[Federal Register Volume 82, Number 107 (Tuesday, June 6, 2017)]
[Notices]
[Pages 26128-26144]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-11679]


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NUCLEAR REGULATORY COMMISSION

[NRC-2017-0131]


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 May 9, 2017, to May 22, 2017. The last 
biweekly notice was published on May 23, 2017.

DATES: Comments must be filed by July 6, 2017. A request for a hearing 
must be filed by August 7, 2017.

ADDRESSES: You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0131. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: TWFN-8-D36M, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-5411; email: [email protected].

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2017-0131, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0131.
     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 this 
document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2017-0131, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

[[Page 26129]]

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

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

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

    Within 60 days after the date of publication of this notice, any 
persons (petitioner) whose interest may be affected by this action may 
file a request for a hearing and petition for leave to intervene 
(petition) with respect to the action. Petitions shall be filed in 
accordance with the Commission's ``Agency Rules of Practice and 
Procedure'' in 10 CFR part 2. Interested persons should consult a 
current copy of 10 CFR 2.309. The NRC's regulations are accessible 
electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of 
the regulations is available at the NRC's Public Document Room, located 
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. If a petition is filed, the 
Commission or a presiding officer will rule on the petition and, if 
appropriate, a notice of a hearing will be issued.
    As required by 10 CFR 2.309(d) the petition should specifically 
explain the reasons why intervention should be permitted with 
particular reference to the following general requirements for 
standing: (1) The name, address, and telephone number of the 
petitioner; (2) the nature of the petitioner's right under the Act to 
be made a party to the proceeding; (3) the nature and extent of the 
petitioner's property, financial, or other interest in the proceeding; 
and (4) the possible effect of any decision or order which may be 
entered in the proceeding on the petitioner's interest.
    In accordance with 10 CFR 2.309(f), the petition must also set 
forth the specific contentions which the petitioner seeks to have 
litigated in the proceeding. Each contention must consist of a specific 
statement of the issue of law or fact to be raised or controverted. In 
addition, the petitioner must provide a brief explanation of the bases 
for the contention and a concise statement of the alleged facts or 
expert opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to the specific sources and 
documents on which the petitioner intends to rely to support its 
position on the issue. The petition must include sufficient information 
to show that a genuine dispute exists with the applicant or licensee on 
a material issue of law or fact. Contentions must be limited to matters 
within the scope of the proceeding. The contention must be one which, 
if proven, would entitle the petitioner to relief. A petitioner who 
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at 
least one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene. 
Parties have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that party's admitted 
contentions, including the opportunity to present evidence, consistent 
with the NRC's regulations, policies, and procedures.
    Petitions must be filed no later than 60 days from the date of 
publication of this notice. Petitions and motions for leave to file new 
or amended contentions that are filed after the deadline will not be 
entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in 
accordance with the filing instructions in the ``Electronic Submissions 
(E-Filing)'' section of this document.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to establish when the hearing is held. If the final determination is 
that the amendment request involves no significant hazards 
consideration, the Commission may issue the amendment and make it 
immediately effective, notwithstanding the request for a hearing. Any 
hearing would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of the amendment unless the Commission finds an imminent 
danger to the health or safety of the public, in which case it will 
issue an appropriate order or rule under 10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission by 
August 7, 2017. The petition must be filed in accordance with the 
filing instructions in the ``Electronic Submissions (E-Filing)'' 
section of this document, and should meet the requirements for 
petitions set forth in this section, except that under 10 CFR 
2.309(h)(2) a State, local governmental body, or federally recognized 
Indian Tribe, or agency thereof does not need to address the standing 
requirements in 10 CFR 2.309(d) if the facility is located within its 
boundaries. Alternatively, a State,

[[Page 26130]]

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

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing and petition for leave to intervene (petition), any 
motion or other document filed in the proceeding prior to the 
submission of a request for hearing or petition to intervene, and 
documents filed by interested governmental entities that request to 
participate under 10 CFR 2.315(c), must be filed in accordance with the 
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 
46562, August 3, 2012). The E-Filing process requires participants to 
submit and serve all adjudicatory documents over the internet, or in 
some cases to mail copies on electronic storage media. Detailed 
guidance on making electronic submissions may be found in the Guidance 
for Electronic Submissions to the NRC and on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit 
paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at 301-415-1677, to (1) request a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign submissions and access the E-Filing 
system for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a petition or 
other adjudicatory document (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a 
digital ID certificate and a docket has been created, the participant 
can then submit adjudicatory documents. Submissions must be in Portable 
Document Format (PDF). Additional guidance on PDF submissions is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the 
time the document is submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
document on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before adjudicatory documents are 
filed so that they can obtain access to the documents via the E-Filing 
system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public Web site at http://www.nrc.gov/site-help/e-submittals.html, by 
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 6 p.m., Eastern Time, Monday through Friday, excluding government 
holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
stating why there is good cause for not filing electronically and 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff. Participants filing adjudicatory documents in this 
manner are responsible for serving the document on all other 
participants. Filing is considered complete by first-class mail as of 
the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service. A presiding officer, having granted an 
exemption request from using E-Filing, may require a participant or 
party to use E-Filing if the presiding officer subsequently determines 
that the reason for granting the exemption from use of E-Filing no 
longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the 
Commission or the presiding officer. If you do not have an NRC-issued 
digital ID certificate as described above, click cancel when the link 
requests certificates and you will be automatically directed to the 
NRC's electronic hearing dockets where you will be able to access any 
publicly-available documents in a particular hearing docket. 
Participants are requested not to include personal privacy information, 
such as social security numbers, home addresses, or personal phone 
numbers in their filings, unless an NRC regulation or other law 
requires submission of such information. For example, in some 
instances, individuals provide home addresses in order to demonstrate 
proximity to a facility or site. With respect to copyrighted works, 
except for limited excerpts that serve the purpose of the adjudicatory 
filings and would constitute a Fair Use application, participants are 
requested not to include copyrighted materials in their submission.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document,

[[Page 26131]]

