Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 26128-26144 [2017-11679]
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Federal Register / Vol. 82, No. 107 / Tuesday, June 6, 2017 / Notices
ATTACHMENT 1—GENERAL TARGET SCHEDULE FOR PROCESSING AND RESOLVING REQUESTS FOR ACCESS TO SENSITIVE
UNCLASSIFIED NON-SAFEGUARDS INFORMATION IN THIS PROCEEDING—Continued
Day
Event/activity
A .......................
If access granted: Issuance of presiding officer or other designated officer decision on motion for protective order for access
to sensitive information (including schedule for providing access and submission of contentions) or decision reversing a
final adverse determination by the NRC staff.
Deadline for filing executed Non-Disclosure Affidavits. Access provided to SUNSI consistent with decision issuing the protective order.
Deadline for submission of contentions whose development depends upon access to SUNSI. However, if more than 25 days
remain between the petitioner’s receipt of (or access to) the information and the deadline for filing all other contentions (as
established in the notice of opportunity to request a hearing and petition for leave to intervene), the petitioner may file its
SUNSI contentions by that later deadline.
(Contention receipt +25) Answers to contentions whose development depends upon access to SUNSI.
(Answer receipt +7) Petitioner/Intervenor reply to answers.
Decision on contention admission.
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[FR Doc. 2017–11211 Filed 6–5–17; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[NRC–2017–0131]
Biweekly Notice; Applications and
Amendments to Facility Operating
Licenses and Combined Licenses
Involving No Significant Hazards
Considerations
Nuclear Regulatory
Commission.
ACTION: Biweekly notice.
AGENCY:
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
https://www.regulations.gov and search
for Docket ID NRC–2017–0131. Address
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SUMMARY:
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questions about NRC dockets to Carol
Gallagher; telephone: 301–415–3463;
email: Carol.Gallagher@nrc.gov. For
technical questions, contact the
individual listed in the FOR FURTHER
INFORMATION CONTACT section of this
document.
• Mail comments to: Cindy Bladey,
Office of Administration, Mail Stop:
TWFN–8–D36M, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001.
For additional direction on obtaining
information and submitting comments,
see ‘‘Obtaining Information and
Submitting Comments’’ in the
SUPPLEMENTARY INFORMATION section of
this document.
FOR FURTHER INFORMATION CONTACT:
Shirley Rohrer, Office of Nuclear
Reactor Regulation, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001; telephone: 301–415–
5411; email: shirley.rohrer@nrc.gov.
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 publiclyavailable information related to this
action by any of the following methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2017–0131.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may obtain publiclyavailable documents online in the
ADAMS Public Documents collection at
https://www.nrc.gov/reading-rm/
adams.html. To begin the search, select
‘‘ADAMS Public Documents’’ and then
select ‘‘Begin Web-based ADAMS
Search.’’ For problems with ADAMS,
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please contact the NRC’s Public
Document Room (PDR) reference staff at
1–800–397–4209, 301–415–4737, or by
email to pdr.resource@nrc.gov. The
ADAMS accession number for each
document referenced (if it is available in
ADAMS) is provided the first time that
it is mentioned in this document.
• NRC’s PDR: You may examine and
purchase copies of public documents at
the NRC’s PDR, Room O1–F21, One
White Flint North, 11555 Rockville
Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC–2017–
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 https://
www.regulations.gov as well as enter the
comment submissions into ADAMS.
The NRC does not routinely edit
comment submissions to remove
identifying or contact information.
If you are requesting or aggregating
comments from other persons for
submission to the NRC, then you should
inform those persons not to include
identifying or contact information that
they do not want to be publicly
disclosed in their comment submission.
Your request should state that the NRC
does not routinely edit comment
submissions to remove such information
before making the comment
submissions available to the public or
entering the comment into ADAMS.
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Federal Register / Vol. 82, No. 107 / Tuesday, June 6, 2017 / Notices
II. Notice of Consideration of Issuance
of Amendments to Facility Operating
Licenses and Combined Licenses and
Proposed No Significant Hazards
Consideration Determination
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The Commission has made a
proposed determination that the
following amendment requests involve
no significant hazards consideration.
Under the Commission’s regulations in
§ 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 60day 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
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action. Petitions shall be filed in
accordance with the Commission’s
‘‘Agency Rules of Practice and
Procedure’’ in 10 CFR part 2. Interested
persons should consult a current copy
of 10 CFR 2.309. The NRC’s regulations
are accessible electronically from the
NRC Library on the NRC’s Web site at
https://www.nrc.gov/reading-rm/doccollections/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
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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,
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local governmental body, Federallyrecognized 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 EFiling process requires participants to
submit and serve all adjudicatory
documents over the internet, or in some
cases to mail copies on electronic
storage media. Detailed guidance on
making electronic submissions may be
found in the Guidance for Electronic
Submissions to the NRC and on the NRC
Web site at https://www.nrc.gov/sitehelp/e-submittals.html. Participants
may not submit paper copies of their
filings unless they seek an exemption in
accordance with the procedures
described below.
To comply with the procedural
requirements of E-Filing, at least 10
days prior to the filing deadline, the
participant should contact the Office of
the Secretary by email at
hearing.docket@nrc.gov, or by telephone
at 301–415–1677, to (1) request a digital
identification (ID) certificate, which
allows the participant (or its counsel or
representative) to digitally sign
submissions and access the E-Filing
system for any proceeding in which it
is participating; and (2) advise the
Secretary that the participant will be
submitting a petition or other
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adjudicatory document (even in
instances in which the participant, or its
counsel or representative, already holds
an NRC-issued digital ID certificate).
