[Federal Register Volume 80, Number 217 (Tuesday, November 10, 2015)]
[Notices]
[Pages 69707-69719]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-28347]


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NUCLEAR REGULATORY COMMISSION

[NRC-2015-0253]


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 October 10, 2015, to October 26, 2015. The 
last biweekly notice was published on October 27, 2015.

DATES: Comments must be filed December 10, 2015. A request for a 
hearing must be filed by January 11, 2016.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0253. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected].
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.

[[Page 69708]]

    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: Lynn Ronewicz, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-2549, email: [email protected].

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2015-0253 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0253.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section of 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-2015-0253, facility name, unit 
number(s), application date, and subject in your comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, (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 should 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. Should the 
Commission take 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. Should the Commission make 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 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC's regulations are accessible electronically from the NRC Library on 
the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed within 60 days, the Commission or a presiding officer designated 
by the Commission or by the Chief Administrative Judge of the Atomic 
Safety and Licensing Board Panel, will rule on the request and/or 
petition; and the Secretary or the Chief Administrative Judge of the 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/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 requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.

[[Page 69709]]

    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall 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 requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion. The petition must include sufficient information to show that 
a genuine dispute exists with the applicant on a material issue of law 
or fact. Contentions shall be limited to matters within the scope of 
the amendment under consideration. The contention must be one which, if 
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements 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, 
and have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that person's admitted 
contentions, including the opportunity to present evidence and to 
submit a cross-examination plan for cross-examination of witnesses, 
consistent with NRC regulations, policies and procedures.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day 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)-(iii).
    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 decide 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 held 
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 any 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 
December 28, 2015. 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 for leave to intervene set forth in this section, except that 
under Sec.  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. A State, local governmental body, Federally-
recognized Indian tribe, or agency thereof may also have the 
opportunity to participate under 10 CFR 2.315(c).
    If a hearing is granted, any person who does not wish, or is not 
qualified, to become a party to the proceeding may, in 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 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. Persons desiring to make a limited 
appearance are requested to inform the Secretary of the Commission by 
December 28, 2015.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, 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 participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007). 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. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at http://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.

[[Page 69710]]

    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are 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 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to 
[email protected], or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 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 
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, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document 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 
http://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 home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, in some instances, a request to intervene will require 
including information on local residence in order to demonstrate a 
proximity assertion of interest in the proceeding. 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.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day 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)-(iii).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

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

    Date of amendment request: July 17, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15232A017.
    Description of amendment request: The proposed amendment corrects a 
usage problem with recently issued Amendment Nos. 382, 384, and 383 
(ADAMS Accession No. ML13231A013), which precludes Oconee Nuclear 
Station Technical Specification (TS) 3.8.1, ``AC [Alternating Current] 
Sources-Operating,'' Condition H from being used as planned. The 
proposed change revises the note to TS 3.8.1 Required Actions L.1, L.2, 
and L.3, to remove the 12-hour time limitation when the second Keowee 
Hydroelectric Unit (KHU) is made inoperable for the purpose of 
restoring the KHU undergoing maintenance to OPERABLE status. Removal of 
the 12-hour time limitation allows use of the full 60-hour Completion 
Time of Required Action H.2 when the unit(s) have been in Condition C 
for greater than 72 hours and both units are made inoperable for 
purposes of restoring the KHU undergoing maintenance to OPERABLE 
status.
    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 revises the note to Technical 
Specification (TS) 3.8.1 Required Actions L.1, L.2, and L.3 to 
indicate the Required Actions are not required when the Condition is 
entered to restore a KHU to OPERABLE status. This change is 
consistent with Amendment Nos. 382, 384, and 383, which approved a 
cumulative 240 hours of allowed outage time over a 3-year period 
when both KHUs are inoperable when in the 45-day Completion Time of 
TS 3.8.1 Required Action C.2.2.5. The proposed TS change does not 
modify the reactor coolant system pressure boundary, nor make any 
physical changes to the facility design, material, or construction 
standards. The probability of any design basis accident (DBA) is not 
affected by this change, nor are the consequences of any DBA 
affected by this change. The proposed change does not involve 
changes to any structures, systems, or components (SSCs) that can 
alter the probability for initiating a LOCA [loss-of-coolant 
accident] event.

[[Page 69711]]

    Therefore, the proposed TS changes do not significantly increase 
the probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed TS change revises the note to TS 3.8.1 Required 
Actions L.1, L.2, and L.3 to indicate the Required Actions are not 
required when the Condition is entered to restore a KHU to OPERABLE 
status. Revision of the note allows the 60 hour Completion Time of 
TS 3.8.1 Condition H to limit the time that both KHUs are 
inoperable. The changes do not alter the plant configuration (no new 
or different type of equipment will be installed) or make changes in 
methods governing normal plant operation. No new failure modes are 
identified, nor are any SSCs required to be operated outside the 
design bases.
    Therefore, the possibility of a new or different kind of 
accident from any kind of accident previously evaluated is not 
created.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed TS change revises the note to TS 3.8.1 Required 
Actions L.1, L.2, and L.3 to indicate the Required Actions are not 
required when the Condition is entered to restore a KHU to OPERABLE 
status. Revision of the note allows the 60 hour Completion Time of 
TS 3.8.1 Condition H to limit the time that both KHUs are 
inoperable. The proposed TS change does not involve: (1) A physical 
alteration of the Oconee Units; (2) the installation of new or 
different equipment; (3) operating any installed equipment in a new 
or different manner; (4) a change to any set points for parameters 
which initiate protective or mitigation action; or (5) any impact on 
the fission product barriers or safety limits.
    Therefore, this request 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: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Robert J. Pascarelli.