see the ``Obtaining Information and Submitting Comments'' section of 
this document.
Energy Northwest, Docket No. 50-397, Columbia Generating Station 
(Columbia), Benton County, Washington
    Date of amendment request: March 27, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17086A586.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) for Columbia and proposes 
changes to the containment leakage rate testing programs of Type A, B 
and C. These tests are required by TS 5.5.12, ``Primary Containment 
Leakage Rate Testing Program,'' and these changes would adopt the more 
conservative allowable test internal extension of Nuclear Energy 
Institute (NEI) 94-01, Revision 3-A and also adopt American National 
Standards Institute/American Nuclear Society 56.8-2002, ``Containment 
System Leakage Testing Requirements.''
    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 activities involve the revision of Columbia 
Generating Station (Columbia) Technical Specification (TS) 5.5.12 to 
allow the extension of the Type A containment test interval to 15 
years, and the extension of the Type C test interval to 75 months. 
The current Type A test interval of 120 months (10 years) would be 
extended on a permanent basis to no longer than 15 years from the 
last Type A test. The current Type C test interval of 60 months for 
selected components would be extended on a performance basis to no 
longer than 75 months. Extensions of up to nine months (total 
maximum interval of 84 months for Type C tests) are permissible only 
for non-routine emergent conditions.
    The proposed extensions do not involve either a physical change 
to the plant or a change in the manner in which the plant is 
operated or controlled. The containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. As such, 
the containment and the testing requirements invoked to periodically 
demonstrate the integrity of the containment exist to ensure the 
plant's ability to mitigate the consequences of an accident, and do 
not involve the prevention or identification of any precursors of an 
accident.
    The change in Type A test frequency to once-per-fifteen-years, 
measured as an increase to the total integrated plant risk for those 
accident sequences influenced by Type A testing, is 2.77E-4 person-
rem [roentgen equivalent man]/yr (a 0.00761% increase). EPRI 
[Electric Power Research Institute] Report No. 1009, Revision 2-A 
states that a very small population dose is defined as an increase 
of less than 1.0 person-rem per year or less than 1 percent of the 
total population dose, whichever is less restrictive for the risk 
impact assessment of the extended ILRT [integrated leakage rate 
test] intervals. Moreover, the risk impact when compared to other 
severe accident risks is negligible. Therefore, the proposed 
extension does not involve a significant increase in the probability 
of an accident previously evaluated.
    In addition, as documented in NUREG-1493, ``Performance-Based 
Containment Leak-Test Program,'' dated January 1995, Types B and C 
tests have identified a very large percentage of containment leakage 
paths, and the percentage of containment leakage paths that are 
detected only by Type A testing is very small. The Columbia Type A 
test history supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and (2) time based. Activity based failure mechanisms are defined as 
degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with ASME [American Society of Mechanical Engineers] 
Section XI, and TS requirements serve to provide a high degree of 
assurance that the containment would not degrade in a manner that is 
detectable only by a Type A test. Based on the above, the proposed 
test interval extensions do not significantly increase the 
consequences of an accident previously evaluated.
    The proposed amendment also deletes two exceptions previously 
granted. The first exception allowed a one-time extension of the 
ILRT test frequency for Columbia. This exception was for an activity 
that has already taken place; therefore, this deletion is solely an 
administrative action that does not result in any change in how 
Columbia is operated. The second exemption to compensate for flow 
metering inaccuracies in excess of those specified in the American 
National Standards Institute (ANSI)/American Nuclear Society (ANS) 
ANSI/ANS 56.8-1994 will be deleted as new test equipment has been 
acquired with accuracies within the tolerances specified in ANSI/ANS 
56.8-1994 and 2002.
    Therefore, the proposed changes do not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment to the TS 5.5.12, ``Primary Containment 
Leakage Rate Testing Program,'' involves the extension of the 
Columbia Type A containment test interval to 15 years and the 
extension of the Type C test interval to 75 months. The containment 
and the testing requirements to periodically demonstrate the 
integrity of the containment exist to ensure the plant's ability to 
mitigate the consequences of an accident.
    The proposed change does not involve a physical modification to 
the plant (i.e., no new or different type of equipment will be 
installed) nor does it alter the design, configuration, or change 
the manner in which the plant is operated or controlled beyond the 
standard functional capabilities of the equipment.
    The proposed amendment also deletes two exceptions previously 
granted. The first exception granted under TS Amendment No. 191 
allowed a one-time extension of the ILRT test frequency for 
Columbia. This exception was for an activity that has already 
occurred; therefore, this deletion is solely an administrative 
action that does not result in any change in how Columbia is 
operated. The second exemption which was originally granted via 
Amendment No. 144 to compensate for flow meter inaccuracies in 
excess of those specified in ANSI/ANS 56.8-1994, will be deleted as 
new test equipment has been acquired with accuracies within the 
tolerances specified in ANSI/ANS 56.8-1994 and 2002. These changes 
to the exceptions in TS 5.5.12 are administrative in nature and do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed amendment to TS 5.5.12 involves the extension of 
the Columbia Type A containment test interval to 15 years and the 
extension of the Type C test interval to 75 months for selected 
components. This amendment does not alter the manner in which safety 
limits, limiting safety system set points, or limiting conditions 
for operation are determined. The specific requirements and 
conditions of the TS Containment Leak Rate Testing Program exist to 
ensure that the degree of containment structural integrity and leak-
tightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by TS 
is maintained.
    The proposed change involves the extension of the interval 
between Type A containment leak rate tests and Type C tests for 
Columbia. The proposed surveillance interval extension is bounded by 
the 15-year ILRT interval and the 75-month Type C test interval 
currently authorized within NEI 94-01, Revision 3-A. Industry 
experience supports the conclusion that Type B and C testing detects 
a large percentage of containment leakage paths and that the 
percentage of containment leakage paths that are detected only by 
Type A testing is small.

[[Page 26132]]

The containment inspections performed in accordance with ASME 
Section Xl, and TS serve to provide a high degree of assurance that 
the containment would not degrade in a manner that is detectable 
only by Type A testing. The combination of these factors ensures 
that the margin of safety in the plant safety analysis is 
maintained. The design, operation, testing methods and acceptance 
criteria for Type A, B, and C containment leakage tests specified in 
applicable codes and standards would continue to be met, with the 
acceptance of this proposed change, since these are not affected by 
changes to the Type A and Type C test intervals. The proposed 
amendment also deletes exceptions previously granted to allow one 
time extension of the ILRT test frequency for Columbia. This 
exception was for an activity that has taken place; therefore, the 
deletion is solely an administrative action and does not change how 
Columbia is operated and maintained. Thus, there is no reduction in 
any margin of safety.
    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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Robert J. Pascarelli.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of amendment request: March 27, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17086A587.
    Description of amendment request: The proposed amendment would 
revise or add surveillance requirements (SRs) to verify that the system 
locations susceptible to gas accumulation are sufficiently filled with 
water and to provide allowances, which permit performance of the 
verification. The changes are being made to address the concerns 
discussed in Generic Letter 2008-01, ``Managing Gas Accumulation in 
Emergency Core Cooling, Decay Heat Removal, and Containment Spray 
Systems.'' The proposed amendment is consistent with Technical 
Specification Task Force (TSTF) TSTF-523, Revision 2, ``Generic Letter 
2008-01, Managing Gas Accumulation.''
    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 or adds SRs that require 
verification that the Emergency Core Cooling System (ECCS), Reactor 
Core Isolation Cooling (RCIC) System, Residual Heat Removal (RHR) 
Shutdown Cooling System, RHR Drywell Spray System, and RHR 
Suppression Pool Cooling System are not rendered inoperable due to 
accumulated gas and to provide allowances which permit performance 
of the revised verification. Gas accumulation in the subject systems 
is not an initiator of any accident previously evaluated. As a 
result, the probability of any accident previously evaluated is not 
significantly increased. The proposed SRs ensure that the subject 
systems continue to be capable to perform their assumed safety 
function and are not rendered inoperable due to gas accumulation. 
Thus, the consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, RCIC System, RHR Shutdown Cooling 
System, RHR Drywell Spray System, and RHR Suppression Pool Cooling 
System are not rendered inoperable due to accumulated gas and to 
provide allowances which permit performance of the revised 
verification. The proposed change does not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed) or a change in the methods governing normal plant 
operation. In addition, the proposed change does not impose any new 
or different requirements that could initiate an accident. The 
proposed change does not alter assumptions made in the safety 
analysis and is consistent with the safety analysis assumptions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, RCIC System, RHR Shutdown Cooling 
System, RHR Drywell Spray System, and RHR Suppression Pool Cooling 
System are not rendered inoperable due to accumulated gas and to 
provide allowances which permit performance of the revised 
verification. The proposed change adds new requirements to manage 
gas accumulation in order to ensure the subject systems are capable 
of performing their assumed safety functions. The proposed SRs are 
more comprehensive than the current SRs and will ensure that the 
assumptions of the safety analysis are protected. The proposed 
change does not adversely affect any current plant safety margins or 
the reliability of the equipment assumed in the safety analysis. 
Therefore, there are no changes being made to any safety analysis 
assumptions, safety limits or limiting safety system settings that 
would adversely affect plant safety as a result of the proposed 
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: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Robert J. Pascarelli.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center (DAEC), Linn County, Iowa
    Date of amendment request: March 31, 2017. A publicly-available 
version is in ADAMS under Package Accession No. ML17102B194.
    Description of amendment request: The proposed amendment by NextEra 
Energy Duane Arnold, LLC (NextEra Duane Arnold) would modify the DAEC 
Emergency Plan (E Plan) that revises the Emergency Planning Zone (EPZ) 
boundary for an area beyond the 10 mile required EPZ, specifically, 
subarea 24 of the EPZ by designating U.S. Highway 30 as its southern 
boundary. Currently, there is a tract within the DAEC EPZ subarea 24 
that is to the south of US Highway 30. This tract in subarea 24 is 
unique--otherwise, the entire DAEC EPZ is to the north of US Highway 
30, which is a four lane, divided highway. Subarea 24 is within Linn 
County, Iowa. The EPZ boundary change requires that a new Evacuation 
Time Estimates (ETE) study be performed for the DAEC host counties of 
Linn and Benton, Iowa, and this revision is also included in the 
proposal. The proposed change to the southern boundary of the EPZ is 
considered a reduction in effectiveness as defined in 10 CFR 50, 
Paragraph 50.54(q)(1)(iv) due to the reduction in EPZ area beyond the 
10 mile boundary, and as such, it requires prior NRC approval in 
accordance with the requirements of 10 CFR 50.54(q)(4). The

