Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 154-164 [2016-31813]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 82, Number 1 (Tuesday, January 3, 2017)]
[Notices]
[Pages 154-164]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-31813]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0273]


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

[[Page 155]]

publish notice of any amendments issued, or proposed to be issued, and 
grants the Commission the authority to issue and make immediately 
effective any amendment to an operating license or combined license, as 
applicable, upon a determination by the Commission that such amendment 
involves no significant hazards consideration, notwithstanding the 
pendency before the Commission of a request for a hearing from any 
person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued, from December 6 to December 19, 2016. The last 
biweekly notice was published on December 20, 2016.

DATES: Comments must be filed by February 2, 2017. A request for a 
hearing must be filed by March 6, 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-2016-0273. 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: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Janet Burkhardt, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-1384, email: Janet.Burkhardt@nrc.gov.

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0273, 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-2016-0273.
     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-2016-0273, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

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

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

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if

[[Page 156]]

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 March 
6, 2017. The petition must be filed in accordance with the filing 
instructions in the ``Electronic Submissions (E-Filing)'' section of 
this document, and should meet the requirements for petitions set forth 
in this section, except that under 10 CFR 2.309(h)(2) a State, local 
governmental body, or federally recognized Indian Tribe, or agency 
thereof does not need to address the standing requirements in 10 CFR 
2.309(d) if the facility is located within its boundaries. 
Alternatively, a State, local governmental body, Federally-recognized 
Indian Tribe, or agency thereof may participate as a non-party under 10 
CFR 2.315(c).
    If a hearing is granted, any person who is not a party to the 
proceeding and is not affiliated with or represented by a party may, at 
the discretion of the presiding officer, be permitted to make a limited 
appearance pursuant to the provisions of 10 CFR 2.315(a). A person 
making a limited appearance may make an oral or written statement of 
his or her position on the issues but may not otherwise participate in 
the proceeding. A limited appearance may be made at any session of the 
hearing or at any prehearing conference, subject to the limits and 
conditions as may be imposed by the presiding officer. Details 
regarding the opportunity to make a limited appearance will be provided 
by the presiding officer if such sessions are scheduled.

B. Electronic Submissions (E-Filing)

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

[[Page 157]]

adjudicatory documents. Submissions should 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 7 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://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or personal phone numbers in their filings, unless an 
NRC regulation or other law requires submission of such information. 
For example, in some instances, individuals provide home addresses in 
order to demonstrate proximity to a facility or site. With respect to 
copyrighted works, except for limited excerpts that serve the purpose 
of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.
    Dominion Nuclear Connecticut, Inc. (DNC), Docket No. 50-336, 
Millstone Power Station, Unit No. 2 (MPS2), New London County, 
Connecticut.
    Date of amendment request: December 14, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16354A424.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) Surveillance Requirement (SR) 4.1.3.1.2 
regarding control element assembly (CEA) freedom of movement 
surveillance, such that CEA 39 may be excluded from the last remaining 
quarterly performance of the SR in MPS2 Cycle 24.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would exclude CEA 39 from SR 4.1.3.1.2 
for the remainder of MPS2 Cycle 24 operation. The function of CEA 39 
is to provide negative reactivity addition into the core upon 
receipt of a signal from the Reactor Protection System (RPS). CEA 39 
was demonstrated to be moveable and trippable during the last 
performance of SR 4.1.3.1.2. Since the functionality of CEA 39 has 
not been affected, the assumptions and conclusions of the Final 
Safety Analysis Report (FSAR) Chapter 14, Safety Analysis, are not 
affected by this license amendment request.
    The misoperation of a CEA, which includes a CEA drop event, has 
been evaluated in the MPS2 FSAR and found acceptable. The proposed 
change would minimize the potential for inadvertent insertion of CEA 
39 into the core by eliminating the requirement to place the CEA on 
the UGC to perform freedom of movement testing. The proposed change 
does not significantly increase the probability of a failure of a 
CEA to insert into the core on a reactor trip or the probability of 
an inadvertent CEA drop into the core at power.
    No modifications are proposed to the RPS or associated Control 
Element Drive Mechanism (CEDM) system logic.
    Based on the above, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment would exclude CEA 39 from SR 4.1.3.1.2 
for the remainder of MPS2 Cycle 24 operation. CEA 39 was 
demonstrated to be moveable and trippable during the last 
performance of SR 4.1.3.1.2; therefore, the functionality of CEA 39 
has not been affected. The proposed change will not introduce any 
new design changes or systems that can prevent the CEA from 
performing its specified safety function to insert on a reactor 
trip. The current MPS2 FSAR safety analysis considers the drop of a 
CEA into the core as an initiating event. This change does not alter 
assumptions made in the FSAR Chapter 14 safety analysis.
    Based on the above, 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 the margin of safety?
    Response: No.
    The proposed amendment would exclude CEA 39 from SR 4.1.3.1.2 
for the remainder of MPS2 Cycle 24 operation. SR 4.1.3.1.2 is 
intended to verify freedom of movement of CEAs (i.e., trippable). 
CEA 39 was demonstrated to be moveable and trippable during the last 
performance of SR 4.1.3.1.2.

