no adverse effect or 
challenges imposed on any safety-related system as a result of this 
amendment.
    Therefore, the proposed changes do not create the possibility of 
a new or different accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their intended functions. 
These barriers include the fuel cladding, the reactor coolant system 
pressure boundary, and the containment barriers. The proposed 
changes will not have any impact on these barriers. No accident 
mitigating equipment will be adversely impacted.
    Therefore, existing safety margins will be preserved. None of 
the proposed changes will 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

[[Page 23623]]

amendment request involves no significant hazards consideration.
    Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station (MNS), Units 1 and 2, Mecklenburg County, North 
Carolina

    Date of amendment request: January 11, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17025A069.
    Description of amendment request: The amendments would modify 
Technical Specification (TS) 3.9.6, ``Residual Heat Removal (RHR) and 
Coolant Circulation--Low Water Level,'' to add Note 1 to the Limiting 
Condition for Operation (LCO) Section of TS 3.9.6 to allow the securing 
of the operating train of RHR for up to 15 minutes to support switching 
operating trains. The allowance is restricted to three conditions: (a) 
The core outlet temperature is maintained greater than 10 degrees 
Fahrenheit below saturation temperature; (b) no operations are 
permitted that would cause an introduction of coolant into the Reactor 
Coolant System (RCS) with boron concentration less than that required 
to meet the minimum required boron concentration of LCO 3.9.1; and (c) 
no draining operations to further reduce RCS water volume are 
permitted. Additionally, the amendments would modify the LCO Section of 
TS 3.9.6 to add Note 2 which would allow one required RHR loop to be 
inoperable for up to 2 hours for surveillance testing, provided that 
the other RHR loop is operable and in operation. These proposed changes 
are consistent with Technical Specification Task Force (TSTF) Travelers 
TSTF-349-A, Revision 1, ``Add Note to LCO 3.9.5 Allowing Shutdown 
Cooling Loops Removal from Operation,'' TSTF-361-A, Revision 2, ``Allow 
Standby SDC [Shutdown Cooling]/RHR/DHR [Decay Heat Removal] Loop to be 
Inoperable to Support Testing,'' and TSTF-438-A, Revision 0, ``Clarify 
Exception Notes to be Consistent with the Requirement Being Excepted.''
    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 add two notes to MNS TS LCO 3.9.6. Note 1 
would allow securing the operating train of Residual Heat Removal 
(RHR) for up to 15 minutes to support switching operating trains, 
subject to certain restrictions. Note 2 to would allow one RHR loop 
to be inoperable for up to 2 hours for surveillance testing provided 
the other RHR loop is Operable and in operation. These provisions 
are operational allowances. Neither operational allowance is an 
initiator to any accident previously evaluated. In addition, the 
proposed changes will not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of any accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not involve a physical alteration to the 
plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the assumptions made in the 
safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    An operational allowance is proposed which would allow securing 
the operating train of RHR for up to 15 minutes to support switching 
operating trains, subject to certain restrictions. Considering these 
restrictions, combined with the short time frame allowed to swap 
operating RHR trains, and the ability to start an operating RHR 
train, if needed, the occurrence of an event that would require 
immediate operation of an RHR train is extremely remote.
    An operational allowance is also proposed which would allow one 
RHR loop to be inoperable for up to 2 hours for surveillance testing 
provided the other RHR loop is operable and in operation. A similar 
allowance currently appears in MNS TS 3.4.7, ``Reactor Coolant 
System (RCS) Loops--MODE 5, Loops Filled,'' and MNS TS 3.4.8, ``RCS 
Loops--MODE 5, Loops Not Filled,'' and the conditions under which 
the operational allowance would be applied in TS 3.9.6 are not 
significantly different from those specifications. This operational 
allowance provides the flexibility to perform surveillance testing, 
while ensuring that there is reasonable time for operators to 
respond to and mitigate any expected failures. The purpose of the 
RHR System is to remove decay and sensible heat from the Reactor 
Coolant System, to provide mixing of borated coolant, and to prevent 
boron stratification. Removal of system components from service as 
described above, and with limitations in place to maintain the 
ability of the RHR System to perform its safety function, does not 
significantly impact the margin of safety. Operators will continue 
to have adequate time to respond to any off-normal events. Removing 
the system from service, for a limited period of time, with other 
operational restrictions, limits the consequences to those already 
assumed in the Updated Final Safety Analysis Report (UFSAR).
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: March 30, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17089A380.
    Description of amendment request: The proposed amendment would 
revise the PNP Cyber Security Plan (CSP) Milestone 8 full 
implementation date from December 15, 2017, to May 31, 2020. This 
amendment request is in support of PNP's transition, starting on 
October 1, 2018, from an operating power plant to a decommissioned 
plant.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the CSP implementation schedule is 
administrative in nature. This change 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