Duke Energy Florida, Inc., et al., Docket No. 50-302, Crystal River 
Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida

    Date of amendment request: August 27, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15246A231.
    Description of amendment request: The amendment would approve 
changes to the Permanently Defueled Emergency Plan (PDEP) to reflect 
the planned use of an Independent Spent Fuel Storage Installation 
(ISFSI) located in the Crystal River Unit 3 Nuclear Plant Protected 
Area while the spent fuel pool contains spent fuel assemblies.
    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 site PDEP and PD EAL [Permanently Defueled 
Emergency Action Level] Bases Manual revisions are commensurate with 
the ongoing and anticipated reduction in radiological source term at 
the CR-3 site and reflects the addition of spent fuel being 
transferred to the ISFSI facility. These changes add the 
responsibility for responding to ISFSI emergencies to the CR-3 PDEP 
Shift Supervisor/Certified Fuel Handler, and accompanying changes to 
the PD EAL Bases Manual due to the creation of a potential or actual 
release path to the environment, degradation of one or more storage 
canisters or fuel assemblies due to environmental factors, and 
configuration changes that could cause challenges in removing the 
canister or fuel from storage.
    There are no longer design basis accidents or postulated beyond 
design basis accidents that could result in doses to the public and 
the environment beyond the exclusion area boundary that would exceed 
the EPA PAGs [Protective Action Guidelines]. CR-3 was shut down on 
September 26, 2009, and will not be restarted. With the reactor 
permanently defueled, the spent fuel pool and its support systems 
are dedicated to spent fuel storage only. With the spent fuel in wet 
storage for some time, the spectrum of postulated accidents is much 
smaller than for an operational plant, with the majority of design 
basis accidents no longer possible. The only remaining credible 
design basis accident is the fuel handling accident, which does not 
result in exceeding the EPA Protective Action Guidelines at the 
exclusion area boundary. Spent fuel located in the spent fuel pools 
will be transferred to the ISFSI facility. Emergency Planning Zones 
beyond the exclusion area boundary and the associated protective 
actions are no longer required. No corporate personnel, personnel 
involved in off-site dose projections, or personnel with special 
qualifications are required to augment the ERO [Emergency Response 
Organization].
    The credible events for the ISFSI facility remain unchanged. The 
indications of damage to a loaded Dry Shielded Canister CONFINEMENT 
BOUNDARY have been revised to be twice the design basis dose rate as 
described in Draft Amendment 14 to COC [Certificate of Compliance] 
1004 Technical Specifications for the Standardized NUHOMS Horizontal 
Modular Storage System, Sections 5.2.4 `Radiation Protection 
Program' and 5.4.2 HSM [horizontal storage module] or HSM-H Dose 
Rate Evaluation Program (Reference 7), while in transit or HSM 
storage.
    Damage to Dry Shielded Canister CONFINEMENT BOUNDARY as 
indicated by the following on-contact radiation readings at some 
prescribed distance from the transfer cask or HSM:
    1300 mrem/hr (gamma + neutron) on the radial surface of the fuel 
transfer cask while in transit to the ISFSI HSM

OR
1050 mrem/hr (gamma + neutron)--HSM Front Bird Screen
4 mrem/hr (gamma + neutron)--HSM Outside Door
40 mrem/hr (gamma + neutron)--HSM End Shield Wall Exterior while in 
HSM storage.

    This change is consistent with industry practices previously 
approved by the NRC to distinguish whether a degraded containment 
barrier condition exists.
    The probability of occurrence of previously evaluated accidents 
is not increased, since most previously analyzed accidents can no 
longer occur and the probability of the remaining credible design 
basis accident is unaffected by the proposed amendment.
    The deletion of the Communicator position does not impact 
Emergency Notifications from the plant since the Emergency 
Coordinator has shown the capability to perform this function. This 
function is not involved in operations or evolutions that could 
cause an accident since it is not performed until after the 
emergency is declared, and has no effect on accident mitigation.
    Therefore, the proposed changes do not affect any plant system, 
the operation and maintenance of CR-3 and the ISFSI facility, or 
increase 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 have no impact on facility structures, 
systems, or components (SSCs) affecting the safe storage of 
irradiated fuel, or on the methods of operation of such SSCs, or on 
the handling and storage of irradiated fuel itself. Additionally, 
the proposed changes have no impact on a Fuel Handling Accident, 
which is the remaining credible design basis accident evaluated. The 
CR-3 PDEP is applicable for the plant's defueled condition. There is 
no impact on the prevention, diagnosis, or mitigation of reactor-
related transients as there are no longer any reactor-related 
accidents. Accidents cannot result in different or more adverse 
failure modes or accidents than previously evaluated because the 
reactor is permanently shut down and defueled, and CR-3 is no longer 
authorized to operate the reactor.
    There are no longer credible events that would result in doses 
to the public beyond the exclusion area boundary that would exceed 
the EPA [Environmental Protection

[[Page 69712]]

Agency] PAGs. Spent fuel waste will be transferred to the ISFSI 
facility. Emergency Planning Zones beyond the site boundary and the 
associated protective actions are no longer required. No corporate 
personnel, personnel involved in offsite dose projections, or 
personnel with special qualifications are required to augment the 
ERO.
    The credible events for the ISFSI facility remain unchanged. The 
indications of damage to a loaded Dry Shielded Canister CONFINEMENT 
BOUNDARY have been revised to be twice the design basis dose rate as 
described in Draft Amendment 14 to COC 1004 Technical Specifications 
for the Standardized NUHOMS Horizontal Modular Storage System, 
Sections 5.2.4 `Radiation Protection Program' and 5.4.2 HSM or HSM-H 
Dose Rate Evaluation Program (Reference 7), while in transit or HSM 
storage.
    Damage to Dry Shielded Canister CONFINEMENT BOUNDARY as 
indicated by the following on-contact radiation readings at some 
prescribed distance from the transfer cask or HSM:
    1300 mrem/hr (gamma + neutron) on the radial surface of the fuel 
transfer cask while in transit to the ISFSI horizontal storage 
module (HSM)

OR
1050 mrem/hr (gamma + neutron)--HSM Front Bird Screen
4 mrem/hr (gamma + neutron)--HSM Outside Door
40 mrem/hr (gamma + neutron)--HSM End Shield Wall Exterior while in 
HSM storage.