[[Page 26133]]

proposed change to the subarea 24 boundary will enhance law 
enforcement's ability to evacuate subareas in the Cedar Rapids area as 
well as improve their ability to control the access back into evacuated 
metro areas. Further, the proposed change to subarea 24 will make the 
overall DAEC EPZ boundary more consistent.
    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.
    This amendment request would alter portions of the southern, 
outer EPZ boundary defined in the DAEC E Plan to align with the EPZ 
boundaries requested by the Linn County Emergency Management 
Commission. The proposed amendment does not involve any 
modifications or physical changes to plant systems, structures, or 
components. The proposed amendment does not change plant operations 
or maintenance of plant systems, structures, or components, nor does 
the proposed amendment alter any DAEC E Plan facility or equipment. 
Changing the EPZ boundaries cannot increase the probability of an 
accident since emergency plan functions would be implemented after a 
postulated accident occurs. The proposed amendment does not alter or 
prevent the ability of the DAEC emergency response organization to 
perform intended emergency plan functions to mitigate the 
consequences of, and to respond adequately to, radiological 
emergencies.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of a previously evaluated 
accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This amendment request alters the EPZ boundary described in the 
DAEC E Plan. The proposed amendment does not involve any design 
modifications or physical changes to the plant, does not change 
plant operation or maintenance of equipment, and does not alter DAEC 
E Plan facilities or equipment. The proposed amendment to the DAEC E 
Plan does not alter any DAEC emergency actions that would be 
implemented in response to postulated accident events.
    The proposed amendment does not create any credible new failure 
mechanisms, malfunctions, or accident initiators not previously 
considered.
    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.
    This amendment request would alter one subarea in the EPZ 
boundary defined in the DAEC E Plan. The proposed amendment does not 
involve any design or licensing bases functions of the plant, no 
physical changes to the plant are to be made, it does not impact 
plant operation or maintenance of equipment, and it does not alter 
DAEC E Plan facilities or equipment. This change does not alter any 
DAEC emergency actions that would be implemented in response to 
postulated accident events. The DAEC E Plan continues to meet 10 CFR 
50.47 and 10 CFR 50, Appendix E requirements for emergency response.
    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, P. O. Box 14000 Juno Beach, 
FL 33408-0420.
    NRC Branch Chief: David J. Wrona.
Northern States Power Company--Minnesota (NSPM), Docket Nos. 50-263, 
50-282 and 50-306, Monticello Nuclear Generating Plant (MNGP), Wright 
County, and Prairie Island Nuclear Generating Plant, Units 1 and 2 
(PINGP), Goodhue County, Minnesota
    Date of amendment request: March 31, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17090A201.
    Description of amendment request: The proposed amendment would 
revise the PINGP technical specification (TS) 5.3, ``Plant Staff 
Qualifications'' and MNGP TS 5.3, ``Unit Staff Qualifications,'' 
subsections 5.3.1 to add an exception for licensed operators from the 
education and experience eligibility requirements of American National 
Standards Institute (ANSI) N18.1-1971, ``Selection and Training of 
Nuclear Power Plant Personnel,'' by requiring that licensed operators 
comply only with the requirements of 10 CFR part 55, ``Operators' 
Licenses.'' Additionally, the proposed change would revise the PINGP 
and MNGP TS 5.0, ``Administrative Controls,'' sub-sections 5.1-5.3 by 
making changes to standardize and align formatting to the extent 
possible between the TSs.
    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 revises TS 5.3.1 to take exception to ANSI 
N18.1-1971 requirements for the education and experience 
qualifications requirements for licensed operators and requires 
compliance with 10 CFR 55 and standardizes language between the TS 
without modifying meaning. An allowance for utilization of a 
Commission-approved training program that is based upon a SAT [site 
access training] is contained within 10 CFR 55. The NRC has also 
stated that the NANT [National Academy for Nuclear Training] 
guidelines, as endorsed, for initial licensed operator training and 
qualification are an acceptable way to meet the requirements of 10 
CFR 55.
    The proposed changes are administrative and do not affect any 
system that is a contributor to initiating events for previously 
evaluated accidents. Nor do the changes affect any system that is 
used to mitigate any previously evaluated accidents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    The proposed change revises TS 5.3.1 to take exception to ANSI 
N18.1-1971 requirements for the education and experience 
qualifications requirements for licensed operators and requires 
compliance with 10 CFR 55 and standardizes language between the TS 
without modifying the meaning. An allowance for utilization of a 
Commission-approved training program that is based upon a SAT is 
contained within 10 CFR 55. The NRC has also stated that the NANT 
guidelines, as endorsed, for initial licensed operator training and 
qualification are an acceptable way to meet the requirements of 10 
CFR 55. The proposed change is administrative and does not alter the 
design, function, or operation of any plant component, nor do they 
involve installation of any new or different equipment.
    Therefore, the proposed change does not create the possibility 
of a new or difference [different] kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises TS 5.3.1 to take exception to ANSI 
N18.1-1971 requirements for the education and experience 
qualifications requirements for licensed operators and requires 
compliance with 10 CFR 55 and standardizes language between the TS 
without modifying the meaning. An allowance for utilization of a 
Commission-approved training program that is based upon a SAT is 
contained within 10 CFR 55. The NRC has also stated that the NANT 
guidelines, as endorsed, for initial licensed operator training and 
qualification are an acceptable way to meet the

[[Page 26134]]

requirements of 10 CFR 55. The proposed change is administrative and 
does not alter the design, function, or operation of any plant 
component, nor do they involve installation of any new or different 
equipment.
    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: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: David J. Wrona.
Northern States Power Company--Minnesota (NSPM), Docket No. 50-263, 
Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota
    Date of amendment request: March 31, 2017. A publicly-available 
version is in ADAMS under Accession Package No. ML17095A107.
    Description of amendment request: The proposed amendment would 
revise the current emergency action levels (EAL) scheme used at MNGP to 
the EAL scheme contained in NEI 99-01, Revision 6, ``Development of 
Emergency Action Levels.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the MNGP EAL scheme does not impact the 
physical function of plant structures, systems or components (SSC) or 
the manner in which the SSCs perform their design function. The 
proposed change neither adversely affects accident initiators or 
precursors, nor alters design assumptions. Therefore, the proposed 
change does not alter or prevent the ability of SSCs to perform their 
intended function to mitigate the consequences of an event. The 
Emergency Plan, including the associated EALs, is implemented when an 
event occurs and cannot increase the probability of an accident. 
Further, the proposed change does not reduce the effectiveness of the 
Emergency Plan to meet the emergency planning requirements established 
in 10 CFR 50.47 and 10 CFR 50, Appendix E.
    Therefore, the proposed EAL scheme change does not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change does not involve any physical alteration to the 
plant, that is, no new or different type of equipment will be 
installed. The proposed change also does not change the method of plant 
operation and does not alter assumptions made in the safety analysis. 
Therefore, the proposed change will not create new failure modes or 
mechanisms that could result in a new or different kind of accident. 
The Emergency Plan, including the associated EAL scheme, is implemented 
when an event occurs and is not an accident initiator.
    Therefore, the proposed EAL scheme change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Margin of safety is provided by the ability of accident mitigation 
SSCs to perform at their analyzed capability. The change proposed in 
this license amendment request does not modify any plant equipment and 
there is no impact to the capability of the equipment to perform its 
intended accident mitigation function. The proposed change does not 
impact operation of the plant or its response to transients or 
accidents. Additionally, the proposed changes will not change any 
criteria used to establish safety limits or any safety system settings. 
The applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E 
will continue to be met.
    Therefore, the proposed EAL scheme change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: David J. Wrona.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1 (FCS), Washington County, Nebraska
    Date of amendment request: March 24, 2017. A publicly-available 
version is available in ADAMS under Accession No. ML17094A810.
    Description of amendment request: The amendment would revise the 
renewed facility operating license Paragraph 3.C, ``Security and 
Safeguards Contingency Plans.'' The amendment would revise the FCS 
Cyber Security Plan (CSP) implementation schedule for Milestone 8 (MS8) 
full implementation date from December 31, 2017, to December 28, 2018.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The amendment request proposes a change to the FCS CSP MS8 
completion date as set forth in the CSP implementation schedule and 
associated regulatory commitments. The NRC staff has concluded that 
the proposed change: (1) Does not alter accident analysis 
assumptions, add any initiators, or affect the function of plant 
systems or the manner in which systems are operated, maintained, 
modified, tested, or inspected; (2) does not require any plant 
modifications which affect the performance capability of the 
structures, systems, and components relied upon to mitigate the 
consequences of postulated accidents; and (3) has no impact on the 
probability or consequences of an accident previously evaluated. In 
addition, the NRC staff has concluded that the proposed change to 
the CSP implementation schedule is administrative in nature.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The NRC staff has concluded the proposed change: (1) Does not 
alter accident analysis assumptions, add any initiators, or affect 
the function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected; and (2) does 
not require any plant modifications which affect the performance 
capability of the structures, systems, and components relied upon to 
mitigate the consequences of