[[Page 23624]]

plant modifications which affect the performance capability of the 
structures, system, and components 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 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 CSP implementation schedule is 
administrative in nature. This proposed change 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 
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 accident 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, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed changes to 
the CSP implementation schedule is administrative in nature. In 
addition, the milestone date delay for full implementation of the 
CSP has no substantive impact because other measures have been taken 
which provide adequate protection during this period of time. 
Because there is no change to established safety margins as a result 
of this change, the proposed change does not involve a significant 
reduction in a margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeanne Cho, Senior Counsel, Entergy 
Services, Inc., 440 Hamilton Ave., White Plains, NY 10601.
    NRC Branch Chief: David J. Wrona.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: March 30, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17101A608.
    Description of amendment request: The amendment would revise the 
renewed facility operating license Paragraph 3.G, ``Physical 
Protection.'' The amendment would revise the Pilgrim Nuclear Power 
Station Cyber Security Plan (CSP) implementation schedule for Milestone 
8 full implementation date from December 15, 2017, to December 31, 
2020.
    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 the CSP implementation schedule is 
administrative in nature. The change 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 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 change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the CSP implementation schedule is 
administrative in nature. The proposed change 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 
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, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed change to 
the CSP implementation schedule is administrative in nature. In 
addition, the milestone date delay for full implementation of the 
CSP has no substantive impact because other measures have been taken 
which provide adequate protection during this period of time. 
Because there is no change to established safety margins as a result 
of this change, the proposed change does not involve a significant 
reduction in a margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Douglas A. Broaddus.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc. 
(the licensees), Docket Nos. 50-416 and 72-50, Grand Gulf Nuclear 
Station, Unit 1 (Grand Gulf), and Independent Spent Fuel Storage 
Installation (ISFSI), Claiborne County, Mississippi

    Date of amendment request: March 29, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17093A729.
    Description of amendment request: The proposed amendment would make 
an administrative change to the name of South Mississippi Electric 
Power Association, one of the licensees for Grand Gulf and its ISFSI. 
Effective November 10, 2016, South Mississippi Electric Power 
Association changed its corporate name from ``South Mississippi 
Electric Power Association'' to ``Cooperative Energy, a Mississippi 
Electric Cooperative.'' The corporate name was changed for commercial 
reasons. The changes proposed herein to the Grand Gulf operating 
license solely reflects the changed licensee name. This name change is 
purely administrative in nature. This request does not involve a 
transfer of control or of an interest in the license.

[[Page 23625]]

    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. Do the proposed changes ``involve a significant increase in 
the probability or consequences of an accident previously 
evaluated''?
    Response: No.
    The proposed amendments simply change the name of a licensee. 
The name change is purely administrative. None of the functions or 
responsibility of any of the Grand Gulf licensees will change as a 
result of the amendments. The proposed amendments do not alter the 
design, function, or operation of any plant equipment. As such, the 
accident and transient analyses contained in the facility updated 
final safety analysis report will not be affected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes ``create the possibility of a new or 
different kind of accident from any accident previously evaluated''?
    Response: No.
    The proposed amendments simply change the name of a licensee. 
The proposed name change is purely administrative. None of the 
functions or responsibility of any of the Grand Gulf licensees will 
change as a result of the amendments. The proposed amendments do not 
alter the design, function, or operation of any plant equipment. As 
such, the accident and transient analyses contained in the facility 
updated final safety analysis report will not be affected.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Do the proposed changes ``involve a significant reduction in 
the margin of safety''?
    Response: No.
    The proposed amendments simply change the name of a licensee. 
The name change is purely administrative. None of the functions or 
responsibility of any of the Grand Gulf licensees will change as a 
result of the amendments. The proposed amendments do not alter the 
design, function, or operation of any plant equipment. As such, the 
accident and transient analyses contained in the facility updated 
final safety analysis report will not be affected.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William B. Glew, Jr., Associate General 
Counsel--Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New 
York 10601.
    NRC Branch Chief: Robert J. Pascarelli.

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 (GGNS), Claiborne 
County, Mississippi