    This change is consistent with industry practices previously 
approved by the NRC to distinguish whether a degraded containment 
barrier condition exists. The proposed amendment does not introduce 
a new mode of plant operation or new accident pre-cursors, does not 
involve any physical alterations to plant configurations, or make 
changes to plant system set points that initiate a new or different 
kind of accident.
    The deletion of the Communicator position does not impact 
Emergency Notifications from the plant since the Emergency 
Coordinator has shown the capability to perform this function. This 
function is not involved in operations or evolutions that could 
cause or create new or different kinds of accidents since the 
communication of Emergency Notifications is not performed until 
after the emergency is declared and cannot affect an accident or 
event already in progress.
    Therefore, the proposed changes have no direct effect on any 
plant system, the operation and maintenance of CR-3 or the ISFSI 
facility, or create the possibility of a new or different kind of 
accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes have no direct effect on any plant system, 
do not involve any physical plant limit or parameter, License 
Condition, Technical Specification Limiting Condition of Operability 
or operating philosophy, and therefore cannot affect any margin of 
safety. The margin of safety is maintained by conforming to the CR-3 
Technical Specifications or the ISFSI Technical Specifications. The 
proposed CR-3 PDEP and PD EAL Bases Manual revisions are 
commensurate with the on-going and anticipated reduction in 
radiological source term at the CR-3 site and reflect spent fuel 
being transferred to the ISFSI facility. These changes add the 
responsibility for implementing the emergency plan for the ISFSI 
facility to the Shift Supervisor/Certified Fuel Handler.
    The only remaining credible accident for CR-3, while the SFP is 
operable and prior to the transference of all spent fuel to dry 
shielded canisters, is a fuel handling accident. The proposed 
amendment does not adversely affect the inputs or assumptions of any 
design basis analysis that impact the fuel handling accident. There 
are no longer credible events that would result in doses to the 
public beyond the exclusion area boundary that would exceed the EPA 
PAGs. Emergency Planning Zones beyond the exclusion area boundary 
and the associated protective actions are no longer required. No 
corporate personnel, personnel involved in offsite dose projections, 
or personnel with special qualifications are required to augment the 
ERO. The credible events for the ISFSI facility remain unchanged. 
The indications of damage to a loaded Dry Shielded Canister 
CONFINEMENT BOUNDARY have been revised to be twice the design basis 
dose rate as described in Draft Amendment 14 to COC 1004 Technical 
Specifications for the Standardized NUHOMS Horizontal Modular 
Storage System, Sections 5.2.4 `Radiation Protection Program' and 
5.4.2 HSM or HSM-H Dose Rate Evaluation Program (Reference 7), while 
in transit or HSM storage.
    Damage to Dry Shielded Canister CONFINEMENT BOUNDARY as 
indicated by the following on-contact radiation readings at some 
prescribed distance from the transfer cask or HSM:
    1300 mrem/hr (gamma + neutron) on the radial surface of the fuel 
transfer cask while in transit to the ISFSI HSM

OR
1050 mrem/hr (gamma + neutron)--HSM Front Bird Screen
4 mrem/hr (gamma + neutron)--HSM Outside Door
40 mrem/hr (gamma + neutron)--HSM End Shield Wall Exterior while in 
HSM storage.

    This change is consistent with industry practices previously 
approved by the NRC to distinguish whether a degraded containment 
barrier condition exists. The proposed changes are limited to the 
CR-3 PDEP and PD EAL Bases Manual and do not impact the safe storage 
of irradiated fuel. The proposed revisions do not affect any 
requirements for SSCs credited in the remaining analyses of 
applicable postulated accidents, and as such, do not affect the 
margin of safety associated with these accident analyses.
    The deletion of the Communicator position does not impact 
Emergency Notifications from the plant since the Emergency 
Coordinator has shown the capability to perform this function. This 
function is not involved in design basis analyses or operations that 
could cause any decrease in any previously analyzed safety margin.
    Therefore, the proposed changes do not create the possibility of 
reduction in any safety margin.

    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: Lara S. Nichols, 550 South Tryon Street, 
Charlotte NC 28202.
    NRC Branch Chief: Bruce A. Watson, CHP.

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

    Date of amendment request: September 8, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15258A185.
    Description of amendment request: The proposed amendment would 
replace the Technical Specification (TS) Figure 4.1-1, ``Site and 
Exclusion Area Boundaries and Low Population Zone,'' with a text 
description in TS 4.1, ``Site Location.'' In addition, a typographical 
error would be corrected from ``LGHR'' to ``LHGR'' [Linear Heat 
Generation Rate] in TS 1.1, ``Definitions.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change removes a figure, replaces that figure with 
a text description of the site location and corrects a typographical 
error. An administrative change such as this is not an initiator of 
any accident previously evaluated. As a result, the probability of 
an accident previously evaluated is not affected. The consequences 
of an accident with the incorporation of this administrative change 
are not different than the consequences of the same accident without 
this change. As a result, the consequences of an accident previously 
evaluated are not affected by this change.
    Based on the above, it is concluded that the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not modify the plant design, nor does 
the proposed change alter the operation of the plant or equipment 
involved in either routine plant operation or

[[Page 69713]]

in the mitigation of design basis accidents. The proposed change is 
administrative only.
    Based on the above, it is concluded that the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change consists of an administrative change to 
remove a figure, replace that figure with a text description of the 
site location, and correct a typographical error. The change does 
not alter the manner in which safety limits, limiting safety system 
settings, or limiting conditions for operation are determined. The 
safety analysis acceptance criteria are not affected by this change. 
The proposed change will not result in plant operation in a 
configuration outside of the design basis.
    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: Mr. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: July 24, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15246A408.
    Description of amendment request: The amendment would make 
editorial corrections to Technical Specification (TS) Section 1.4, 
``Frequency.'' Example 1.4-1 would be revised to be consistent with 
NRC-approved Industry Technical Specification Task Force (TSTF) 
Standard Technical Specification Change Traveler, TSTF-485, Revision 0, 
``Correct Example 1.4-1.'' In addition, Example 1.4-5 and Example 1.4-6 
would be revised to correct typographical errors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are editorial in nature and have no effect 
on accident scenarios previously evaluated. The proposed changes 
consist of editorial corrections to TS Section 1.4, ``Frequency,'' 
that would make the Duane Arnold Energy Center (DAEC) TS consistent 
with the Standard Technical Specifications for General Electric BWR/
4 Plants (NUREG-1433). The proposed changes do not affect initiating 
events for accidents previously evaluated and do not affect or 
modify plant systems or procedures used to mitigate the progression 
or outcome of those accident scenarios.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are editorial in nature consisting of 
editorial corrections to TS Section 1.4, ``Frequency.'' The proposed 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or a change in 
the methods governing normal plant operation. No new accident 
scenarios, failure mechanisms, or limiting single failures are 
introduced as a result of the proposed changes.
    The proposed changes do not introduce any new accident 
precursors, nor do they impose any new or different requirements or 
eliminate any existing requirements. The proposed changes do not 
alter assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to confidence in the ability of the 
fission product barriers (fuel cladding, reactor coolant system, and 
primary containment) to perform their design functions during and 
following postulated accidents. The proposed changes are editorial 
in nature consisting of editorial corrections to TS Section 1.4, 
``Frequency.'' No setpoints at which protective actions are 
initiated are altered by the proposed changes. The proposed changes 
do not alter the manner in which the safety limits are determined. 
These changes are consistent with plant design and do not change the 
TS operability requirements; thus, previously evaluated accidents 
are not affected by this proposed change.
    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: Mr. William Blair, P.O. Box 14000, Juno 
Beach, FL 33408-0420.
    NRC Branch Chief: David L. Pelton.