[[Page 26135]]

postulated accidents and does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
In addition, the NRC staff has concluded that the proposed change to 
the FCS CSP MS8 implementation schedule is administrative in nature.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The delay of the full 
implementation date for the FCS CSP MS8 has no substantive impact 
because other measures have been taken which provide adequate 
protection for the plant during this period of time. Therefore, the 
NRC staff has concluded that there is no significant reduction in a 
margin of safety. In addition, the NRC staff has concluded that the 
proposed change to the FCS CSP MS8 implementation schedule is 
administrative in nature.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700 
K Street NW., Washington, DC 20006-3817.
    NRC Branch Chief: Douglas A. Broaddus.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1 (FCS), Washington County, Nebraska
    Date of amendment request: March 31, 2017. A publicly-available 
version is available in ADAMS under Accession No. ML17093A309.
    Description of amendment request: The amendment would revise the 
FCS license conditions, definitions, and Technical Specifications (TS) 
sections to align with those required for the Permanently Defueled 
Technical Specifications (PDTS) that will reflect decommissioning 
requirements.
    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.
    Because the 10 CFR part 50 license for FCS will no longer 
authorize operation of the reactor or emplacement or retention of 
fuel into the reactor vessel with the certifications required by 10 
CFR part 50.82(a)(1) submitted, as specified in 10 CFR part 
50.82(a)(2), the occurrence of postulated accidents associated with 
reactor operation is no longer credible. The only remaining credible 
accident is a [fuel handling accident (FHA)]. The proposed amendment 
does not adversely affect the inputs or assumptions of any of the 
design basis analyses that impact the FHA.
    The only remaining [Update Safety Analysis Report (USAR)] 
Chapter 14 postulated accident scenario that could potentially occur 
at a permanently defueled facility would be a[n] FHA. Remaining 
Chapter 14 events include an accidental release of waste liquid and 
heavy load drop. Since the waste gas decay tanks have been purged of 
their content, and the volume control tanks, liquid holdup tanks, 
reactor coolant drain tank, and associated systems, contain waste 
that does not exceed any of the 10 CFR 50.67 limits if an event were 
to occur. The analyzed accident that remains applicable to FCS in 
the permanently shutdown and defueled condition is a[n] FHA in the 
auxiliary building where the SFP is located. The FHA analyses for 
FCS shows that, following 100 days of decay time after reactor 
shutdown and provided the [spent fuel pool (SFP)] water level 
requirements of TS 2.8.3(2) are met, the dose consequences are 
acceptable without relying on [structures, systems, and components 
(SSCs)] remaining functional for accident mitigation during and 
following the event. The one exception to this is the continued 
function of the passive SFP structure.
    The probability of occurrence of previously evaluated accidents 
is not increased, since extended operation in a defueled condition 
and safe storage and handling of fuel will be the only operations 
performed, and therefore bounded by the existing analyses. 
Additionally, the occurrence of postulated accidents associated with 
reactor operation will no longer be credible in a permanently 
defueled reactor. This significantly reduces the scope of applicable 
accidents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes have no impact on facility SSCs affecting 
the safe storage of irradiated fuel, or on the methods of operation 
of such SSCs, or on the handling and storage of irradiated fuel 
itself. The removal of TS that are related only to the operation of 
the nuclear reactor or only to the prevention, diagnosis, or 
mitigation of reactor-related transients or accidents, cannot result 
in different or more adverse failure modes or accidents than 
previously evaluated because the reactor is permanently shutdown and 
defueled and FCS is no longer authorized to operate the reactor.
    The proposed modification or deletion of requirements in the FCS 
10 CFR part 50 License and TS do not affect systems credited in the 
accident analysis for the FHA at FCS.
    The proposed license and TS will continue to require proper 
control and monitoring of systems associated with significant 
parameters and activities. The TSs continue to preserve the 
requirements for safe storage and movement of irradiated fuel.
    The proposed amendment does not result in any new mechanisms 
that could initiate damage to the remaining credited barriers for 
defueled plants (fuel cladding, spent fuel racks, SFP integrity, and 
SFP water level). Since extended operation in a defueled condition 
and safe fuel handling will be the only operations performed, and 
therefore bounded by the existing analyses, such a condition does 
not create the possibility of a new or different kind of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Because the 10 CFR part 50 license for FCS no longer authorizes 
operation of the reactor or emplacement or retention of fuel into 
the reactor vessel with the certifications required by 10 CFR part 
50.82(a)(1) submitted, as specified in 10 CFR part 50.82(a)(2), the 
occurrence of postulated accidents associated with reactor operation 
is no longer credible. The only remaining credible postulated 
accident is a[n] FHA. The proposed amendment does not adversely 
affect the inputs or assumptions of any of the design basis analyses 
that impact the FHA.
    The proposed changes are limited to those portions of the 
license and TS that are not related to the safe storage or movement 
of irradiated fuel. The requirements that are proposed to be revised 
or deleted from the FCS license and TS are not credited in the 
existing accident analysis for the remaining applicable postulated 
accident; and as such, do not contribute to the margin of safety 
associated with the accident analysis. Postulated [design-basis 
accidents (DBAs)] involving the reactor will no longer be possible 
because the reactor will be permanently shutdown and defueled and 
FCS will no longer be authorized to operate the reactor.
    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: David A. Repka, Esq., Winston & Strawn, 1700 
K Street NW., Washington, DC 20006-3817.

[[Page 26136]]

    NRC Branch Chief: Douglas A. Broaddus.
PSEG Nuclear, LLC, and Exelon Generation Company, LLC, Docket Nos. 50-
272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, 
Salem County, New Jersey
    Date of amendment request: March 6, 2017, as supplemented by letter 
dated May 4, 2017. Publicly-available versions are in ADAMS under 
Accession Nos. ML17065A241 and ML17125A051, respectively.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3.6.2.3, ``Containment Cooling System,'' 
to extend the containment fan coil unit allowed outage time (AOT) from 
7 days to 14 days for one or two inoperable containment fan coil units.
    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 containment fan cooling units (CFCUs) are safety related 
components which provide the minimum containment cooling as assumed 
by the containment response analysis for a design-basis loss of 
coolant accident (LOCA) or main steam line break (MSLB) event. The 
CFCUs are not accident initiators; the CFCUs are designed to 
mitigate the consequences of previously evaluated accidents 
including a design basis LOCA or MSLB event. Extending the AOT for 
one or two inoperable CFCUs would not affect the previously 
evaluated accidents since the remaining three CFCUs supplying 
cooling to containment would continue to be available to perform the 
accident mitigation functions. Thus allowing one or two CFCUs to be 
inoperable for an additional 7 days for performance of maintenance 
or testing does not increase the probability of a previously 
evaluated accident.
    Deterministic and probabilistic risk assessments evaluated the 
effect of the proposed Technical Specification change on the 
acceptability of operating with one or two CFCUs inoperable for up 
to 14 days. These assessments concluded that the proposed Technical 
Specification change does not involve a significant increase in the 
risk from CFCU unavailability.
    The calculated impact on risk associated with continued 
operation for an additional 7 days with one or two CFCUs inoperable 
is very small and is consistent with the acceptance guidelines 
contained in Regulatory Guides 1.174 and 1.177. This risk is judged 
to be reasonably consistent with the risk associated with operations 
for 7 days with one or two CFCUs inoperable as allowed by the 
current Technical Specifications. The remaining 3 operable CFCUs, in 
conjunction with the Containment Spray System, are adequate to 
supply cooling to remove sufficient heat from the reactor 
containment, following the initial LOCA/MSLB containment pressure 
transient, to keep the containment pressure from exceeding the 
design pressure.
    The consequences of previously evaluated accidents will remain 
the same during the proposed 14 day AOT as during the current 7 day 
AOT. The ability of the remaining 3 TS required CFCUs to maintain 
containment pressure and temperature within limits following a 
postulated design basis LOCA or MSLB event will not be affected.
    There will be no impact on the source term or pathways assumed 
in accidents previously evaluated. No analysis assumptions will be 
changed and there will be no adverse effects on onsite or offsite 
doses as the result of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed Technical Specification change does not involve a 
change in the plant design, system operation, or procedures involved 
with the CFCUs. The proposed changes allow one or two CFCUs to be 
inoperable for additional time. There are no new failure modes or 
mechanisms created due to plant operation for an extended period to 
perform CFCU maintenance or testing. Extended operation with one or 
two inoperable CFCUs does not involve any modification in the 
operational limits or physical design of plant systems. There are no 
new accident precursors generated due to the extended AOT.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. 
The proposed change, which would increase the AOT from 7 days to 14 
days for one or two inoperable CFCUs, does not exceed or alter a 
setpoint, design basis or safety limit.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC-N21, 
P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Branch Chief: James G. Danna.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028, 
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield, South 
Carolina
    Date of amendment request: May 2, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17122A353.
    Description of amendment request: The amendment request proposes 
changes to the Protection and Safety Monitoring System (PMS) including 
the reactor trip system (RTS) and the engineered safety feature 
actuation system (ESFAS), the passive core cooling system (PXS), the 
steam generator blowdown system (BDS), and the spent fuel pool cooling 
system (SFS). In addition, revisions are proposed to COL Appendix A, 
Technical Specifications. Because, this proposed change requires a 
departure from Tier 1 information in the Westinghouse Electric 
Company's AP1000 Design Control Document (DCD), the licensee also 
requested an exemption from the requirements of the Generic DCD Tier 1 
in accordance with 10 CFR 52.63(b)(1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to add IRWST lower narrow range level 
instruments addresses the accuracy required to initiate IRWST 
containment recirculation following a design basis accident in order 
to mitigate the consequences of the accident. The proposed change to 
add the new defense-in-depth refueling cavity and SFS isolation on 
Low IRWST wide range level addresses a seismic or other event 
resulting in a pipe rupture in the nonsafety-related, nonseismic SFS 
when connected to the IRWST that could potentially result in a loss 
of IRWST inventory. Isolation of the SFS from the IRWST to mitigate 
the consequences of a design basis accident continues to be 
implemented by the existing containment isolation function, and does 
not rely on the new defense-in-depth refueling cavity and SFS 
isolation on Low IRWST wide range level. The addition of RTS and 
ESFAS P-9 interlocks and blocks does not affect the availability of 
the actuated equipment to perform their design functions to mitigate 
the consequences of an accident. The proposed