    Date of amendment request: December 29, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16364A338.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) for GGNS. The amendment would 
allow for a one cycle extension to the 10-year frequency of the GGNS 
containment integrated leakage rate test (ILRT) or Type A test and the 
drywell bypass leak rate test (DWBT). These tests are required by TS 
5.5.12, ``10 CFR part 50, Appendix J [Primary Reactor Containment 
Leakage Testing for Water-Cooled Power Reactors], Testing Program,'' 
and TS Surveillance Requirement 3.6.5.1.1, respectively. The proposed 
change would permit the existing ILRT and DWBT frequency to be extended 
from 10 years to 11.5 years.
    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 edits in [brackets]:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment to the Technical Specifications (TS) 
involves the extension of the Grand Gulf Nuclear Station, Unit 1 
(GGNS) Type A integrated leakage rate test and the drywell bypass 
leakage rate test intervals to 11.5 years.
    The proposed extension does not involve either a physical change 
to the plant or a change in the manner in which the plant is 
operated or controlled. The containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. Type B 
and C testing ensures that individual containment isolation valves 
are essentially leak tight. In addition, aggregate Type B and C 
leakage rates support the leakage tightness of primary containment 
by minimizing potential leakage paths. The assessment of the [leak-
tightness] of the drywell will continue to be performed at least 
once each operating cycle. The proposed amendment will not change 
the leakage rate acceptance requirements. As such, the containment 
will continue to perform its design function as a barrier to fission 
product releases. In addition, the containment and the testing 
requirements invoked to periodically demonstrate the integrity of 
the containment and the assessment of the [leak-tightness] of the 
drywell exist to ensure the plant's ability to mitigate the 
consequences of an accident, and do not involve the prevention or 
identification of any precursors of an accident. Therefore, this 
proposed extension does not involve a significant increase in the 
probability of an accident previously evaluated.
    Therefore, the proposed change does not result in 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 amendment to the Technical Specifications (TS) 
involves the extension of the Grand Gulf Nuclear Station, Unit 1 
(GGNS) Type A integrated leakage rate test and the drywell bypass 
leakage rate test intervals to 11.5 years. The containment and the 
testing requirements to periodically demonstrate the integrity of 
the containment exist to ensure the plant's ability to mitigate the 
consequences of an accident do not involve any accident precursors 
or initiators. The proposed change does not involve a physical 
change to the plant (i.e., no new or different type of equipment 
will be installed) or a change to the manner in which the plant is 
operated or controlled.
    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 
the margin of safety?
    Response: No.
    The proposed amendment to the Technical Specifications (TS) 
involves the extension of the Grand Gulf Nuclear Station, Unit 1 
(GGNS) Type A integrated leakage rate test and the drywell bypass 
leakage rate test intervals to 11.5 years. This amendment does not 
alter the manner in which safety limits, limiting safety system set 
points, or limiting conditions for operation are determined. The 
specific requirements and conditions of the TS 10 CFR part 50, 
Appendix J, Testing Program for containment leak rate testing exist 
to ensure that the degree of containment structural integrity and 
leak-tightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by TS 
is maintained.
    The proposed change involves the extension of the interval for 
only the Type A containment leakage rate test and the drywell bypass 
leakage rate test for GGNS. The proposed surveillance interval 
extension is bounded by the 15-year Type A test interval currently 
authorized within NEI 94-01, Revision 3-A. The design, operation, 
testing methods, and acceptance criteria for Types A, B, and C 
containment leakage tests specified in applicable codes and 
standards would continue to be met with the

[[Page 23626]]

acceptance of this proposed change, since these are not affected by 
the proposed changes to the Type A test interval. In addition to the 
scheduled performance of DWBT GGNS will continue to monitor the 
drywell for significant leakage during operation.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William B. Glew, Jr., Associate General 
Counsel--Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New 
York 10601.
    NRC Branch Chief: Robert J. Pascarelli.

Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear 
Generating Station (OCNGS), Ocean County, New Jersey

    Date of amendment request: April 10, 2017. A publicly-available 
version is available in ADAMS under Accession No. ML17100A844.
    Description of amendment request: The amendment would revise the 
OCNGS Cyber Security Plan (CSP) Milestone 8 (MS8) full implementation 
completion date, as set forth in the CSP implementation schedule, and 
revise the physical protection license condition in the renewed 
facility operating license. The licensee proposes to revise the CSP MS8 
completion date from December 31, 2017, to August 31, 2021.
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review 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 amendment request proposes a change to the OCNGS CSP MS8 
completion date as set forth in the CSP implementation schedule and 
associated regulatory commitments. The NRC staff has concluded that 
the proposed change: (1) 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; (2) does not require any plant 
modifications which affect the performance capability of the 
structures, systems, and components relied upon to mitigate the 
consequences of postulated accidents; and (3) has no impact on the 
probability or consequences of an accident previously evaluated. In 
addition, the NRC staff has concluded that the proposed change to 
the CSP implementation schedule is administrative in nature.
    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 NRC staff has concluded the proposed change: (1) 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; and (2) does 
not require any plant modifications which affect the performance 
capability of the structures, systems, and components 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. In addition, the NRC staff has 
concluded that the proposed change to the OCNGS CSP MS8 
implementation schedule is administrative in nature.
    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.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The delay of the full 
implementation date for the OCNGS CSP MS8 has no substantive impact 
because other measures have been taken which provide adequate 
protection for the plant during this period of time. Therefore, the 
NRC staff has concluded that there is no significant reduction in a 
margin of safety. In addition, the NRC staff has concluded that the 
proposed change to the OCNGS CSP MS8 implementation schedule is 
administrative in nature.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: March 24, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17087A012.
    Description of amendment request: The proposed changes would modify 
Technical Specifications (TS) Section 3.7.2, ``Steam Generator Stop 
Valves (SGSVs),'' to incorporate the SGSV actuator trains into the 
Limiting Condition for Operation and provide associated Conditions and 
Required Actions. In addition, Surveillance Requirement (SR) 3.7.2.2 
would be revised to clearly identify that the SGSV actuator trains are 
required to be tested in accordance with the SR.
    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 provide requirements for SGSVs that have 
dual actuators which receive signals from separate instrumentation 
trains. The design and functional performance requirements, 
operational characteristics, and reliability of the SGSVs and 
actuator trains are unchanged. There is no impact on the design 
safety function of the SGSVs to close (as an accident mitigator), 
nor is there any change with respect to inadvertent closure of an 
SGSV (as a potential transient initiator). Since no failure mode or 
initiating condition that could cause an accident (including any 
plant transient) is created or affected, the change cannot involve a 
significant increase in the probability of an accident previously 
evaluated.
    With regard to the consequences of an accident and the equipment 
required for mitigation of the accident, the proposed changes 
involve no design or physical changes to the SGSVs or any other 
equipment required for accident mitigation. With respect to SGSV 
actuator train Completion Times, the consequences of an accident are 
independent of equipment Completion Times as long as adequate 
equipment availability is maintained. The proposed SGSV actuator 
Completion Times take into account the redundancy of the actuator 
trains and are limited in extent consistent with other Completion 
Times specified in the TS. Adequate equipment availability would 
therefore continue to be required by the TS. On this basis, the 
consequences of applicable, analyzed accidents are not significantly 
affected by the proposed changes.