[[Page 158]]

The physical and electrical design of the CEAs, and past operating 
experience, provides high confidence that CEAs remain trippable 
whether or not exercised during each SR interval. Eliminating 
further exercise of CEA 39 for the remainder of MPS2 Cycle 24 
operation does not directly relate to the potential for CEA binding 
to occur. The current MPS2 FSAR safety analysis is unaffected by 
this license amendment request and there is no reduction in the 
margin of safety.
    There is no known failure mechanism (e.g., crud deposition) that 
would preclude the CEA from inserting into core on a valid trip 
signal or loss of power.
    Based on the above, the proposed amendment does not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resources Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 
23219.
    NRC Acting Branch Chief: Stephen S. Koenick.

    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: June 30, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16193A656.
    Description of amendment request: The amendments would modify 
Technical Specification 3.6.14, ``Divider Barrier Integrity,'' to 
revise Conditions A and D to allow one steam generator (SG) enclosure 
hatch or one pressurizer enclosure hatch to be open for up to 48 hours 
to facilitate potential inspections and maintenance and to enhance 
personnel and radiation safety.
    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.
    Implementation of this amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Removal of the SG enclosure hatch or the 
pressurizer enclosure hatch will not cause an increase in the 
probability of an accident that has been previously evaluated 
because the hatches are not accident initiators.
    The consequences of an accident, which have been previously 
evaluated, will not be significantly increased by removal of the 
pressurizer enclosure or SG enclosure hatch. As discussed in the 
technical justification supporting this amendment request, the new 
containment compression peak pressure will remain well below the 
acceptance criteria. Additionally, the long term containment peak 
pressure will not be adversely affected due to the delay time in 
melting of the ice.
    The removal of the pressurizer enclosure hatch itself has been 
previously evaluated in Modes 1 through 4 in accordance with the 
analytical method described in NUREG-0612 and the NRC's December 22, 
1980, letter regarding the control of heavy loads at nuclear power 
plants [(ADAMS Accession Nos. ML070250180 and ML071080219, 
respectively)]. Because the SG enclosure hatch weighs less than 300 
pounds, it would not be considered a heavy load as defined by NUREG-
0612. As such, it is not subject to heavy lift considerations. 
Regardless, there is no safety-related equipment directly under 
these hatch covers, so in the unlikely event that one fell, no 
damage is expected to be caused. The changes proposed in this 
[license amendment request (LAR)] have no adverse effect on the 
procedures used for the handling of heavy loads at McGuire.
    In summary, the proposed changes will not involve any 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.
    Operation in accordance with the proposed amendment does not 
create a new plant configuration and does not adversely affect how 
the plant is operated, so implementation of this amendment would not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. No new accident causal mechanisms 
are created as a result of the NRC approval of this license 
amendment request. As discussed above, extending the time that the 
pressurizer enclosure hatch or SG enclosure hatch is allowed to be 
open or inoperable does not create any new or different accidents 
from those previously evaluated. Removal of the pressurizer 
enclosure hatch to perform inspections or maintenance has been 
previously evaluated and determined to be acceptable. The analysis 
contained in the technical justification for this license amendment 
request provides results concluding that the containment compression 
peak pressure and the long term containment peak pressure are 
acceptable with either a pressurizer enclosure hatch or [an] SG 
enclosure hatch open. This proposed amendment does not impact any 
plant systems that are accident initiators; therefore, no new 
accident types are being created.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    Implementation of this amendment would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. The pressurizer enclosure hatch 
and the SG enclosure hatch, as well as their performances, have a 
direct impact on the containment boundary since peak containment 
pressure due to an accident could be affected. However, the analysis 
supporting this amendment request concludes that the containment 
compression peak pressure and the long term containment peak 
pressure continue to be acceptable with the increased time a single 
hatch is open.
    Therefore, the performance of the fission product barriers will 
not be significantly impacted by implementation of this amendment, 
and no safety margins will be significantly impacted.