[[Page 26137]]

changes do not involve any accident initiating component/system 
failure or event, thus the probabilities of the accidents previously 
evaluated are not affected.
    The affected equipment does not adversely affect or interact 
with safety-related equipment or a radioactive material barrier, and 
this activity does not involve the containment of radioactive 
material. Thus, the proposed changes would not adversely affect any 
safety-related accident mitigating function. The radioactive 
material source terms and release paths used in the safety analyses 
are unchanged, thus the radiological release in the UFSAR accident 
analyses are not affected.
    These proposed changes to the PMS design do not have an adverse 
effect on any of the design functions of the affected actuated 
systems. The proposed changes do not affect the support, design, or 
operation of mechanical and fluid Systems required to mitigate the 
consequences of an accident. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to the predicted radioactive releases due to postulated 
accident conditions. The plant response to previously evaluated 
accidents or external events is not adversely affected, nor do the 
proposed changes create any new accident precursors.
    Therefore, the requested amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to add IRWST lower narrow range level 
instruments include requirements similar in function and 
qualification to many safety-related instruments already performing 
the affected safety functions as described in the current licensing 
basis to enable the RTS and ESFAS to perform required design 
functions, and are consistent with other Updated Final Safety 
Analysis Report (UFSAR) information. The proposed change to add the 
new defense-in-depth refueling cavity and SFS isolation on Low IRWST 
wide range level addresses a seismic or other event resulting in a 
postulated pipe rupture in the nonsafety-related, nonseismic SFS 
when connected to the IRWST that could potentially result in a loss 
of IRWST inventory. Isolation of the SFS from the IRWST to mitigate 
the consequences of a design basis accident continues to be 
implemented by the existing containment isolation function, and does 
not rely on the new defense-in-depth refueling cavity and SFS 
isolation on Low IRWST wide range level. The addition of RTS and 
ESFAS P-9 interlocks and blocks does not affect the availability of 
the actuated equipment to perform their design functions to mitigate 
the consequences of an accident. This activity does not allow for a 
new radioactive material release path, result in a new radioactive 
material barrier failure mode, or create a new sequence of events 
that would result in significant fuel cladding failures.
    The proposed changes revise the PMS design. The proposed changes 
do not adversely affect the design requirements for the PMS, or the 
design requirements of associated actuated systems. The proposed 
changes do not adversely affect the design function, support, 
design, or operation of mechanical and fluid systems. The proposed 
changes to the PMS do not result in a new failure mechanism or 
introduce any new accident precursors. No design function described 
in the UFSAR is adversely affected by the proposed changes.
    Therefore, the requested amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    No safety analysis or design basis acceptance limit or 
acceptance criterion is challenged or exceeded by the proposed 
changes, and no margin of safety is reduced. The proposed change to 
add the new defense-in-depth refueling cavity and SFS isolation of 
Low IRWST wide range level addresses a seismic or other event 
resulting in a postulated pipe rupture in the nonsafety-related, 
nonseismic SFS when connected to the IRWST, maintaining the required 
IRWST inventory and preserving the original margin of safety assumed 
for the PXS and SFS.
    Therefore, the requested amendment does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius, LLC, 1111 Pennsylvania NW., Washington, DC 20004-2514.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: May 10, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17130A999.
    Description of amendment request: The VEGP amendment request 
proposes changes which involve departures from incorporated plant-
specific Tier 2 and Tier 2* Updated Final Safety Analysis Report 
(UFSAR) information in order to make changes to the design of certain 
components of the auxiliary building roof reinforcement and roof 
girders, and other related changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design functions of the auxiliary building roof are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the auxiliary 
building. The auxiliary building is a seismic Category I structure 
and is designed for dead, live, thermal, pressure, safe shutdown 
earthquake loads, and loads due to postulated pipe breaks. The 
auxiliary building roof is designed for snow, wind, and tornado 
loads and postulated external missiles. The proposed changes to 
UFSAR descriptions and figures are intended to address changes in 
the detail design of the auxiliary building roof. The thickness and 
strength of the auxiliary building roof are not reduced. As a 
result, the design function of the auxiliary building structure is 
not adversely affected by the proposed changes. There is no change 
to plant systems or the response of systems to postulated accident 
conditions. There is no change to the predicted radioactive releases 
due to postulated accident conditions. The plant response to 
previously evaluated accidents or external events is not adversely 
affected, nor do the changes described create any new accident 
precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to UFSAR descriptions and figures are 
proposed to address changes in the detail design of the auxiliary 
building roof. The thickness, geometry, and strength of the 
structures are not adversely altered. The concrete and reinforcement 
materials are not altered. The properties of the concrete are not 
altered. The changes to the design details of the auxiliary building 
structure do not create any new accident precursors. As a result, 
the design function of the auxiliary building structure is not 
adversely affected by the proposed changes.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The criteria and requirements of American Concrete Institute 
(ACI) 349 and American Institute of Steel Construction (AISC) N690 
provide a margin of safety to structural failure. The design of the 
auxiliary building structure conforms to applicable criteria and 
requirements in ACI 349 and AISC N690 and therefore maintains the 
margin of safety. The proposed changes to the UFSAR address changes 
in the detail design of the auxiliary