[[Page 23627]]

    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 incorporate requirements for the SGSV 
actuator trains in TS 3.7.2 do not involve any design or physical 
changes to the facility, including the SGSVs and actuator trains 
themselves. No physical alteration of the plant is involved, as no 
new or different type of equipment is to be installed. The proposed 
changes do not alter any assumptions made in the safety analyses, 
nor do they involve any changes to plant procedures for ensuring 
that the plant is operated within analyzed limits. As such, no new 
failure modes or mechanisms that could cause a new or different kind 
of accident from any previously evaluated are being introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to incorporate requirements for the SGSV 
actuator trains do not alter the manner in which safety limits or 
limiting safety system settings are determined. No changes to 
instrument/system actuation setpoints are involved. The safety 
analysis acceptance criteria are not affected by this change and the 
proposed changes will not permit plant operation in a configuration 
outside the design basis.
    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: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: David J. Wrona.

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

    Date of amendment request: March 24, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17086A442.
    Description of amendment request: The proposed change would 
relocate cycle specific minimum critical power ratio (MCPR) values to 
the DAEC core operating limits report (COLR). The proposed amendment 
would revise the DAEC technical specifications (TS) to modify TS Table 
3.3.2.1-1, ``Control Rod Block Instrumentation,'' Footnotes (a) through 
(e), and would relocate cycle specific MCPR values previously specified 
in TS Table 3.3.2.1-1, Footnotes (a) through (e) to TS 5.6.5(a)(4) by 
reference to the DAEC COLR.
    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 is an administrative change that does not 
affect any plant systems, structures, or components designed for the 
prevention or mitigation of previously evaluated accidents. No new 
equipment is added nor is installed equipment being changed or 
operated in a different manner.
    Relocation of the Control Rod Block Instrumentation MCPR values 
to the COLR has no influence or impact on, nor does it contribute in 
any way to the probability or consequences of transients or 
accidents. The COLR will continue to be controlled by the NextEra 
programs and procedures that comply with TS 5.6.5. Transient 
analyses addressed in the Final Safety Analysis Report will continue 
to be performed in the same manner with respect to changes in the 
cycle-dependent parameters obtained from the use of NRC-approved 
reload design methodologies, which ensures that the transient 
evaluation of new reloads are bounded by previously accepted 
analyses.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of a previously evaluated 
accident.
    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 administrative change does not involve any changes 
to the operation, testing, or maintenance of any safety-related, or 
otherwise important to safety systems. All systems important to 
safety will continue to be operated and maintained within their 
design bases. Relocation of the Control Rod Block Instrumentation 
MCPR values to the COLR has no influence or impact on new or 
different kind of accidents.
    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 not affected by the relocation of cycle-
specific Control Rod Block Instrumentation MCPR values from the TS 
to the COLR. Appropriate measures exist to control the values of 
these cycle-specific limits since it is required by TS that only 
NRC-approved methods be used to determine the limits. The proposed 
change continues to require operation within the core thermal limits 
as obtained from NRC-approved reload design methodologies and the 
actions to be taken if a limit is exceeded remain unchanged, again, 
in accordance with existing TS.
    Therefore, the proposed change has no impact to the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William Blair, P.O. Box 14000, Juno Beach, 
FL 33408-0420.
    NRC Branch Chief: David J. Wrona.

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

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: March 30, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17093A688.
    Description of amendment request: The amendments would revise 
technical specification requirements to operate ventilation systems 
with charcoal filters from 10 hours to 15 minutes in accordance with 
Technical Specifications Task Force (TSTF) Traveler TSTF-522, Revision 
0, ``Revise Ventilation System Surveillance Requirements to Operate for 
10 hours per Month.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change replaces an existing Surveillance 
Requirement to operate the CREMAFS [Control Room Emergency Makeup 
Air and Filtration System], FSBEACS [Fuel Storage Building Emergency 
Air Cleaning System], and SBVS [Shield Building Ventilation System] 
equipped with electric heaters for at least a continuous 10-hour 
period in accordance with the SFCP [Surveillance Frequency Control 
Program] with a requirement to operate the systems for 15 continuous 
minutes with heaters operating.