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

    Date of amendment request: August 6, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15246A410.
    Description of amendment request: The proposed amendment would 
resolve a 10 CFR part 21 condition concerning a potential to 
momentarily violate Reactor Core Safety Limit 2.1.1.1 during Pressure 
Regulator Failure Maximum Demand (Open) transient.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    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 reactor steam dome pressure from 785 
psig to 685 psig in TS [Technical Specification] SLs [Safety Limits] 
2.1.1.1 and 2.1.1.2 does not alter the use of the analytical methods 
used to determine the safety limits that have been previously 
reviewed and approved by the NRC. The proposed change is in 
accordance with an NRC approved critical power correlation 
methodology and as such maintains required safety margins. The 
proposed change does not adversely affect accident initiators or 
precursors nor does it alter the design assumptions, conditions, or 
configuration of the facility or the manner in which the plant is 
operated and maintained.
    The proposed change does not alter or prevent the ability of 
structures, systems, and components (SSCs) from performing their 
intended function to mitigate the consequences of an initiating 
event within the assumed acceptance limits. The proposed change does 
not require any physical change to any plant SSCs nor does it 
require any change in systems or plant operations. The proposed 
change is consistent with the 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 does not involve a physical alteration of 
the plant (i.e., no new or different type of equipment will be

[[Page 69714]]

installed) or a change in the methods governing normal plant 
operation. No new accident scenarios, failure mechanisms, or 
limiting single failures are introduced as a result of the proposed 
change.
    The proposed change does not introduce any new accident 
precursors, nor does it impose any new or different requirements or 
eliminate any existing requirements. The proposed change does not 
alter assumptions made in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is related to confidence in the ability of the 
fission product barriers (fuel cladding, reactor coolant system, and 
primary containment) to perform their design functions during and 
following postulated accidents. Evaluation of the 10 CFR part 21 
condition by General Electric determined that there was no decrease 
in the safety margin, the Minimum Critical Power Ratio improves 
during the transient, and therefore is not a threat to fuel cladding 
integrity.
    The proposed change to Reactor Core Safety Limits 2.1.1.1 and 
2.1.1.2 is consistent with, and within the capabilities of the 
applicable NRC approved critical power correlation, and thus 
continues to ensure that valid critical power calculations are 
performed. No setpoints at which protective actions are initiated 
are altered by the proposed change. The proposed change does not 
alter the manner in which the safety limits are determined. This 
change is consistent with plant design and does not change the TS 
operability requirements; thus, previously evaluated accidents are 
not affected by this proposed change.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

NextEra Energy, Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: June 12, 2015, as supplemented by 
letters dated August 11, 2015, and August 28, 2015. Publicly-available 
versions are in ADAMS under Accession Nos. ML15166A042, ML15223B277, 
and ML15240A017, respectively.
    Description of amendment request: The amendments would revise the 
Point Beach Emergency Plan, to increase the staff augmentation times 
for Emergency Response Organization (ERO) response functions, from 30 
and 60 minutes, to 60 minutes and 90 minutes, respectively. Additional 
changes include relocation of the Emergency Director and Emergency 
Action Level Monitor positions, and the addition of an Assistant 
Emergency Operations Facility Manager position.
    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 increase in staff augmentation times has no effect 
on normal plant operation or on any accident initiator or precursors 
and does not impact the function of plant structures, systems, or 
components (SCCs). The proposed change does not alter or prevent the 
ability of the ERO to perform their intended functions to mitigate 
the consequences of an accident or event. The ability of the ERO to 
respond adequately to radiological emergencies has been demonstrated 
as acceptable through a staffing analysis as required by 10 CFR 50 
Appendix E.IV.A.9.
    Therefore, the proposed Emergency Plan changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not impact the accident analysis. The 
change does not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed), a change in 
the method of plant operation, or new operator actions. The proposed 
change does not introduce failure modes that could result in a new 
accident, and the change does not alter assumptions made in the 
safety analysis. This proposed change increases the staff 
augmentation response times in the Emergency Plan, which are 
demonstrated as acceptable through a staffing analysis as required 
by 10 CFR 50 Appendix E.IV.A.9. The proposed change does not alter 
or prevent the ability of the ERO to perform their intended 
functions to mitigate the consequences of an accident or event.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel cladding, reactor coolant 
system pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed change is 
associated with the Emergency Plan staffing and does not impact 
operation of the plant or its response to transients or accidents. 
The change does not affect the Technical Specifications. The 
proposed change does not involve a change in the method of plant 
operation, and no accident analyses will be affected by the proposed 
change. Safety analysis acceptance criteria are not affected by this 
proposed change. The revised Emergency Plan will continue to provide 
the necessary response staff with the proposed change. A staffing 
analysis and a functional analysis were performed for the proposed 
change on the timeliness of performing major tasks for the 
functional areas of Emergency Plan. The analysis concluded that an 
extension in staff augmentation times would not significantly affect 
the ability to perform the required Emergency Plan tasks. Therefore, 
the proposed change is determined to not adversely affect the 
ability to meet 10 CFR 50.54(q)(2), the requirements of 10 CFR 50 
Appendix E, and the emergency planning standards as described in 10 
CFR 50.47(b).
    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 Blair, Managing Attorney--Nuclear, 
Florida Power & Light Company, P.O. Box 14000, 700 Universe Boulevard, 
Juno Beach, FL 33408-0420.
    NRC Branch Chief: David L. Pelton.