[[Page 26138]]

building roof. There is no change to design requirements of the 
auxiliary building structure. There is no change to the method of 
evaluation from that used in the design basis calculations. There is 
not a significant change to the in structure response spectra. No 
safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes, thus no margin of 
safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
    Date of amendment request: March 31, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17090A209.
    Description of amendment request: The requested amendment proposes 
changes to combined operating license (COL) Appendix C (and plant-
specific Tier 1) and Updated Final Safety Analysis Report (UFSAR) Tier 
2 that describe; (1) the inspection and analysis of, and specifies the 
maximum calculated flow resistance acceptance criteria for, the fourth-
stage (automatic depressurization system (ADS) loops; (2) revises 
licensing basis text in COL Appendix C (and plant-specific Tier 1) and 
UFSAR Tier 2 that describes the testing of, and specifies the allowable 
flow resistance acceptance criteria for, the in-containment refueling 
water storage tank (IRWST) injection line; (3) revises licensing basis 
text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 
that describes the testing of, and specifies the maximum flow 
resistance acceptance criteria for, the containment recirculation line; 
(4) revises licensing basis text in COL Appendix C (and plant-specific 
Tier 1) and UFSAR Tier 2 that specifies acceptance criteria for the 
maximum flow resistance between the IRWST drain line and the 
containment; and (5) removes licensing basis text from UFSAR Tier 2 
that discusses the operation of swing check valves in current operating 
plants. Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption 
from elements of the design as certified in the 10 CFR part 52, 
appendix D, design certification rule is also requested for the plant-
specific Design Control Document Tier 1 material departures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes do not adversely affect the operation of 
any systems or equipment that initiate an analyzed accident or alter 
any structures, systems, and components (SSCs) accident initiator or 
initiating sequence of events. The proposed changes do not adversely 
affect the physical design and operation of the in-containment 
refueling water storage tank (IRWST) injection, drain, containment 
recirculation, or fourth-stage automatic depressurization system 
(ADS) valves, including as-installed inspections and maintenance 
requirements as described in the Updated Final Safety Analysis 
Report (UFSAR). Inadvertent operation or failure of the fourth-stage 
ADS valves are considered as an accident initiator or part of an 
initiating sequence of events for an accident previously evaluated. 
However, the proposed change to the test methodology and calculated 
flow resistance for the fourth-stage ADS lines does not adversely 
affect the probability of inadvertent operation or failure. 
Therefore, the probabilities of the accidents previously evaluated 
in the UFSAR are not affected.
    The proposed changes do not adversely affect the ability of 
IRWST injection, drain, containment recirculation, and fourth-stage 
ADS valves to perform their design functions. The designs of the 
IRWST injection, drain, containment recirculation, and fourth-stage 
ADS valves continue to meet the same regulatory acceptance criteria, 
codes, and standards as required by the UFSAR. In addition, the 
proposed changes maintain the capabilities of the IRWST injection, 
drain, containment recirculation, and fourth-stage ADS valves to 
mitigate the consequences of an accident and to meet the applicable 
regulatory acceptance criteria. The proposed changes do not 
adversely affect the prevention and mitigation of other abnormal 
events, e.g., anticipated operational occurrences, earthquakes, 
floods and turbine missiles, or their safety or design analyses. 
Therefore, the consequences of the accidents evaluated in the UFSAR 
are not affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that might initiate a new or different kind of 
accident, or alter any SSC such that a new accident initiator or 
initiating sequence of events is created. The proposed changes do 
not adversely affect the physical design and operation of the IRWST 
injection, drain, containment recirculation, and fourth-stage ADS 
valves, including as-installed inspections, and maintenance 
requirements, as described in the UFSAR. Therefore, the operation of 
the IRWST injection, drain, containment recirculation, and fourth-
stage ADS valves is not adversely affected. These proposed changes 
do not adversely affect any other SSC design functions or methods of 
operation in a manner that results in a new failure mode, 
malfunction, or sequence of events that affect safety-related or 
nonsafety-related equipment. Therefore, this activity does not allow 
for a new fission product release path, result in a new fission 
product barrier failure mode, or create a new sequence of events 
that result in significant fuel cladding failures.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes maintain existing safety margins. The 
proposed changes verify and maintain the capabilities of the IRWST 
injection, drain, containment recirculation, and fourth-stage ADS 
valves to perform their design functions. The proposed changes 
maintain existing safety margin through continued application of the 
existing requirements of the UFSAR, while updating the acceptance 
criteria for verifying the design features necessary to ensure the 
IRWST injection, drain, containment recirculation, and fourth-stage 
ADS valves perform the design functions required to meet the 
existing safety margins in the safety analyses. Therefore, the 
proposed changes satisfy the same design functions in accordance 
with the same codes and standards as stated in the UFSAR. These 
changes do not adversely affect any design code, function, design 
analysis, safety analysis input or result, or design/safety margin.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes, and no margin of 
safety is reduced.
    Therefore, the requested amendment does not involve a 
significant reduction in a margin of safety.

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

[[Page 26139]]

Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Jennifer Dixon-Herrity.
Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna 
Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
    Date of amendment request: January 25, 2017, as supplemented by 
letter dated March 21, 2017. Publicly-available versions are in ADAMS 
under Accession Nos. ML17044A149 and ML17080A405.
    Description of amendment request: The amendments would revise 
certain Surveillance Requirements (SRs) in Technical Specification (TS) 
3.8.1, ``AC [Alternating Current] Sources--Operating.'' The request is 
for changes in the use of steady state voltage and frequency acceptance 
criteria for onsite standby power source of the diesel generators 
(DGs), allowing for the use of new and more conservative design 
analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of any accident previously 
evaluated?
    Response: No.
    The proposed amendment would provide more restrictive acceptance 
criteria for certain DG technical specification surveillance tests. 
The proposed acceptance criteria changes would help to ensure the 
DGs are capable of carrying the electrical loading assumed in the 
safety analyses that take credit for the operation of the DGs. [The 
proposed changes] would not affect the capability of other 
structures, systems, and components to perform their design 
function, and would not increase the likelihood of a malfunction.
    Therefore, the proposed amendment does not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes would provide more restrictive acceptance 
criteria to be applied to existing technical specification 
surveillance tests that demonstrate the capability of the facility 
DGs to perform their design function. The proposed acceptance 
criteria changes would not create any new failure mechanisms, 
malfunctions, or accident initiators not considered in the design 
and licensing bases.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed DG surveillance requirement changes to voltage and 
frequency test acceptance criteria are conservative because the 
minimum steady state voltage increase and the narrowing of the 
acceptable steady-state frequency range validates use of existing 
design basis analysis for these test acceptance criteria. Both 
changes support the use of conservative administrative controls that 
remain in place, allowing [the] use of the new test acceptance 
criteria in test procedures until technical specifications reflect 
these new requirements. The conduct of surveillance tests on safety 
related plant equipment is a means of assuring that the equipment is 
capable of maintaining the margin of safety established in the 
safety analyses for the facility. The proposed amendment does not 
affect DG performance as described in the design basis analyses, 
including the capability for the DG to attain and maintain required 
voltage and frequency for accepting and supporting plant safety 
loads, should a DG start signal occur. The proposed amendment does 
not introduce changes to limits established in accident analysis.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Damon D. Obie, Associate General Counsel, 
Talen Energy Supply, LLC, 835 Hamilton St., Suite 150, Allentown, PA 
18101.
    NRC Branch Chief: James G. Danna.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3 (BFN), Limestone County, 
Alabama
Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar 
Nuclear Plant, Units 1 and 2 (WBN), Rhea County, Tennessee
    Date of amendment request: April 5, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17096A620.
    Description of amendment request: The amendments would modify 
technical specification surveillance requirements (SRs) that currently 
operate ventilation systems with charcoal filters for 10 hours each 
month in accordance with Technical Specification Task Force (TSTF) 
Traveler TSTF-522, Revision 0, ``Revise Ventilation System Surveillance 
Requirements to Operate for 10 hours per Month.'' Specifically, BFN SRs 
3.6.4.3.1 and 3.7.3.1, and WBN SRs 3.6.9.1 and 3.7.12.1 are being 
revised to require operation of the systems for 15 continuous minutes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change replaces existing Surveillance Requirements 
to operate the SGT [Standby Gas Treatment] and CREV [Control Room 
Emergency Ventilation] systems for BFN and the EGT [Emergency Gas 
Treatment] and ABGT [Auxiliary Building Gas Treatment] systems for 
WBN, equipped with electric heaters for a continuous 10 hour period 
every 31 days with a requirement to operate the systems for 15 
continuous minutes with heaters operating.
    These systems are not accident initiators and therefore, these 
changes do not involve a significant increase in the probability of 
an accident. The proposed system and filter testing changes are 
consistent with current regulatory guidance for these systems and 
will continue to assure that these systems perform their design 
function which may include mitigating accidents. Thus the change 
does not involve a significant increase in the consequences of an 
accident.
    Therefore, it is concluded that this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change replaces existing Surveillance Requirements 
to operate the SGT and CREV systems for BFN and the EGT and ABGT 
systems for WBN, equipped with electric heaters for a continuous 10 
hour period every 31 days with a requirement to operate the systems 
for 15 continuous minutes with heaters operating.
    The change proposed for these ventilation systems does not 
change any system operations or maintenance activities. Testing 
requirements will be revised and will continue to demonstrate that 
the Limiting Conditions for Operation are met and the system 
components are capable of performing their intended safety 
functions. The change does not create new failure modes or 
mechanisms and no new accident precursors are generated.
    Therefore, it is concluded that this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