[[Page 23628]]

    These systems are not accident initiators and therefore, these 
changes do not involve a significant increase in the probability of 
an accident. The proposed system and filter testing changes are 
consistent with current regulatory guidance for these systems and 
will continue to assure that these systems perform their design 
function which may include mitigating accidents. Thus, the change 
does not involve a significant increase in the consequences of an 
accident.
    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 change replaces an existing Surveillance 
Requirement to operate the CREMAFS, FSBEACS, and SBVS equipped with 
electric heaters for at least a continuous 10-hour period in 
accordance with the SFCP with a requirement to operate the systems 
for 15 continuous minutes with heaters operating.
    The change proposed for these ventilation systems does not 
change any system operations or maintenance activities. Testing 
requirements will be revised and will continue to demonstrate that 
the Limiting Conditions for Operation are met and the system 
components are capable of performing their intended safety 
functions. The change does not create new failure modes or 
mechanisms and no new accident precursors are generated.
    Therefore, it is concluded that this 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 change replaces an existing Surveillance 
Requirement to operate the CREMAFS, FSBEACS, and SBVS equipped with 
electric heaters for at least a continuous 10-hour period in 
accordance with the SFCP with a requirement to operate the systems 
for 15 continuous minutes with heaters operating.
    The design basis for the ventilation systems' heaters is to heat 
the incoming air which reduces the relative humidity. The heater 
testing change proposed will continue to demonstrate that the 
heaters are capable of heating the air and will perform their design 
function. The proposed change is consistent with regulatory 
guidance.
    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: William Blair, Managing Attorney--Nuclear, 
Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-
0420.
    NRC Branch Chief: James G. Danna.

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

    Date of amendment request: March 29, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17094A565.
    Brief description of amendment request: The proposed amendments 
would revise the current emergency action levels (EAL) scheme used at 
PINGP to the EAL scheme contained in NEI 99-01, Revision 6, 
``Development of Emergency Action Levels.''
    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 the PINGP EAL scheme does not impact the 
physical function of plant structures, systems or components (SSC) 
or the manner in which the SSCs perform their design function. The 
proposed change neither adversely affects accident initiators or 
precursors, nor alters design assumptions. Therefore, the proposed 
change does not alter or prevent the ability of SSCs to perform 
their intended function to mitigate the consequences of an event. 
The Emergency Plan, including the associated EALs, is implemented 
when an event occurs and cannot increase the probability of an 
accident. Further, the proposed change does not reduce the 
effectiveness of the Emergency Plan to meet the emergency planning 
requirements established in 10 CFR 50.47 and 10 CFR part 50, 
Appendix E.
    Therefore, the proposed EAL scheme 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 involve any physical alteration to 
the plant, that is, no new or different type of equipment will be 
installed. The proposed change also does not change the method of 
plant operation and does not alter assumptions made in the safety 
analysis. Therefore, the proposed change will not create new failure 
modes or mechanisms that could result in a new or different kind of 
accident. The emergency plan, including the associated EAL scheme, 
is implemented when an event occurs and is not an accident 
initiator.
    Therefore, the proposed EAL scheme change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is provided by the ability of accident 
mitigation SCCs to perform at their analyzed capability. The change 
proposed in this license amendment request does not modify any plant 
equipment and there is no impact to the capability of the equipment 
to perform its intended accident mitigation function. The proposed 
change does not impact operation of the plant or its response to 
transients or accidents. Additionally, the proposed changes will not 
change any criteria used to establish safety limits or any safety 
system settings. The applicable requirements of 10 CFR 50.47 and 10 
CFR part 50, Appendix E will continue to be met.
    Therefore, the proposed EAL scheme 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 requests involve no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: David J. Wrona.

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

    Date of amendment request: March 27, 2017, as supplemented by 
letter dated April 28, 2017. Publicly-available versions are in ADAMS 
under Accession Nos. ML17086A364 and ML17118A092, respectively.
    Description of amendment request: The amendment would amend the 
Hope Creek Generating Station (Hope Creek) Technical Specifications 
(TSs) to revise and relocate the pressure-temperature (P-T) limit 
curves to a licensee-controlled pressure and temperature limits report 
(PTLR). The request was submitted in accordance with guidance provided 
in NRC Generic Letter 96-03, ``Relocation of the Pressure Temperature 
Limit Curves and Low Temperature Overpressure Protections System 
Limits,'' dated January 31, 1996.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 23629]]