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 amendment request: September 16, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15259A576.
    Description of amendment request: The amendment would revise the 
Reactor Coolant System (RCS) minimum flow specified in Technical 
Specification (TS) 3.4.1, ``RCS Pressure, Temperature, and Flow 
Departure from Nucleate Boiling (DNB) Limits.'' The proposed change is 
necessary to correct a non-conservative TS value for DCPP, Unit 1. The 
Unit 1 RCS flow specified in TS 3.4.1 for 100 percent power is 359,000 
gallons per minute (gpm). However, the TS value is less than the 
359,200 gpm RCS minimum measured flow (MMF) value specified in the 
Updated Final Safety Analyses Report

[[Page 69715]]

(UFSAR) Table 4.1-1, ``Reactor Design Comparison.'' The UFSAR RCS MMF 
value represents the RCS flow value used in the reactor core DNB safety 
analyses. This issue has been entered in the DCPP corrective action 
program, and the actual Unit 1 RCS flow value has been verified to be 
within the limits required by the applicable safety analyses.
    In order to resolve the non-conservative TS value, the proposed 
change would revise the RCS flow requirements in DCPP TS 3.4.1 to be 
consistent with TS 3.4.1 in NUREG-1431, Revision 4, Volume 1, 
``Standard Technical Specifications--Westinghouse Plants,'' April 2012 
(ADAMS Accession No. ML12100A222). The proposed change to the RCS flow 
requirements in TS 3.4.1 would also be consistent with the NRC-approved 
Technical Specification Task Force (TSTF) Traveler-339-A, Revision 2, 
``Relocate TS Parameters to [Core Operating Limits Report] COLR,'' and 
NRC-approved WCAP-14483-A, ``Generic Methodology for Expanded Core 
Operating Limits Report,'' dated June 13, 2000 (ADAMS Accession No. 
ML003723269).
    The proposed change would delete the current DCPP, Units 1 and 2 TS 
3.4.1 RCS flow Tables 3.4.1-1 and 3.4.1-2, and would add the DCPP, 
Units 1 and 2 RCS thermal design flow values of 350,800 gpm and 354,000 
gpm, respectively, to the requirements of TS 3.4.1. In addition, the 
proposed change would add the RCS MMF values of 359,200 gpm and 362,500 
gpm, to the DCPP, Units 1 and 2 COLR, respectively. Consistent with the 
Standard Technical Specifications (STS), the proposed change would also 
include a reference to the RCS COLR flow requirements in the TS 3.4.1 
Limiting Condition for Operation and Surveillance Requirements. Due to 
the elimination of RCS flow Tables 3.4.1-1 and 3.4.1-2, a reference to 
these tables is also deleted from Figure 2.1.1-1, ``Reactor Core Safety 
Limit.''
    As such, the proposed change would resolve the non-conservative TS 
value for Unit 1 and serve to make the DCPP, Units 1 and 2 TS more 
consistent with the STS in NUREG-1431.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the DCPP Unit 1 and Unit 2 RCS flow 
requirements in TS 3.4.1, ``RCS Pressure, Temperature, and Flow 
Departure from Nucleate Boiling (DNB) Limits,'' to be more 
consistent with TS 3.4.1 in NUREG-1431 and with the applicable DCPP 
safety analyses. The proposed RCS flow values will ensure the 
assumptions of the safety analyses continue to be met.
    As such, the proposed change does not affect the design or 
function of any plant structures, systems, and components (SSCs). 
Thus, the proposed change does not affect plant operation, design 
features, or any analysis that verifies the capability of an SSC to 
perform a design function. As the proposed change is consistent with 
the RCS flow assumptions of the safety analyses, the proposed change 
does not affect any previously evaluated accidents in the UFSAR. In 
addition, the proposed change does not affect any SSCs, operating 
procedures, and administrative controls which have the function of 
preventing or mitigating any accident previously evaluated in the 
UFSAR.
    The proposed change will not alter any accident analyses 
assumptions discussed in the UFSAR and will continue to assure the 
DCPP units operate within the assumptions of the applicable safety 
analyses described in the UFSAR.
    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 accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the DCPP Unit 1 and Unit 2 RCS flow 
requirements in TS 3.4.1, ``RCS Pressure, Temperature, and Flow 
Departure from Nucleate Boiling (DNB) Limits,'' to be more 
consistent with TS 3.4.1 in NUREG-1431 and with the applicable DCPP 
safety analyses. The proposed RCS flow values will ensure the 
assumptions of the safety analyses continue to be met.
    The proposed change does not change any system functions or 
maintenance activities. The change does not involve physical 
alteration of the plant, that is, no new or different type of 
equipment will be installed. The proposed change involves no 
physical plant modification or changes in plant operation, therefore 
no new failure modes are created. As such, the proposed change does 
not create new failure modes or mechanisms that are not identifiable 
during testing, and no new accident precursors are generated.
    Therefore, the proposed change does not create the possibility 
of a new or different 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 change does not physically alter safety-
related systems, nor does it 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 change. 
Therefore, sufficient equipment remains available to actuate upon 
demand for the purpose of mitigating an analyzed event. The proposed 
RCS flow value changes are consistent with the plant safety 
analyses. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, CA 94120.
    NRC Branch Chief: Michael T. Markley.