[[Page 26140]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change replaces existing Surveillance Requirements 
to operate the SGT and CREV systems for BFN and the EGT and ABGT 
systems for WBN, equipped with electric heaters for a continuous 10 
hour period every 31 days with a requirement to operate the systems 
for 15 continuous minutes with heaters operating.
    The design basis for the ventilation systems' heaters is to heat 
the incoming air which reduces the relative humidity. The heater 
testing change proposed will continue to demonstrate that the 
heaters are capable of heating the air and will perform their design 
function. The proposed change is consistent with regulatory 
guidance.
    Therefore, it is concluded that this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Dr., WT 6A, Knoxville, TN 37902.
    NRC Branch Chief: Benjamin G. Beasley.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee
    Date of amendment request: March 16, 2017. A publicly-available 
version is in ADAMS under Accession No. ML17075A229.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' Table 3.3.1-1, to increase the values for the 
nominal trip setpoint and the allowable value for Function 14.a. 
``Turbine Trip--Low Fluid Oil Pressure.'' The proposed amendment also 
requests changes in accordance with Technical Specifications Task Force 
(TSTF) Traveler TSTF-493, Revision 4, ``Clarify Application of Setpoint 
Methodology for LSSS [Limiting Safety System Settings] Functions,'' 
Option A, for the affected turbine trip on low fluid oil pressure 
function setpoints only.
    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 reflects a design change to the turbine 
control system that results in the use of an increased control oil 
[system pressure], necessitating a change to the value at which a 
low fluid oil pressure initiates a reactor trip on turbine trip. The 
low fluid oil pressure is an input to the reactor trip 
instrumentation in response to a turbine trip event. The value at 
which the low fluid oil initiates a reactor trip is not an accident 
initiator. A change in the nominal control oil pressure does not 
introduce any mechanisms that would increase the probability of an 
accident previously analyzed. The reactor trip on turbine trip 
function is initiated by the same protective signal as used for the 
existing auto stop low fluid oil system trip signal. There is no 
change in form or function of this signal and the probability or 
consequences of previously analyzed accidents are not impacted.
    The proposed change also adds test requirements to the low fluid 
oil pressure TS instrument function related to those variables to 
ensure that instruments will function as required to initiate 
protective systems or actuate mitigating systems at the point 
assumed in the applicable setpoint calculation. Surveillance tests 
are not an initiator to any accident previously evaluated. As a 
result, the probability of any accident previously evaluated is not 
significantly increased. The systems and components required by the 
low fluid oil pressure TS instrument function for which surveillance 
tests are added are still required to be operable, meet the 
acceptance criteria for the surveillance requirements, and be 
capable of performing any mitigation function.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    The EHC [electrohydraulic control] fluid oil pressure rapidly 
decreases in response to a turbine trip signal. The value at which 
the low fluid oil pressure switches initiates a reactor trip is not 
an accident initiator. The proposed TS change reflects the higher 
pressure that will be sensed after the pressure switches are 
relocated from the auto stop low fluid oil system to the EHC high 
pressure header. Failure of the new switches would not result in a 
different outcome than is considered in the current design basis. 
Further, the change does not alter assumptions made in the safety 
analysis but ensures that the instruments perform as assumed in the 
accident analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The change involves a parameter that initiates an anticipatory 
reactor trip following a turbine trip. The safety analyses do not 
credit this anticipatory trip for reactor core protection. The 
original pressure switch configuration and the new pressure switch 
configuration both generate the same reactor trip signal. The 
difference is that the initiation of the trip will now be adjusted 
to a different system of higher pressure. This system function of 
sensing and transmitting a reactor trip signal on turbine trip 
remains the same. Also, the proposed change adds test requirements 
that will assure that technical specifications instrumentation 
allowable values: (1) Will be limiting settings for assessing 
instrument channel operability and; (2) will be conservatively 
determined so that evaluation of instrument performance history and 
the as left tolerance requirements of the calibration procedures 
will not have an adverse effect on equipment operability. The 
testing methods and acceptance criteria for systems, structures, and 
components, specified in applicable codes and standards (or 
alternatives approved for use by the NRC) will continue to be met as 
described in the plant licensing basis including the updated Final 
Safety Analysis Report. There is no impact to safety analysis 
acceptance criteria as described in the plant licensing basis 
because no change is made to the accident analysis assumptions.
    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: Sherry A. Quirk, General Counsel, Tennessee 
Valley Authority, 400 West Summit Hill Drive, 6A West Tower, Knoxville, 
TN 37902.
    NRC Branch Chief: Benjamin G. Beasley.

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,

[[Page 26141]]

and opportunity for a hearing in connection with these actions, was 
published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3 (PVNGS), Maricopa County, Arizona
    Date of amendment request: June 29, 2016.
    Description of amendment request: The amendments revised the 
Technical Specifications (TSs) for PVNGS, by modifying the TS 
requirements to address Generic Letter 2008-01, ``Managing Gas 
Accumulation in Emergency Core Cooling, Decay Heat Removal, and 
Containment Spray Systems,'' as described in TS Task Force [TSTF]-523, 
Revision 2, ``Generic Letter 2008-01, Managing Gas Accumulation.''
    Date of issuance: May 16, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 1 year from the date of issuance.
    Amendment Nos.: Unit 1--202, Unit 2--202, and Unit 3--202. A 
publicly available version is in ADAMS under Accession No. ML17123A435; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: 
The amendments revised the Operating Licenses and TSs.
    Date of initial notice in Federal Register: August 16, 2016 (81 FR 
54613).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 16, 2017.
    No significant hazards consideration comments received: No.
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 21, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16209A223.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) for the Oconee Nuclear Station, Units 1, 
2, and 3 (ONS); specifically, TS 2.1.1.1, ``Reactor Core SLs [Safety 
Limits],'' and TS 5.6.5, ``Core Operating Limits Report (COLR),'' to 
allow the use of the COPERNIC fuel performance code.
    Date of issuance: May 11, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 403, 405, and 404. A publicly-available version is 
in ADAMS under Accession No. ML17103A509; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: February 14, 2017 (82 
FR 10593).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 11, 2017.
    No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of application for amendment: June 28, 2016, as supplemented 
by letter dated, August 11, 2016, August 18, 2016, November 14, 2016, 
December 8, 2016, December 12, 2016, January 9, 2017, January 12, 2017, 
February 16, 2017, February 21, 2017, March 7, 2017.
    Brief description of amendment: The amendment would revise the 
operating license and technical specifications to implement an increase 
in rated thermal power from the current licensed thermal power of 3486 
megawatts (MWt) to a measurement uncertainty recapture thermal power of 
3544 MWt.
    Date of issuance: May 11, 2017.
    Effective date: As of its date of issuance and shall be implemented 
within 120 days from the date of issuance, or during the 2017 Refueling 
Outage if issued on May 13, 2017, or earlier.
    Amendment No.: 241. A publicly-available version is in ADAMS under 
Accession No. ML17095A117; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-21: The amendment 
revised the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: October 4, 2016 (81 FR 
68470). The supplemental letter(s) dated August 11, 2016, August 18, 
2016, November 14, 2016, December 8, 2016, December 12, 2016, January 
9, 2017, January 12, 2017, February 16, 2017, February 21, 2017, and 
March 7, 2017, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 11, 2017.
    No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
2 (ANO-2), Pope County, Arkansas
    Date of application for amendment: October 27, 2016, as 
supplemented by letters dated December 2, 2016, and February 21, 2017.
    Brief description of amendment: The amendment authorized a new 
risk-informed, performance-based fire protection licensing basis for 
ANO-2, with revised modifications, recovery actions, ignition 
frequencies, and the application of an NRC-approved fire modeling 
method. The amendment also revised Attachments M, ``License Condition 
Changes''; Attachment S, ``Plant Modifications and Items to be 
Completed during Implementation''; and Attachment W, ``Fire PRA 
[Probabilistic Risk Assessment] Insights,'' of the previously approved 
National Fire Protection Association (NFPA) 805 amendment.
    Date of issuance: May 12, 2017.
    Effective date: As of the date of issuance and shall be implemented 
as described in the transition license conditions.
    Amendment No.: 306. A publicly-available version is in ADAMS under 
Accession No. ML17096A235; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
renewed facility operating license.