    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed license amendment adopts the NRC approved 
methodology described in Boiling Water Reactor Owner's Group (BWROG) 
Licensing Topical Report (LTR) (BWROG-TP-11-022-A, SIR-05-044), 
``Pressure Temperature Limits Report Methodology for Boiling Water 
Reactors.'' The Hope Creek PTLR was developed based on the 
methodology and template provided in the BWROG LTR.
    10 CFR part 50, Appendix G establishes requirements to protect 
the integrity of the reactor coolant pressure boundary (RCPB) in 
nuclear power plants.
    Implementing this NRC approved methodology does not reduce the 
ability to protect the RCPB as specified in Appendix G, nor will 
this change increase the probability of malfunction of plant 
equipment, or the failure of plant structures, systems, or 
components. Incorporation of the new methodology for calculating P-T 
curves, and the relocation of the P-T curves from the TS to the PTLR 
provides an equivalent level of assurance that the RCPB is capable 
of performing its intended safety functions.
    The proposed changes do not adversely affect accident initiators 
or precursors, and do not alter the design assumptions, conditions, 
or configuration of the plant or the manner in which the plant is 
operated and maintained. The ability of structures, systems, and 
components to perform their intended safety functions is not altered 
or prevented by the proposed changes, and the assumptions used in 
determining the radiological consequences of previously evaluated 
accidents are not affected.
    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 in methodology for calculating P-T limits and the 
relocation of those limits to the PTLR do not alter or involve any 
design basis accident initiators. RCPB integrity will continue to be 
maintained in accordance with 10 CFR part 50, Appendix G, and the 
assumed accident performance of plant structures, systems and 
components will not be affected. The proposed changes do not involve 
a physical alteration of the plant (i.e., no new or different type 
of equipment will be installed), and the installed equipment is not 
being operated in a new or different manner.
    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 changes involve a significant reduction in 
a margin of safety?
    Response: No.
    The proposed changes do not affect the function of the RCPB or 
its response during plant transients. Calculating the Hope Creek P-T 
limits using the NRC approved SI methodology ensures adequate 
margins of safety relating to RCPB integrity are maintained. The 
proposed changes do not alter the manner in which the Limiting 
Conditions for Operation P-T limits for the RCPB are determined. 
There are no changes to the setpoints at which protective actions 
are initiated, and the operability requirements for equipment 
assumed to operate for accident mitigation are not affected.
    Therefore, it is concluded that 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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: James G. Danna.

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

    Date of amendment request: April 12, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17102B032.
    Description of amendment request: The requested amendment proposes 
changes to combined license (COL) Appendix C (and plant-specific Tier 
1) and Updated Final Safety Analysis Report (UFSAR) Tier 2 that 
describe: (1) The inspection and analysis of, and specifies the maximum 
calculated flow resistance acceptance criteria for, the fourth-stage 
automatic depressurization system loops; (2) revises licensing basis 
text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 
that describes the testing of, and specifies the allowable flow 
resistance acceptance criteria for, the in-containment refueling water 
storage tank (IRWST) injection line; (3) revises licensing basis text 
in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 that 
describes the testing of, and specifies the maximum flow resistance 
acceptance criteria for, the containment recirculation line; (4) 
revises licensing basis text in COL Appendix C (and plant-specific Tier 
1) and UFSAR Tier 2 that specifies acceptance criteria for the maximum 
flow resistance between the IRWST drain line and the containment; and 
5) removes licensing basis text from UFSAR Tier 2 that discusses the 
operation of swing check valves in current operating plants.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not adversely affect the operation of 
any systems or equipment that initiate an analyzed accident or alter 
any structures, systems, and components (SSCs) accident initiator or 
initiating sequence of events. The proposed changes do not adversely 
affect the physical design and operation of the in-containment 
refueling water storage tank (IRWST) injection, drain, containment 
recirculation, or fourth-stage automatic depressurization system 
(ADS) valves, including as-installed inspections and maintenance 
requirements as described in the Updated Final Safety Analysis 
Report (UFSAR). Inadvertent operation or failure of the fourth-stage 
ADS valves are considered as an accident initiator or part of an 
initiating sequence of events for an accident previously evaluated. 
However, the proposed change to the test methodology and calculated 
flow resistance for the fourth-stage ADS lines does not adversely 
affect the probability of inadvertent operation or failure. 
Therefore, the probabilities of the accidents previously evaluated 
in the UFSAR are not affected.
    The proposed changes do not adversely affect the ability of 
IRWST injection, drain, containment recirculation, and fourth-stage 
ADS valves to perform their design functions. The designs of the 
IRWST injection, drain, containment recirculation, and fourth-stage 
ADS valves continue to meet the same regulatory acceptance criteria, 
codes, and standards as required by the UFSAR. In addition, the 
proposed changes maintain the capabilities of the IRWST injection, 
drain, containment recirculation, and fourth-stage ADS valves to 
mitigate the consequences of an accident and to meet the applicable 
regulatory acceptance criteria.
    The proposed changes do not adversely affect the prevention and 
mitigation of other abnormal events, e.g., anticipated operational 
occurrences, earthquakes, floods and turbine missiles, or their 
safety or design analyses. Therefore, the consequences of the 
accidents evaluated in the UFSAR are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that might initiate a new or different kind of 
accident, or alter any SSC such that a new accident initiator or 
initiating sequence of events is created. The proposed changes do