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

    Date of amendment request: August 20, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15236A018.
    Description of amendment request: The proposed amendment would 
revise Appendix 3A of the Updated Final Safety Analysis Report to more 
fully reflect the permanently shutdown status of the SONGS, Units 2 and 
3. The revision would include a limited set of exceptions and 
clarifications to referenced Regulatory Guides to reflect the 
significantly reduced decay heat loads in the SONGS, Units 2 and 3 
Spent Fuel Pools and to support corresponding design basis changes and 
modifications that will allow for the implementation of the ``cold and 
dark'' strategy outlined in the SONGS Post-Shutdown Decommissioning 
Activities Report (PSDAR).
    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?
    The only accident previously evaluated, is the Spent Fuel Pool 
Boiling Event. The initiating event (loss of cooling) would no 
longer lead to a rapid increase in pool temperature to the boiling 
point or to a relatively short-term reduction in pool level due to 
evaporative losses. Currently a loss of

[[Page 69716]]

cooling would lead to a very slow heat-up toward the boiling point 
taking at least a week or more. From that point the slower 
evaporative losses would take several weeks to reduce inventory to 
unacceptable levels.
    The most likely cause of a loss of function of the Spent Fuel 
Pool Cooling System (SFPCS) is not a failure of components in the 
cooling system, but instead a loss of electrical power. The 
probability of a loss of power is substantially higher than the 
probability of a contemporaneous common cause failure of active 
components in the cooling system. For example, NRC has collected 
operating experience on loss of Spent Fuel Pool (SFP) cooling for 
nuclear plants in the U.S., which includes both safety-related and 
non-safety-related cooling systems. As indicated in NUREG-1275, 
Volume 12, the causes of loss of SFP cooling were the loss of the 
SFP cooling pumps due to loss of electrical power (39 of 56 events), 
loss of suction from the spent fuel pool, flow blockage, loss of the 
heat sink, and one case of inadequate configuration control. As 
concluded by the NRC: ``The dominant cause of the actual loss of SFP 
cooling events was loss of electrical power to the SFP cooling 
pumps.'' There were no cases involving a common cause failure mode, 
such as seismic events or tornados. Given this operating experience, 
any increase in the probability of a spent fuel pool boiling event 
due to the seismic re-classification of the system would be minimal 
in comparison to the failure rate due to loss of electrical power.
    The change in commitment does not affect the consequences of the 
spent fuel pool boiling accident (which by definition assumes loss 
of the spent fuel pool cooling system). Revised dose calculations 
were completed to support the changes to the Updated Final Safety 
Analysis Report (UFSAR) Chapter 15 Accident Analysis, and the UFSAR 
was revised to reflect the new analysis. These were recently 
reviewed to verify they remain bounding for the much slower event, 
even if it is not terminated (through restored cooling or adequate 
make-up) prior to reaching levels approaching the top of the stored 
fuel. This re-evaluation confirmed the doses previously calculated 
remain bounding and several orders of magnitude below applicable 
limits.
    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?
    The only accident relevant to this proposed change would be an 
unmitigated Spent Fuel Pool Boiling Event (i.e., boiling without 
restoration of cooling or make-up prior to uncovering of the spent 
fuel). The initiating event (loss of cooling) would no longer lead 
to a rapid increase in pool temperature to the boiling point and a 
relatively short-term reduction in pool level due to evaporative 
losses. Currently a loss of cooling would lead to a very slow heatup 
toward the boiling point taking at least a week or more. From that 
point the slower evaporative losses would take several weeks to 
reduce inventory to unacceptable levels. The only safety function 
remaining relates to maintaining the fuel cladding in the SFP 
(cooling is not a safety-related function as defined in the updated 
Chapter 15 Fuel Pool Boiling Accident Analysis, only maintaining 
water level--Reference 6.12). The only remaining safety related SSCs 
at SONGS Units 2 and 3 are the Spent Fuel Pool and related 
structural components (pool liner, structure, and racks).
    The Make-up System will ensure that sufficient water is supplied 
to the SFPs in the event of loss of cooling. In addition to the 
Seismic Category I make-up source, currently there are numerous 
other diverse sources of make-up for the SFPs, including:
     As provided in SONGS Units 2 and 3 procedures, the 
Nuclear Service Water connections located on the SFP operating level 
can be used via hoses to fill the pool. These connections are QC 
III, Seismic Category II.
     As provided in SONGS Units 2 and 3 Mitigation 
Strategies, water from Fire Water Tanks T-102 and T-103 via Fire 
Pumps P-220 (diesel driven), P-221 or P-222 (both of which are motor 
driven) can be provided through the installed fire system piping to 
two fire hose cabinets located on the Spent Fuel Pool Operating 
level. The tanks, pumps and piping are QC III-EPS and Seismic 
Category II.
     As provided in SONGS Units 2 and 3 Mitigation 
Strategies, make-up to the SFPs can be provided using water from one 
or more of the following sources: Demineralized Water Tanks T-266, 
T-267 or T-268, all are located at a higher elevation at the Make-up 
Demineralizer Area at the south end of the plant. Skid mounted pump 
P-i1058 delivers water from these sources to the seismic standpipe 
and from the standpipe to the SFP. T-266, T-267 and T-268 are QC 
III, Seismic Category II. P-1058 is QC III-EPS and Seismic Category 
III.
     As discussed in SONGS Units 2 and 3 Mitigation 
Strategies, the 10'' City Water Line Supply Line can be used as an 
alternate source of SFP make-up water.
     Another make-up path is available using the Seismic 
Category I Demineralized Water Storage Tank (T-351) located in the 
North Industrial Area along with Seismic Category I portable diesel 
driven Fire Pump (P-i1065) using strategically staged hoses between 
the tank, pump, Seismic Category I standpipe and the Spent Fuel 
Pool. The hoses are pressure tested annually and are inspected for 
location quarterly per SONGS Units 2 and 3 procedures.
    The Mitigation Strategies are sequenced to assure the strategies 
can be deployed in 2 hours or less. The capability to achieve this 
time requirement was evaluated in a formal study and further 
demonstrated in the field using actual staff, procedures and 
equipment.
    Given the number and diversity of make-up sources, and the time 
available to supply make-up to the SFPs in the loss of spent fuel 
pool cooling, it is not credible to postulate a complete loss of 
make-up to a SFP. As discussed in NRC's June 30, 2014, letter 
concerning San Onofre Nuclear Generating Station, Units 2 and 3--
Rescission of Order EA-12-049:
    [T]he time to boil off water inventory in the SFP to a level of 
10 feet above the spent fuel will be sufficiently long to obviate 
the need for additional strategies to restore SFP cooling. The NRC 
staff concludes that given the low decay heat levels and the long 
time to boil off, the reliance on the SFP inventory for passive 
cooling provides an equivalent level of protection as that which 
would be provided by the initial phase of the guidance and 
strategies for maintaining or restoring SFP cooling capabilities 
that would be necessary for compliance with Order EA-12-049 using 
installed equipment. The staff further concludes that the long time 
to boil off the SFP inventory to a point at which make-up would be 
necessary for radiation shielding purposes obviates the need for the 
transition phase of the guidance and strategies that would be 
necessary for compliance with Order EA-12-049 using on-site portable 
equipment. The staff also concludes that the low decay heat and long 
boil-off period provides sufficient time for the licensee to obtain 
off-site resources on an ad hoc basis to sustain the SFP cooling 
function indefinitely, obviating the need for the final phase of the 
guidance and strategies that would be necessary for compliance with 
Order EA-12-049.
    Similarly, as described in NRC's 2015 exemption from certain 
emergency planning requirements for SONGS Units 2 and 3:
    Additionally, in its letters to the NRC dated October 6, 2014, 
and December 15, 2014, SCE described the SFP make-up strategies that 
could be used in the event of a catastrophic loss of SFP inventory. 
The multiple strategies for providing make-up water to the SFP 
include: Using existing plant systems for inventory make-up; an 
internal strategy that relies on installed fire water pumps and 
service water or fire water storage tanks; or an external strategy 
that uses portable pumps to initiate make-up flow into the SFPs 
through a seismic standpipe and standard fire hoses routed to the 
SFPs or to a spray nozzle. These strategies will continue to be 
required as a license condition. Considering the very low 
probability of beyond-design-basis accidents affecting the SFP, 
these diverse strategies provide defense-in-depth and time to 
provide additional make-up or spray water to the SFP before the 
onset of any postulated off-site radiological release.
    It is not necessary to postulate both a loss of spent fuel pool 
cooling in conjunction with a loss of spent fuel pool make-up, and 
such an event is not postulated in UFSAR Section 15.7.3.8 related to 
SFP boiling and is not credible given the number of diverse sources 
of make-up and the time available to supply make-up.
    As currently discussed in UFSAR 9.1.2.3, spent fuel pool boiling 
also will not adversely affect the integrity of the SFPs. The 
reinforced concrete temperature differences and gradients were 
determined based on an inside face temperature of 230[emsp14][deg]F 
(water temperature of 212[emsp14][deg]F and gamma heating of 
18[emsp14][deg]F). That analysis indicates that the SFP walls have 
sufficient structural capability to accommodate this thermal 
loading.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?