[[Page 26142]]

    Date of initial notice in Federal Register: January 31, 2017 (82 FR 
8869). The supplemental letter dated February 21, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 12, 2017.
    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
    Date of amendment request: July 26, 2016.
    Brief description of amendments: The amendments revised the 
Technical Specification (TS) requirements relating to the inservice 
inspection program required by the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Code and the inservice testing 
program required by the ASME Code for Operation and Maintenance of 
Nuclear Power Plants. The changes are based in part on Technical 
Specifications Task Force (TSTF) Traveler TSTF-545, Revision 3, ``TS 
Inservice Testing Program Removal & Clarify SR [Surveillance 
Requirement] Usage Rule Application to Section 5.5 Testing.''
    Date of issuance: May 16, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment Nos.: 225 (Unit 1) and 188 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17103A081; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-39 and NPF-85: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: October 25, 2016 (81 FR 
73435).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 16, 2017.
    No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, 
Pennsylvania
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
    Date of application for amendments: May 24, 2016, as supplemented 
by letter dated October 25, 2016.
    Brief description of amendments: The amendments eliminated the 
technical specifications (TS), Section 5.5, ``Inservice Testing 
Program,'' to remove requirements duplicated in American Society of 
Mechanical Engineers (ASME) Code for Operations and Maintenance of 
Nuclear Power Plants (OM Code), Case OMN-20, ``Inservice Test 
Frequency.'' A new defined term, ``INSERVICE TESTING PROGRAM,'' was 
added to TS Section 1.1, ``Definitions.'' This change to the TS is 
consistent with TSTF-545, Revision 3, ``TS Inservice Testing Program 
Removal & Clarify SR [Surveilance Requirement] Usage Rule Application 
to Section 5.5 Testing,'' with deviations as described in the license 
amendment request dated May 24, 2016 (ADAMS Accession No. ML16148A047).
    Date of issuance: May 11, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 150 days from the date of issuance.
    Amendment Nos.: 298 for DPR-66, 186 for NPF-73, 295 for NPF-3, and 
175 for NPF-58. A publicly-available version is in ADAMS under 
Accession No. ML17081A509; the documents related to these amendments 
are listed in the Safety Evaluation enclosed with the amendment(s).
    Facility Operating License Nos. DPR-66, NPF-73, NPF-3, and NPF-58: 
The amendments revised the Technical Specifications and the Licenses.
    Date of initial notice in Federal Register: August 2, 2016 (81 FR 
50732). The supplement dated October 25, 2016, contained clarifying 
information and did not change the NRC staff's initial proposed finding 
of no significant hazards consideration.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 11, 2017.
    No significant hazards consideration comments received: No.
PSEG Nuclear, LLC, Docket Nos. 50-354, 50-272, and 50-311, Hope Creek 
Generating Station, and Salem Nuclear Generating Station, Unit Nos. 1 
and 2, Salem County, New Jersey
    Date of amendment request: June 30, 2016.
    Brief description of amendments: The amendments revised the Cyber 
Security Plan (CSP) Milestone 8 implementation schedule for Hope Creek 
Generating Station (Hope Creek) and Salem Nuclear Generating Station 
(Salem), Unit Nos. 1 and 2. Specifically, this change extended the PSEG 
Nuclear LLC (PSEG) CSP Milestone 8 full implementation date as set 
forth in the PSEG CSP implementation schedule and revised the Renewed 
Facility Operating Licenses.
    Date of issuance: May 16, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 204 (Hope Creek), 318 (Salem, Unit No. 1), and 299 
(Salem, Unit No. 2). A publicly-available version is in ADAMS under 
Accession No. ML17093A870; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-57, DPR-70, and DPR-75: 
The amendments revised the Renewed Facility Operating Licenses.
    Date of initial notice in Federal Register: October 4, 2016 (81 FR 
68471).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 16, 2017.
    No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company and South Carolina Public Service 
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear 
Station (VCSNS), Units 2 and 3, Fairfield, South Carolina
    Date of amendment request: July 19, 2016.
    Brief description of amendments: The amendments change Combined 
License (COL) Nos. NPF-93 and NPF-94 for the VCSNS, Units 2 and 3. The 
amendments change the station's Updated Final Safety Analysis Reports 
(UFSAR) by departing from the incorporated AP1000 Design Control 
Document Tier 2 information and involve related changes to the combined 
operating license (COL) Appendix A Technical Specifications (TS). 
Specifically, the changes revise the COLs and plant-specific UFSAR Tier 
2 information and TS to update the Protection and Safety Monitoring

[[Page 26143]]

System (PMS) to align with the standards of the Institute of Electrical 
and Electronics Engineers (IEEE) 603-1991, ``IEEE Standard Criteria for 
Safety Systems for Nuclear Power Generating Stations.''
    Date of issuance: April 10, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 69. Publicly-available versions are in ADAMS under 
Accession Nos. ML17041A020 and ML17041A022; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Facility Operating License Nos. NPF-93 and NPF-94: Amendments 
revised the COL UFSAR in the form of departures from the incorporated 
plant-specific DCD Tier 2 information and COL Appendix A TS.
    Date of initial notice in Federal Register: August 30, 2016 (81 FR 
59659).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 10, 2017.
    No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: December 16, 2016, and supplemented by 
letters dated January 12 and February 22, 2017.
    Description of amendment: The amendment consists of changes to the 
VEGP Units 3 and 4 Updated Final Safety Analysis Report (UFSAR) in the 
form of departures from the incorporated plant specific Design Control 
Document Tier 2 information. Specifically, the amendment consists of 
changes to the UFSAR to provide clarification of the interface criteria 
for nonsafety-related instrumentation that monitors safety-related 
fluid systems.
    Date of issuance: May 1, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 76 and 75. A publicly-available version is in ADAMS 
under Accession Package No. ML17094A845; documents related to this 
amendment are listed in the Safety Evaluation enclosed with the 
amendment.
    Facility Combined License Nos. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: February 28, 2017 (82 
FR 12130). The supplemental letters dated January 12, and February 22, 
2017, provided additional information that clarified the application, 
did not expand the scope of the application request as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in the Safety Evaluation dated May 1, 2017.
    No significant hazards consideration comments received: No.
Southern Nuclear Operating Company (SNC), Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke 
County, Georgia
    Date of amendment request: March 4, 2016, as supplemented on 
January 31, 2017.
    Description of amendment: This amendment revises License Condition 
(LC) 2.D(12)(d) related to initial Emergency Action Levels (EALs). The 
LC will require SNC to submit a fully-developed set of EALs before 
initial fuel load in accordance with the criteria defined in this 
license amendment.
    Date of issuance: May 18, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 77 (Unit 3) and 76 (Unit 4). A publicly-available 
version is in ADAMS under Accession Package No. ML17045A537; documents 
related to this amendment are listed in the Safety Evaluation enclosed 
with the amendment.
    Facility Combined License Nos. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: August 2, 2016 (81 FR 
50736). The supplemental letter dated January 31, 2017, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in the Safety Evaluation dated May 18, 2017
    No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant, 
Unit 2, Rhea County, Tennessee
    Date of amendment request: February 16, 2017.
    Brief description of amendment: The amendment revised the Technical 
Specification Containment Leakage Rate Testing Program to allow a one-
time extension for the Type C local leak rate test for certain 
containment isolation valves.
    Date of issuance: May 18, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 11. A publicly-available version is in ADAMS under 
Accession No. ML17123A228; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-96: Amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: March 14, 2017 (82 FR 
13671).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 18, 2017.
    No significant hazards consideration comments received: No.
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: May 10, 2016, as supplemented by letter 
dated October 18, 2016.
    Brief description of amendments: The amendments would expand 
primary grade water lockout requirements in Technical Specification 
(TS) 3.2.E from being applicable in refueling shutdown (RSD) and cold 
shutdown (CSD) modes to being applicable in RSD, CSD, intermediate 
shutdown, and hot shutdown modes.
    Date of issuance: May 10, 2017.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 288 (Unit 1) and 288 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML17039A513; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: October 11, 2016 (81 FR 
70187). The supplemental letter dated October 18, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.

[[Page 26144]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 10, 2017.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 24th day of May, 2017.

    For the Nuclear Regulatory Commission.

Kathryn M. Brock,
 Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.
[FR Doc. 2017-11679 Filed 6-5-17; 8:45 am]
 BILLING CODE 7590-01-P