[[Page 23630]]

not adversely affect the physical design and operation of the IRWST 
injection, drain, containment recirculation, and fourth-stage ADS 
valves, including as-installed inspections, and maintenance 
requirements, as described in the UFSAR. Therefore, the operation of 
the IRWST injection, drain, containment recirculation, and fourth-
stage ADS valves is not adversely affected. These proposed changes 
do not adversely affect any other SSC design functions or methods of 
operation in a manner that results in a new failure mode, 
malfunction, or sequence of events that affect safety-related or 
nonsafety-related equipment. Therefore, this activity does not allow 
for a new fission product release path, result in a new fission 
product barrier failure mode, or create a new sequence of events 
that 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.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes maintain existing safety margins. The 
proposed changes verify and maintain the capabilities of the IRWST 
injection, drain, containment recirculation, and fourth-stage ADS 
valves to perform their design functions. The proposed changes 
maintain existing safety margin through continued application of the 
existing requirements of the UFSAR, while updating the acceptance 
criteria for verifying the design features necessary to ensure the 
IRWST injection, drain, containment recirculation, and fourth-stage 
ADS valves perform the design functions required to meet the 
existing safety margins in the safety analyses. Therefore, the 
proposed changes satisfy the same design functions in accordance 
with the same codes and standards as stated in the UFSAR.
    These changes do not adversely affect any design code function, 
design analysis, safety analysis input or result, or design/safety 
margin.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

    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.

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

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3 (PVNGS), Maricopa County, Arizona

    Date of amendment request: April 1, 2016, as supplemented by 
letters dated July 21, September 9, and October 26, 2016.
    Description of amendment request: The amendments revised the 
Technical Specifications (TSs) for PVNGS, by modifying the requirements 
regarding the degraded and loss of voltage relays that are planned to 
be modified to be more aligned with designs generally implemented in 
the industry. Specifically, the licensing basis for degraded voltage 
protection will be changed from reliance on a TS initial condition that 
ensures adequate post-trip voltage support of accident mitigation 
equipment to crediting automatic actuation of the degraded and loss of 
voltage relays to ensure proper equipment performance.
    Date of issuance: April 27, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: Unit 1-201, Unit 2-201, and Unit 3-201. A publicly-
available version is in ADAMS under Accession No. ML17090A164; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
The amendment revised the Operating Licenses and TSs.
    Date of initial notice in Federal Register: May 24, 2016 (81 FR 
32803). The supplements dated July 21, September 9, and October 26, 
2016, provided additional information that clarified the application, 
did not expand the scope of the application as originally noticed, and 
did not change the NRC staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 27, 2017.
    No significant hazards consideration comments received: No.

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

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2 (CNS), York County, South Carolina

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

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2 (MNS), Mecklenburg County, North 
Carolina

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

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

    Date of amendment request: June 23, 2016.
    Brief description of amendments: The amendments modified the 
technical specification (TS) requirements for unavailable barriers by 
adding Limiting

[[Page 23631]]

Condition for Operation (LCO) 3.0.9 to the TS for BSEP, ONS, and RNP. 
The same changes were added as LCO 3.0.10 to the TS for CNS and MNS. 
For HNP, TS requirements for unavailable barriers were modified by 
adding LCO 3.0.6 to the TS. The changes are consistent with Technical 
Specification Task Force Traveler (TSTF)-427, Revision 2, ``Allowance 
for Non-Technical Specification Barrier Degradation on Supported System 
OPERABILITY,'' subject to stated variations.
    Date of issuance: April 26, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos: 274/302 (BSEP), 288/284 (CNS), 155 (HNP), 295/274 
(MNS), 402/404/403 (ONS), and 251 (RNP). A publicly-available version 
is in ADAMS under Accession No. ML17066A374; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Facility Operating License Nos. DPR-71 and DPR-62 (BSEP), NPF-35 
and NPF-52 (CNS), NPF-63 (HNP), NPF-9 and NPF-17 (MNS), DPR-38, DPR-47, 
DPR-55 (ONS), and DPR-23 (RNP): Amendments revised the Facility 
Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: August 16, 2016 (81 FR 
54614).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 26, 2017.
    No significant hazards consideration comments received: No.