[[Page 69717]]

    The proposed changes do not alter any design basis or safety 
limits for the plant. The applicable limits are spent fuel clad 
temperature and spent fuel pool level. The spent fuel cladding 
temperature is assured by maintaining water level to support natural 
circulation cooling within the spent fuel racks. Forced cooling 
keeps evaporative losses and Fuel Handling Building environs within 
nominal limits. Thus, the SSCs that support the design and safety 
limits are limited to those that maintain inventory (Spent Fuel Pool 
and related structural components (pool liner, structure, and racks) 
and sufficient equipment to replace evaporative or other losses. 
Complete loss of make-up is not credible given the existence of 
numerous sources of make-up and the time available to provide make-
up. No changes to the pool and its structures are proposed and make-
up capability remains assured.
    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: Walker A. Matthews, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, CA 
91770.
    NRC Branch Chief: Bruce Watson.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-321 and 50-
366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, GA

    Date of amendment request: August 4, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15216A602.
    Description of amendment request: The licensee describes the 
application as follows: ``This amendment corrects an obvious 
typographical error in the Unit 1 FOL [Facility Operating License], and 
on page 5.0.17 of the Unit 2 TS [Technical Specification]. The Degraded 
Voltage Protection license condition in Part 2.C of the Unit 1 FOL 
(DPR-57) is currently listed as condition number 10, whereas it should 
be listed as condition number 11. In addition, this paragraph should be 
further indented to the right, to clarify that it's a third level 
paragraph (i.e. level 2.C.11). In addition to the FOL change, this 
amendment corrects an incorrect Unit number in Hatch Unit 2 TS page 
5.0.17. This page was inadvertently sent and issued stating Unit 1 on 
the bottom left, whereas it should clearly state Unit 2. Lastly, this 
amendment adds the term STAGGERED TEST BASIS to the Definitions section 
of the Unit 1 and Unit 2 TS. This term was removed from the TS and 
moved to the Surveillance Frequency Control Program (SFCP) when the NRC 
issued the TSTF-425 license amendment in [January 3,] 2012 to relocate 
specific surveillance frequency requirements to a licensee controlled 
program. This term, however, was reintroduced into Section 5 of the TS 
as a defined term when Hatch adopted the Control Room Envelope 
Habitability Program (TSTF-448) [in an amendment issued on August 29, 
2014]. Since it's currently used as a defined term in Section 5 of the 
TS, it needs to be included in the Definitions section of the TS.''
    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 contains no technical changes; all 
proposed changes are administrative. These changes are consistent 
with the intent of what has already been approved by the Nuclear 
Regulatory Commission (NRC).
    There are no accidents affected by this change, and therefore no 
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 contains no technical changes; all 
proposed changes are administrative. These changes are consistent 
with the intent of what has already been approved by the Nuclear 
Regulatory Commission (NRC).
    There are no accidents affected by this change, and therefore no 
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 amendment contains no technical changes; all 
proposed changes are administrative. These changes are consistent 
with the intent of what has already been approved by the Nuclear 
Regulatory Commission (NRC).
    There are no accidents affected by this change, and therefore no 
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, 40 Inverness Center 
Parkway, Birmingham, AL 35201.
    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.

Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, 
Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut

[[Page 69718]]

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
    Date of amendment request: November 17, 2014, as supplemented by 
letter dated August 13, 2015.
    Brief description of amendments: The amendments revised the Cyber 
Security Plan (CSP) Milestone 8 full implementation date as set forth 
in the CSP Implementation Schedule for the following plants: Kewaunee 
Power Station; Millstone Power Station, Unit Nos. 2 and 3; North Anna 
Power Station, Unit Nos. 1 and 2; and Surry Power Station, Unit Nos. 1 
and 2.
    Date of issuance: October 7, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 216, 323, 269, 276, 258, 286, and 286. A publicly-
available version is in ADAMS under Accession No. ML15245A482. 
Documents related to these amendment are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-43, DPR-65, DPR-49, 
NPF-4, NPF-7, DPR-32, and DPR-37: Amendments revised the Facility 
Operating Licenses.
    Date of initial notice in Federal Register: May 5, 2015 (80 FR 
25718). The supplement letter dated August 13, 2015, 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.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 7, 2015.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 10, 2015, as supplemented by 
letters dated September 30 and October 20, 2015.
    Brief description of amendment: The amendment approved a one-time 
extension of the Technical Specification (TS) completion time 
associated with the Division 2 Shutdown Service Water Subsystem from 72 
hours to 7 days in support of maintenance activities.
    Date of issuance: October 22, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No: 207. A publicly-available version is in ADAMS under 
Accession No. ML15280A258; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-62: The amendment revised the 
TSs and License.
    Date of initial notice in Federal Register: September 18, 2015 (80 
FR 56498).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 22, 2015.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373, LaSalle County 
Station, Unit 1 and Unit 2, LaSalle County, Illinois

    Date of amendment request: January 12, 2015.
    Brief description of amendments: The amendments deleted the 
limiting condition for operation (LCO) Note for Technical Specification 
(TS) Section 3.5.1, ``ECCS [emergency core cooling system]--
Operating.'' The current Note allowed the licensee to consider the low 
pressure coolant injection subsystem associated with the residual heat 
removal system to be OPERABLE under specified conditions.
    Date of issuance: October 14, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 217 and 203. A publicly-available version is in 
ADAMS under Accession No. ML15244B410; documents related to this 
amendment are listed in the Safety Evaluation enclosed with the 
amendments.
    Facility Operating License Nos. NPF-11 and NPF-18: Amendments 
revised the Facility Operating License and TSs.
    Date of initial notice in Federal Register: March 31, 2015 (80 FR 
17091).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 14, 2015.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of amendment request: December 19, 2014, as supplemented by 
letter dated June 26, 2015.
    Brief description of amendment: This amendment revised the 
technical specifications (TSs) to adopt performance-based Type C 
testing for the reactor containment, which would allow for extended 
test intervals for Type C valves, and corrects an editorial issue in 
the TSs.
    Date of issuance: October 9, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days from the date of issuance.
    Amendment No.: 288. A publicly-available version is in ADAMS under 
Accession No. ML15239B293; documents related to this amendment are 
listed in the Safely Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-3: Amendment revised the 
Facility Operating License and TSs.
    Date of initial notice in Federal Register: March 31, 2015 (80 FR 
17090), and July 7, 2015 (80 FR 38759). The supplemental letter dated 
June 26, 2015, provided additional information that clarified the 
application, did not expand the scope of the application as previously 
noticed, and did not change the staff's 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 October 9, 2015.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of amendment request: December 30, 2014.
    Brief description of amendment: This amendment revises the 
technical specification (TS) surveillance requirement for the frequency 
to verify that each containment spray system nozzle is unobstructed 
from every 10 years to an event-based frequency.
    Date of issuance: October 20, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days from the date of issuance.
    Amendment No.: 289. A publicly-available version is in ADAMS under 
Accession No. ML15251A046; documents related to this amendment

[[Page 69719]]

are listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-3: Amendment revised the 
Facility Operating License and TSs.
    Date of initial notice in Federal Register: March 31, 2015 (80 FR 
17090).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 20, 2015.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of amendment request: June 30, 2014, as supplemented March 27, 
2015.
    Brief description of amendment: The amendment revised the Humboldt 
Bay Power Plant, Unit 3 License to approve the revised Emergency Plan.
    Date of issuance: September 23, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 46. A publicly-available version is in ADAMS under 
Accession No. ML15148A361; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-7: Amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: August 19, 2014 (79 FR 
49109).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 23, 2015.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, 
Fairfield County, South Carolina

    Date of amendment request: July 22, 2015.
    Brief description of amendment: The amendment revised Technical 
Specification Section 6.0, ``Administrative Controls,'' by changing the 
``Shift Supervisor'' title to ``Shift Manager.''
    Date of issuance: October 15, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 202. A publicly-available version is in ADAMS under 
Accession No. ML15208A029; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-12: Amendment revised 
the Renewed Facility Operating License.
    Date of initial notice in Federal Register: August 14, 2015 (80 FR 
48924), as corrected by Federal Register notice dated August 20, 2015 
(80 FR 50663).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 15, 2015.
    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: June 17, 2015, as supplemented by 
letters dated July 14, August 3, August 28, September 3, and September 
21, 2015.
    Brief description of amendment: The amendment adopted new Technical 
Specification (TS) 3.7.16, ``Component Cooling System (CCS)--
Shutdown,'' and TS 3.7.17, ``Essential Raw Cooling Water (ERCW) 
System--Shutdown,'' and revised TS 3.3.2, ``Engineered Safety Feature 
Actuation System (ESFAS) Instrumentation,'' and TS 3.4.6, ``RCS Loops-
MODE 4,'' to support dual-unit operation of WBN Units 1 and 2.
    Date of issuance: October 20, 2015.
    Effective date: As of the date of issuance and shall be implemented 
after the issuance of the Facility Operating License for Unit 2.
    Amendment No.: 104. A publicly-available version is in ADAMS under 
Accession No. ML15275A042; documents related to this amendment are 
listed in the Safety Evaluation (SE) enclosed with the amendment.
    Facility Operating License No. NPF-90: Amendment revised the 
Facility Operating License and TSs.
    Date of initial notice in Federal Register: July 17, 2015 (80 FR 
42552). The supplemental letters dated July 14, August 3, August 28, 
September 3, and September 21, 2015, provided additional information 
that clarified the application. These supplements did not change the 
staff's proposed no significant hazards consideration. The supplemental 
letter dated September 3, 2015, provided additional information that 
expanded the scope of the application as originally noticed. A notice 
published in the Federal Register on September 15, 2015 (80 FR 55383), 
supersedes the original notice in its entirety to update the expanded 
scope of the amendment description and include the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in an SE dated October 20, 2015.
    No significant hazards consideration determination comments 
received: No.

    Dated at Rockville, Maryland, this 2nd day of November, 2015.
    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2015-28347 Filed 11-9-15; 8:45 am]
 BILLING CODE 7590-01-P