    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 Nolan, Deputy General Counsel, Duke 
Energy Corporation, 526 South Church Street--DEC45A, Charlotte, NC 
28202.
    NRC Branch Chief: Michael T. Markley.
    Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick 
Steam Electric Plant, Units 1 and 2 (BSEP), Brunswick County, North 
Carolina.
    Date of amendment request: September 6, 2016, as supplemented by 
letter dated November 9, 2016. Publicly-available versions are in ADAMS 
under Accession Nos. ML16257A410 and ML16330A504, respectively.
    Description of amendment request: The amendment would revise the 
Technical Specifications to support an expansion of the core power-flow 
operating range (i.e., Maximum Extended Load Line Limit Analysis Plus 
(MELLLA+)).
    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 operation in the MELLLA+ operating domain does not 
significantly increase the probability or consequences of an 
accident previously evaluated. The probability (i.e., frequency of 
occurrence) of

[[Page 159]]

Design Basis Accidents (DBAs) occurring is not affected by the 
MELLLA+ operating domain because BSEP continues to comply with the 
regulatory and design basis criteria established for plant 
equipment. Furthermore, a probabilistic risk assessment demonstrates 
that the calculated core damage frequencies do not significantly 
change due to the MELLLA+.
    There is no change in consequences of postulated accidents when 
operating in the MELLLA+ operating domain compared to the operating 
domain previously evaluated. The results of accident evaluations 
remain within the NRC approved acceptance limits.
    The spectrum of postulated transients has been investigated and 
is shown to meet the plant's currently licensed regulatory criteria. 
Continued compliance with the Safety Limit Minimum Critical Power 
Ratio (SLMCPR) will be confirmed on a cycle-specific basis 
consistent with the criteria accepted by the NRC.
    Challenges to the reactor coolant pressure boundary were 
evaluated for the MELLLA+ operating domain conditions (i.e., 
pressure, temperature, flow, and radiation) and were found to meet 
their respective acceptance criteria for allowable stresses and 
overpressure margin.
    Challenges to the containment were evaluated and the containment 
and its associated cooling systems continue to meet the current 
licensing basis. The calculated post-Loss of Coolant Accident (LOCA) 
suppression pool temperature remains acceptable.
    The proposed changes to the sodium pentaborate (SPB) enrichment 
and volume requirements maintain the capability of the Standby 
Liquid Control (SLC) system to perform this reactivity control 
function and ensure continued compliance with the requirements of 10 
CFR 50.62. The SLC system is not considered to be an initiator of 
any event. The use of the proposed SPB solution with a higher boron-
10 (B-10) isotope enrichment does not alter the design, function, or 
operation of the SLC system or increase the likelihood of 
malfunction that could increase the consequences of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed operation in the MELLLA+ operating domain does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    Equipment that could be affected by the MELLLA+ operating domain 
has been evaluated. No new operating mode, safety-related equipment 
lineup, accident scenario, or equipment failure mode was identified. 
The full spectrum of accident considerations has been evaluated and 
no new or different kind of accident was identified. The MELLLA+ 
operating domain uses developed technology, and applies it within 
the capabilities of existing plant safety-related equipment in 
accordance with the regulatory criteria, including NRC-approved 
codes, standards and methods. The use of the proposed SPB solution 
with a higher B-10 isotope enrichment does not alter the design, 
function, or operation of the SLC system or create the possibility 
of a new or different kind of accident. The proposed changes have 
been assessed and determined not to introduce a different accident 
than that previously evaluated. No new accident or event precursor 
has been identified.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed operation in the MELLLA+ domain does not involve a 
significant reduction in the margin of safety.
    The MELLLA+ operating domain can only affect design and 
operational margins. Challenges to the fuel, reactor coolant 
pressure boundary, and containment were evaluated for the MELLLA+ 
operating domain conditions. Fuel integrity is maintained by meeting 
existing design and regulatory limits. The calculated loads on 
affected structures, systems, and components, including the reactor 
coolant pressure boundary, will remain within their design 
allowables for design basis event categories. No NRC acceptance 
criterion is exceeded. The BSEP configuration and responses to 
transients and postulated accidents do not result in exceeding the 
presently approved NRC acceptance limits, thereby preserving safety 
margins.
    The proposed changes to the SPB enrichment and volume 
requirements ensure SLC system shutdown margins and post-accident pH 
control margins are maintained while maintaining compliance with the 
requirements of 10 CFR 50.62.
    Therefore, the proposed amendments do not result in 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: Kathryn B. Nolan, Deputy General Counsel, 
550 South Tryon St., M/C DEC45A, Charlotte, NC 28202.
    NRC Acting Branch Chief: Jeanne D. Johnston.

    Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 
50-277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3 
(PBAPS), York and Lancaster Counties, Pennsylvania.
    Date of amendment request: November 4, 2016, as supplemented by 
letter dated December 7, 2016. Publicly-available versions are in ADAMS 
under Accession Nos. ML16309A298 and ML16342C455, respectively.
    Description of amendment request: The amendments would revise the 
Allowable Value (AV) for the Turbine Condenser--Low Vacuum scram 
function specified in Technical Specification Table 3.3.1.1-1, 
``Reactor Protection System Instrumentation.'' The licensee stated that 
the purpose of the proposed change is to minimize the potential for 
inadvertent scrams due to low condenser vacuum.
    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 shown in square 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change implements a revised AV [allowable value] 
for the Turbine Condenser--Low Vacuum scram instrument function at 
PBAPS.
    The proposed change to the PBAPS Turbine Condenser- Low Vacuum 
scram AV does not require modifying any system interface or affect 
the probability of any event initiators at the facility. Overall RPS 
[Reactor Protection System] performance will remain within the 
bounds of the previously performed accident analyses, since [the 
Turbine-Condenser--Low Vacuum scram is not specifically credited in 
any accident analysis.]
    There will be no degradation in the performance of, or an 
increase in the number of challenges imposed on safety-related 
equipment that are assumed to function during an accident situation. 
The proposed change will not alter any assumptions or change any 
mitigation actions in the radiological consequence evaluations in 
the UFSAR [Updated Final Safety Analysis Report]. The proposed 
change is consistent with safety analysis assumptions and resultant 
consequences.
    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 PBAPS Turbine Condenser--Low Vacuum 
scram AV does not affect the design, functional performance, or 
operation of the facility. Similarly, the proposed change does not 
affect the design or operation of any SSCs [structures, systems, or 
components] involved in the mitigation of any accidents, nor does it 
affect the design or operation of any component in the facility such 
that new equipment failure modes are created.

[[Page 160]]

    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this change. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not affect the acceptance criteria for 
any analyzed event, nor is there a change to any Safety Analysis 
Limit. There will be no effect on the manner in which safety limits, 
limiting safety system settings, or limiting conditions for 
operation are determined nor will there be any effect on those plant 
systems necessary to assure the accomplishment of protection 
functions.
    The purpose of the condenser low vacuum turbine trip is to 
protect the main condenser against overpressure on loss of condenser 
vacuum. A condenser low vacuum condition provides a signal to trip 
the main turbine by providing automatic closure to the turbine stop 
valves. To anticipate the transient and scram which results from the 
closure of the turbine stop valves, a condenser low vacuum condition 
initiates a reactor scram. The condenser low vacuum scram trip 
setting is selected to initiate a reactor scram prior to initiation 
of closure of the [t]urbine [s]top [v]alves.
    The proposed LAR [license amendment request] does not change the 
sequential relationship of the condenser low vacuum scram and 
turbine trip. The Automatic Scram signal (Actual Trip Setpoint 
greater than or equal to 21.95 inches [mercury (Hg)] vacuum) will 
still occur prior to the Turbine Trip signal (Actual Trip Setpoint 
20.0 inches Hg vacuum). This aligns with UFSAR Section 7.2 in that 
the condenser low vacuum scram is an anticipatory trip prior to the 
scram that would result from the closure of the main turbine stop 
valves.
    The condenser low vacuum scram is not specifically credited in 
any accident analysis. The integrity of the condenser is not 
compromised by the proposed change because the reactor will be shut 
down using both diverse and redundant tripping to ensure fission 
products are not released.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

    Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-348 
and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
County, Alabama.
    Date of amendment request: November 22, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16336A024.
    Description of amendment request: SNC requested to revise the 
licensing basis that support a selected scope application of an 
Alternative Source Term (AST) methodology and incorporate Technical 
Specification Task Force (TSTF) Traveler, TSTF-448-A, Revision 3, 
``Control Room Habitability,'' and TSTF-312-A, ``Administrative Control 
of Containment Penetrations.''
    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 licensee's analysis is 
presented below, with NRC staff edits in square brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    There are no physical changes to the plant being introduced by 
the proposed changes to the accident source term. Implementation of 
Alternative Source Term (AST) and the new atmospheric dispersion 
factors have no impact on the probability for initiation of any 
Design Basis Accidents (DBAs). Once the occurrence of an accident 
has been postulated, the new accident source term and atmospheric 
dispersion factors are an input to analyses that evaluate the 
radiological consequences. The proposed changes do not involve a 
revision to the design or manner in which the facility is operated 
that could increase the probability of an accident previously 
evaluated in Chapter 15 of the Final Safety Analysis Report (FSAR).
    Based on the AST analyses, there are no proposed changes to 
performance requirements and no proposed revision to the parameters 
or conditions that could contribute to the consequences of an 
accident previously discussed in Chapter 15 of the FSAR. Plant-
specific radiological analyses have been performed using the AST 
methodology and new atmospheric dispersion factors (X/Qs) have been 
established. Based on the results of these analyses, it has been 
demonstrated that the Control Room and off-site dose consequences of 
the limiting events considered in the analyses meet the regulatory 
guidance provided for use with the AST, and the doses are within the 
limits established by 10 CFR 50.67.
    Regarding TSTF-312-A, the proposed change would allow 
containment penetrations to be unisolated under administrative 
controls during core alterations or movement of irradiated fuel 
assemblies within containment. The status of containment penetration 
flow paths (i.e., open or closed) is not an initiator for any design 
basis accident or event, and therefore the proposed change does not 
increase the probability of any accident previously evaluated. The 
proposed change does not affect the design of the primary 
containment, or alter plant operating practices such that the 
probability of an accident previously evaluated would be 
significantly increased. The proposed change does not significantly 
change how the plant would mitigate an accident previously 
evaluated, and is bounded by the fuel handling accident (FHA) 
analysis.
    Therefore, it is concluded that the proposed amendment does not 
involve a significant increase in the probability or the 
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.
    No new modes of operation are introduced by the proposed 
changes. The proposed changes will not create any failure mode not 
bounded by previously evaluated accidents. Implementation of AST and 
the associated proposed Technical Specification changes and new X/Qs 
have no impact to the initiation of any DBAs. These changes do not 
affect the design function or modes of operation of structures, 
systems and components in the facility prior to a postulated 
accident. Since structures, systems and components are operated no 
differently after the AST implementation, no new failure modes are 
created by this proposed change. The AST change itself does not have 
the capability to initiate accidents.
    Regarding TSTF-312-A, allowing penetration flow paths to be open 
is not an initiator for any accident. The proposed change to allow 
open penetration flow paths will not affect plant safety functions 
or plant operating practices such that a new or different accident 
could be created. There are no design changes associated with the 
proposed changes, and the change does not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed). The change does not alter assumptions made in 
the safety analysis, and is consistent with the safety analysis 
assumptions and current plant operating practice.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Consequently, the proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The AST analyses have been performed using approved 
methodologies to ensure that analyzed events are bounding and safety

[[Page 161]]

margin has not been reduced. The dose consequences of these limiting 
events are within the acceptance criteria presented in 10 CFR 50.67. 
Thus, by meeting the applicable regulatory limits for AST, there is 
no significant reduction in a margin of safety.
    Regarding TSTF-312-A, TS 3.9.3 provides measures to ensure that 
the dose consequences of a postulated FHA inside containment are 
minimized. The proposed change to LCO 3.9.3 will allow penetration 
flow path(s) to be open during refueling operations under 
administrative control. These administrative controls will provide 
assurance that prompt closure of open penetrations flow paths can 
and will be achieved in the event of an FHA inside containment, and 
will minimize dose consequences. The proposed change does not affect 
the safety analysis acceptance criteria for ay [an] analyzed event, 
nor is there a change to any safety analysis limit. The proposed 
change does not alter the manner in which safety limits, limiting 
safety system settings or limiting conditions for operation are 
deterred, not [nor] is here [there] any adverse effect on those 
plant systems necessary to assure the accomplishment of protective 
functions. The proposed change will not result in plant operation in 
a configuration outside the design basis.
    Therefore, because the proposed changes continue to result in 
dose consequences within the applicable regulatory limits, the 
proposed amendment does not involve a significant reduction in 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, 40 Inverness Center 
Parkway, Birmingham, AL 35201.
    NRC Branch Chief: Michael T. Markley.

    Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama.
    Date of amendment request: November 15, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16320A540.
    Description of amendment request: The proposed changes would revise 
Technical Specification 5.5.17, ``Containment Leakage Rate Testing 
Program.'' The revision would increase the existing Type A integrated 
leakage rate test program test interval from 10 years to 15 years; 
adopt an extension of the containment isolation valve leakage testing 
(Type C) frequency from 60 months to 75 months; adopt the use of 
American National Standards Institute/American Nuclear Society (ANSI/
ANS) 56.8-2002, ``Containment System Leakage Testing Requirements''; 
and adopt a grace interval of 9 months for Type A, Type B, and Type C 
leakage tests, in accordance with Nuclear Energy Institute (NEI) 94-01, 
Revision 3-A.
    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 activity involves the revision of Joseph M. Farley 
Nuclear Plant (FNP), Units 1 and 2, Technical Specification (TS) 
5.5.17, ``Primary Containment Leakage Rate Testing Program,'' to 
allow the extension of the Type A integrated leakage rate test 
(ILRT) containment test interval to 15 years, and the extension of 
the Type C local leakage rate test (LLRT) interval to 75 months. The 
current Type A test interval of 120 months (10 years) would be 
extended on a permanent basis to no longer than 15 years from the 
last Type A test. The current Type C test interval of 60 months for 
selected components would be extended on a performance basis to no 
longer than 75 months. Extensions of up to nine months (total 
maximum interval of 84 months for Type C tests) are permissible only 
for non-routine emergent conditions.
    The proposed extensions do 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. As such, 
the containment and the testing requirements invoked to periodically 
demonstrate the integrity of the containment 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.
    The change in Type A test frequency to once-per-fifteen years, 
measured as an increase to the total integrated plant risk for those 
accident sequences influenced by Type A testing, based on the 
internal events probabilistic risk analysis (PRA) is 1.08E-02 
person-rem/year for Unit 1 and 9.89 E-03 person-rem/year for Unit 2. 
Electric Power Research Institute (EPRI) Report No. 1009325, 
Revision 2-A states that a very small population is defined as an 
increase of <= 1.0 person-rem per year or <= 1% of the total 
population dose, whichever is less restrictive for the risk impact 
assessment of the extended ILRT intervals. This is consistent with 
the Nuclear Regulatory Commission (NRC) Final Safety Evaluation for 
Nuclear Energy Institute (NEI) 94-01 and EPRI Report No. 1009325. 
Moreover, the risk impact when compared to other severe accident 
risks is negligible.
    Therefore, this proposed extension does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    In addition, as documented in NUREG-1493, ``Performance-Based 
Containment Leak-Test Program,'' dated January 1995, Types B and C 
tests have identified a very large percentage of containment leakage 
paths, and the percentage of containment leakage paths that are 
detected only by Type A testing is very small. The FNP Type A test 
history supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and (2) time based. Activity-based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. The LLRT requirements and administrative controls such 
as configuration management and procedural requirements for system 
restoration ensure that containment integrity is not degraded by 
plant modifications or maintenance activities. The design and 
construction requirements of the containment combined with the 
containment inspections performed in accordance with American 
Society of Mechanical Engineers (ASME) Section XI, and TS 
requirements serve to provide a high degree of assurance that the 
containment would not degrade in a manner that is detectable only by 
a Type A test. Based on the above, the proposed test interval 
extensions do not significantly increase the consequences of an 
accident previously evaluated.
    The proposed amendment also deletes exceptions previously 
granted under TS Amendments 159 (FNP Unit 1) and 150 (FNP Unit 2) to 
allow one-time extensions of the ILRT test frequency for FNP. These 
exceptions were for activities that would have already taken place 
by the time this amendment is approved; therefore, their deletion is 
solely an administrative action that has no effect on any component 
and no impact on how the unit is operated.
    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 TS 5.5.17, ``Containment Leakage 
Rate Testing Program,'' involves the extension of the FNP Type A 
containment test interval to 15 years and the extension of the Type 
C test interval to 75 months. 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.
    The proposed amendment also deletes exceptions previously 
granted under TS

[[Page 162]]

Amendments 159 (FNP Unit 1) and 150 (FNP Unit 2) to allow one-time 
extensions of the ILRT test frequency for FNP. These exceptions were 
for activities that would have already taken place by the time this 
amendment is approved; therefore, their deletion is solely an 
administrative action that does not result in any change in how the 
unit is operated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment to TS 5.5.17 involves the extension of 
the FNP Type A containment test interval to 15 years and the 
extension of the Type C test interval to 75 months for selected 
components. 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 Containment Leak Rate Testing Program exist to 
ensure that the degree of containment structural integrity and leak-
tightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by TS 
is maintained.
    The proposed change involves only the extension of the interval 
between Type A containment leak rate tests and Type C tests for FNP. 
The proposed surveillance interval extension is bounded by the 15-
year ILRT interval and the 75-month Type C test interval currently 
authorized within NEI 94-01, Revision 3-A. Industry experience 
supports the conclusions that Type B and C testing detects a large 
percentage of containment leakage paths and that the percentage of 
containment leakage paths that are detected only by Type A testing 
is small. The containment inspections performed in accordance with 
ASME Section XI and Technical Specifications serve to provide a high 
degree of assurance that the containment would not degrade in a 
manner that is detectable only by Type A testing. The combination of 
these factors ensures that the margin of safety in the plant safety 
analysis is maintained. 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 acceptance of this proposed change, since these are 
not affected by changes to the Type A and Type C test intervals.
    The proposed amendment also deletes an exception previously 
granted under TS Amendments 159 (FNP Unit 1) and 150 (FNP Unit 2) to 
allow one-time extensions of the ILRT test frequency for FNP. This 
exception was for an activity that would have already taken place by 
the time this amendment is approved; therefore, the deletion is 
solely an administrative action and does not change how the unit is 
operated and maintained. Therefore, there is no reduction in any 
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: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center 
Parkway, Birmingham, AL 35242.
    NRC Branch Chief: Michael T. Markley.

    Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri.
    Date of amendment request: October 11, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16286A553.
    Description of amendment request: The amendment would revise the 
Technical Specification (TS) requirements to reference and allow use of 
the NRC-approved core reload methodologies described in Westinghouse 
topical reports (TRs) WCAP-16045-P-A, ``Qualification of the Two-
Dimensional Transport Code PARAGON,'' WCAP-16045-P-A, Addendum 1-A, 
``Qualification of the NEXUS Nuclear Data Methodology,'' and WCAP-
10965-P-A, Addendum 2-A, ``Qualification of the New Pin Power Recovery 
Methodology,'' for the Callaway 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 license amendment would revise TS 5.6.5.b to add 
additional TR references for NRC-approved methodologies used in core 
reload designs and the determination of core operating limits, 
thereby specifically approving the use of these methodologies for 
the Callaway Plant. The additional analytical methodologies are 
improvements over the current methodologies in use at Callaway 
Plant. The NRC staff reviewed and approved these methodologies and 
concluded that these analytical methods are acceptable as a 
replacement for the current analytical method.
    This proposed license amendment does not involve any physical 
changes to the Callaway Plant. Additionally, the core operating 
limits determined using the proposed analytical methods will 
continue to assure that the reactor operates safely. On that basis, 
the proposed changes do not involve an increase in the probability 
of an accident.
    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 and therefore, does not 
increase the likelihood of any failure mechanisms or precursors to 
transients or accidents postulated and analyzed in the Callaway 
Plant FSAR [Final Safety Analysis Report]. Operation of the reactor 
with core operating limits determined by use of the proposed 
analytical methods does not increase the reactor power level, does 
not increase the core fission product inventory, and does not change 
any radiological release assumptions. The proposed changes will not 
alter any accident analysis assumptions discussed in the FSAR, nor 
do they involve any changes to the requirement for Callaway Plant to 
operate within the power distribution limits and shutdown margins 
required by the TS and within the assumptions of the safety analyses 
described in the FSAR. Therefore the proposed methodology and TS 
changes do 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 provides revised analytical methods for 
determining core operating limits, and does not change any system 
functions or requirements. Acceptance criteria required to be met 
for analyzed core performance under normal, transient and accident 
conditions are not being changed, as the core operating limits will 
continue to be established in accordance with NRC-approved methods. 
The change does not involve physical alteration of the plant, as no 
new or different type of equipment will be installed. The change 
does not alter assumptions made in the safety analyses, but ensures 
that the core will operate within safe limits. Consequently, this 
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 the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are initiated. The proposed changes do not physically alter safety-
related systems, nor do they affect the way in which safety related 
systems perform their functions. The setpoints at which protective 
actions are initiated are not altered by the proposed changes. The 
availability of equipment

[[Page 163]]

required to be available to actuate upon demand for mitigating an 
analyzed event is unchanged by the proposed amendment. The proposed 
analytical methodologies are an improvement that allows more 
accurate modeling of core performance. The NRC has reviewed and 
approved the additional methodologies for use in lieu of the current 
methodology; thus, the margin of safety is not reduced due to this 
change.
    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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Robert J. Pascarelli.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.

    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: September 26, 2013, as supplemented by 
letters dated January 8, 2014; October 13, 2014; November 12, 2014; 
December 12, 2014; January 26, 2015; February 27, 2015; March 13, 2015; 
July 15, 2015; August 20, 2015; September 9, 2015; October 1, 2015; 
January 14, 2016; April 26, 2016; September 29, 2016; and November 21, 
2016.
    Brief description of amendments: The amendments revised the 
condition for the fire protection program (FPP) in Facility Operating 
Licenses such that the FPP is now based on the requirements of 10 CFR 
50.48(c), ``National Fire Protection Association Standard NFPA 805.''
    Date of issuance: December 6, 2016.
    Effective date: As of the date of issuance and shall be implemented 
as stated in the revised License Condition 2.C(4).
    Amendment Nos.: 291 and 270. A publicly-available version is in 
ADAMS under Accession No. ML16077A135; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. NPF-9 and NPF-17: 
Amendments revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: February 19, 2014 (79 
FR 9492). The supplemental letters dated January 8, 2014; October 13, 
2014; November 12, 2014; December 12, 2014; January 26, 2015; February 
27, 2015; March 13, 2015; July 15, 2015; August 20, 2015; September 9, 
2015; October 1, 2015; January 14, 2016; April 26, 2016; September 29, 
2016; and November 21, 2016, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposal 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 6, 2016.
    No significant hazards consideration comments received: No.

    Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal 
River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida.
    Date of application for amendment: July 28, 2015.
    Brief description of amendment: The amendment incorporates into the 
license the transfer of ownership, held by Seminole Electric 
Cooperative, Inc. (SEC), in CR-3 to DEF. The transfer of ownership will 
take place pursuant to the Settlement, Release and Acquisition 
Agreement, dated April 30, 2015, wherein DEF will purchase the 1.6994 
percent ownership share in CR-3 held by SEC, leaving DEF as the sole 
remaining licensee for CR-3.
    Date of issuance: November 30, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 251. A publicly-available version is in ADAMS under 
Accession No. ML16293A200; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the letter dated August 
10, 2016 (ADAMS Accession No. ML16173A022).
    Facility Operating License No. DPR-72: This amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: September 29, 2015 (80 
FR 58513), and January 4, 2016 (81 FR 98).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 10, 2016.
    No significant hazards consideration comments received: No.

    Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal 
River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida.
    Date of application for amendment: August 27, 2015, as supplemented 
by letters dated March 2, 2016, and July 14, 2016.
    Brief description of amendment: The amendment approved the CR-3 
Permanently Defueled Emergency Plan, and Permanently Defueled Emergency 
Action Level Bases Manual, for the Independent Spent Fuel Storage 
Installation.
    Date of issuance: December 5, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 252. A publicly-available version is in ADAMS under 
Accession No. ML16244A099; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.

[[Page 164]]

    Facility Operating License No. DPR-72: This amendment revised the 
License.
    Date of initial notice in Federal Register: November 10, 2015 (80 
FR 69711).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 5, 2016.
    No significant hazards consideration comments received: No.

    Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades 
Nuclear Plant, Van Buren County, Michigan.
    Date of amendment request: March 3, 2016, as supplemented by letter 
dated June 7, 2016.
    Brief description of amendment: The amendment approves the 
implementation of an alternate repair criteria (ARC) called C-star, for 
the portion of the steam generator (SG) tubes within the cold-leg 
tubesheet. In addition, the amendment clarifies the intent and improves 
the wording of the technical specifications regarding the previously 
incorporated ARC for the hot-leg side of the SG's tubesheet. This was 
previously approved by letter dated May 31, 2007, and Amendment No. 
225.
    Date of issuance: December 19, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 261. A publicly-available version is in ADAMS under 
Accession No. ML16300A030; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-20: Amendment revised 
the Renewed Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 2, 2016 (81 FR 
50747).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 19, 2016.
    No significant hazards consideration comments received: No.

    Southern Nuclear Operating Company, Inc.; Georgia Power Company; 
Oglethorpe Power Corporation; Municipal Electric Authority of Georgia; 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia.
    Date of amendment request: December 15, 2015, as supplemented by 
letter dated April 11, 2016.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) to incorporate risk-informed 
requirements regarding selected Required Action end states. 
Additionally, it modified TS Required Actions with a Note prohibiting 
the use of Limiting Condition for Operation Applicability 3.0.4.a when 
entering the preferred end state (Mode 3).
    Date of issuance: December 19, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 281 (Unit No. 1); 225 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML16257A724; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-57 and NPF-5: 
Amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: February 16, 2016 (81 
FR 7841). The supplemental letter dated April 11, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 19, 2016.
    No significant hazards consideration comments received: No.

    Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear 
Plant (WBN), Unit 1, Rhea County, Tennessee.
    Date of amendment request: February 23, 2016, as supplemented by 
letter dated July 22, 2016.
    Brief description of amendment: The amendment approved revisions to 
the WBN Dual Unit Fire Protection Report and revised the associated 
License Condition regarding the WBN fire protection program.
    Date of issuance: December 12, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 108. A publicly-available version is in ADAMS under 
Accession No. ML16307A013; 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.
    Date of initial notice in Federal Register: May 10, 2016 (81 FR 
28901). The supplemental letter dated July 22, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 12, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 21st day of December 2016.

    For the Nuclear Regulatory Commission.
George A. Wilson,
Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2016-31813 Filed 12-30-16; 8:45 am]
 BILLING CODE 7590-01-P