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

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

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

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: September 27, 2016, as supplemented by 
letters dated November 22, 2016, and April 20, 2017.
    Description of amendment request: The amendments revised Technical 
Specification Surveillance Requirements to require operating 
ventilation systems with charcoal filters for 15 continuous minutes 
every 31 days or at a frequency controlled in accordance with the 
Surveillance Frequency Control Program. The amendments are consistent 
with NRC-approved Technical Specifications Task Force (TSTF) Traveler 
TSTF-522, Revision 0, ``Revise Ventilation System Surveillance 
Requirements to Operate for 10 hours per Month,'' as published in the 
Federal Register on September 20, 2012 (77 FR 58428), with variations 
due to plant-specific differences.
    Date of issuance: May 8, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 275 (Unit 1) and 303 (Unit 2) for the Brunswick 
Steam Electric Plant; 289 (Unit 1) and 285 (Unit 2) for the Catawba 
Nuclear Station; 296 (Unit 1) and 275 (Unit 2) for the McGuire Nuclear 
Station; 156 (Unit 1) for the Shearon Harris Nuclear Power Plant; and 
252 (Unit No. 2) for the H. B. Robinson Steam Electric Plant. A 
publicly-available version is in ADAMS under Accession No. ML17055A647; 
documents related to these amendments are listed in the Safety 
Evaluations enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-71 and DPR-62, for the 
Brunswick Steam Electric Plant, Units 1 and 2; NPF-35 and NPF-52, for 
the Catawba Nuclear Station, Units 1 and 2; NPF-9 and NPF-17, for the 
McGuire Nuclear Station, Units 1 and 2; NPF-63, for the Shearon Harris 
Nuclear Power Plant, Unit 1; and DPR-23, for the H. B. Robinson Steam 
Electric Plant, Unit No. 2: The amendments revised the Renewed Facility 
Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: January 17, 2017 (82 FR 
4929). The supplemental letter dated April 20, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the NRC staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluations of the amendments are 
contained in Safety Evaluations dated May 8, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: February 10, 2016, as supplemented by 
letters dated October 10 and December 16, 2016; and January 31, 
February 7, February 16, and March 29, 2017.
    Brief description of amendment: The amendment revised Technical 
Specification 5.5.11, ``Primary Containment Leakage Rate Testing 
Program,'' to increase the containment integrated leakage rate test 
program Test A interval from 10 to 15 years.
    Date of issuance: April 25, 2017.
    Effective date: As of the date of issuance and shall be implemented 
prior to the startup from the 2017 refueling outage.
    Amendment No.: 193. A publicly-available version is in ADAMS under 
Accession No. ML17103A235; 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: April 26, 2016 (81 FR 
24663). The supplemental letters dated October 10 and December 16, 
2016; and January 31, February 7, February 16, and March 29, 2017, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the NRC staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 25, 2017.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant (DCPP), Units 1 and 2, San Luis Obispo 
County, California

    Date of application for amendments: June 17, 2015, as supplemented 
by letters dated August 31, October 22, November 2, November 6, and 
December 17, 2015; and February 1, February 10, April 21, June 9, 
September 15, October 6, and December 27, 2016.
    Brief description of amendments: The amendments revised the 
licensing bases to adopt the alternative source term (AST) as allowed 
by 10 CFR 50.67, ``Accident source term.'' The AST

[[Page 23632]]

methodology, as established in NRC Regulatory Guide 1.183, 
``Alternative Radiological Source Terms for Evaluating Design Basis 
Accidents at Nuclear Power Reactors,'' July 2000 (ADAMS Accession No. 
ML003716792), is used to calculate the offsite and control room 
radiological consequences of postulated accidents for DCPP, Units 1 and 
2. The amendments revised Technical Specification (TS) 1.1, 
``Definitions,'' for the definition of Dose Equivalent I-131; TS 
3.4.16, ``RCS [Reactor Coolant System] Specific Activity,'' to revise 
the noble gas activity limit; TS 3.6.3, ``Containment Isolation 
Valves,'' to require the 48-inch containment purge supply and exhaust 
valves to be sealed closed during Modes 1, 2, 3, and 4; TS 5.5.11, 
``Ventilation Filter Testing Program (VFTP),'' to change the allowable 
methyl iodide penetration testing criteria for the auxiliary building 
system charcoal filter; TS 5.5.19, ``Control Room Habitability 
Program,'' to replace ``whole body or its equivalent to any part of the 
body,'' with ``Total Effective Dose Equivalent,'' which is the dose 
criteria specified in 10 CFR 50.67, and Appendix D, ``Additional 
Conditions,'' for Facility Operating License Nos. DPR-80 and DPR-82 for 
DCPP, Units 1 and 2, to add additional license conditions.
    Date of issuance: April 27, 2017.
    Effective date: As of its date of issuance and shall be implemented 
within 365 days from the date of issuance.
    Amendment Nos.: Unit 1-230; Unit 2-232. A publicly-available 
version is in ADAMS under Accession No. ML17012A246; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: The license amendment 
request was originally noticed in the Federal Register on October 13, 
2015 (80 FR 61486). As a result of the supplemental letters dated 
October 22, November 2, November 6, and December 17, 2015; and February 
1, February 10, April 21, June 9, and September 15, 2016, the notice 
was reissued in its entirety to include the revised scope, description 
of the amendment request, and proposed no significant hazards 
consideration determination on November 8, 2016 (81 FR 78664).
    The supplemental letters dated October 6 and December 27, 2016, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the NRC staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 27, 2017.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: June 7, 2016.
    Brief description of amendment: The amendment modified the 
Technical Specifications to allow the use of Component Cooling System 
(CCS) pump 2B-B to support Train 1B operability when the normally 
aligned CCS pump C-S is removed from service.
    Date of issuance: April 27, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 113. A publicly-available version is in ADAMS under 
Accession No. ML17081A263; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-90: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: September 13, 2016 (81 
FR 62932).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 27, 2017.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 11th day of May 2017.

    For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2017-10570 Filed 5-22-17; 8:45 am]
 BILLING CODE 7590-01-P