Based upon this information, the
Secretary will establish an electronic
docket for the hearing in this proceeding
if the Secretary has not already
established an electronic docket.
Information about applying for a
digital ID certificate is available on the
NRC’s public Web site at https://
www.nrc.gov/site-help/e-submittals/
getting-started.html. Once a participant
has obtained a digital ID certificate and
a docket has been created, the
participant can then submit
adjudicatory documents. Submissions
must be in Portable Document Format
(PDF). Additional guidance on PDF
submissions is available on the NRC’s
public Web site at https://www.nrc.gov/
site-help/electronic-sub-ref-mat.html. A
filing is considered complete at the time
the document is submitted through the
NRC’s E-Filing system. To be timely, an
electronic filing must be submitted to
the E-Filing system no later than 11:59
p.m. Eastern Time on the due date.
Upon receipt of a transmission, the EFiling system time-stamps the document
and sends the submitter an email notice
confirming receipt of the document. The
E-Filing system also distributes an email
notice that provides access to the
document to the NRC’s Office of the
General Counsel and any others who
have advised the Office of the Secretary
that they wish to participate in the
proceeding, so that the filer need not
serve the document on those
participants separately. Therefore,
applicants and other participants (or
their counsel or representative) must
apply for and receive a digital ID
certificate before adjudicatory
documents are filed so that they can
obtain access to the documents via the
E-Filing system.
A person filing electronically using
the NRC’s adjudicatory E-Filing system
may seek assistance by contacting the
NRC’s Electronic Filing Help Desk
through the ‘‘Contact Us’’ link located
on the NRC’s public Web site at https://
www.nrc.gov/site-help/esubmittals.html, by email to
MSHD.Resource@nrc.gov, or by a tollfree 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
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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 publiclyavailable 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,
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see the ‘‘Obtaining Information and
Submitting Comments’’ section of this
document.
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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 nonroutine 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
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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?
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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.
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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.
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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
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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
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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
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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?
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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:
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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
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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.
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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
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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.
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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
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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.
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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
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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
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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.
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NRC Branch Chief: Douglas A.
Broaddus.
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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
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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.
PO 00000
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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
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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-indepth 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 nonsafetyrelated, 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
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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 nonsafetyrelated, 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 DixonHerrity.
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
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26137
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
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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.
mstockstill on DSK30JT082PROD with NOTICES
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 DixonHerrity.
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
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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 fourthstage 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?
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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 nonsafetyrelated 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
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Sixth Avenue North, Birmingham, AL
35203–2015.
NRC Branch Chief: Jennifer DixonHerrity.
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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. Publiclyavailable 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.
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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 steadystate 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)
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26139
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.
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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.
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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
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20:52 Jun 05, 2017
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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
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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,
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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.
mstockstill on DSK30JT082PROD with NOTICES
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.
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22:57 Jun 05, 2017
Jkt 241001
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.
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Amendment No.: 241. A publiclyavailable 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 riskinformed, 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 publiclyavailable 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.
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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.
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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.
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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
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
FirstEnergy Nuclear Operating
Company, et al., Docket No. 50–346,
Davis-Besse Nuclear Power Station, Unit
No. 1, Ottawa County, Ohio
Date of amendment request: 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 publiclyavailable 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.
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.
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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
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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. Publiclyavailable 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 plantspecific 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
VerDate Sep<11>2014
20:52 Jun 05, 2017
Jkt 241001
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 fullydeveloped 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
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Sfmt 4703
26143
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 publiclyavailable 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.
E:\FR\FM\06JNN1.SGM
06JNN1
26144
Federal Register / Vol. 82, No. 107 / Tuesday, June 6, 2017 / Notices
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
NUCLEAR REGULATORY
COMMISSION
[NRC–2017–0001]
Sunshine Act Meeting Notice
Weeks of June 5, 12, 19, 26, July
3, 10, 2017.
PLACE: Commissioners’ Conference
Room, 11555 Rockville Pike, Rockville,
Maryland.
STATUS: Public and Closed.
DATE:
Week of June 5, 2017
There are no meetings scheduled for
the week of June 5, 2017.
Week of June 12, 2017—Tentative
Tuesday, June 13, 2017
10:00 a.m. Briefing on Human
Capital and Equal Employment
Opportunity (Public Meeting);
(Contact: Tanya Parwani-Jaimes:
301–287–0730)
This meeting will be webcast live at
the Web address—https://www.nrc.gov/.
Thursday, June 15, 2017
9:00 a.m. Briefing on Results of the
Agency Action Review Meeting
(Public Meeting); (Contact: Andrew
Waugh: 301–415–5601)
This meeting will be webcast live at
the Web address—https://www.nrc.gov/.
Week of June 19, 2017—Tentative
There are no meetings scheduled for
the week of June 19, 2017.
There are no meetings scheduled for
the week of June 26, 2017.
mstockstill on DSK30JT082PROD with NOTICES
Week of July 3, 2017—Tentative
There are no meetings scheduled for
the week of July 3, 2017.
Week of July 10, 2017—Tentative
There are no meetings scheduled for
the week of July 10, 2017.
*
*
*
*
*
The schedule for Commission
meetings is subject to change on short
20:52 Jun 05, 2017
Jkt 241001
Dated: June 1, 2017.
Denise L. McGovern,
Policy Coordinator, Office of the Secretary.
[FR Doc. 2017–11731 Filed 6–2–17; 11:15 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[Docket Nos. 72–1014, 72–59, and 50–271;
NRC–2017–0134]
Entergy Nuclear Operations, Inc.;
Vermont Yankee Nuclear Power
Station, Independent Spent Fuel
Storage Installation
Nuclear Regulatory
Commission.
ACTION: Environmental assessment and
finding of no significant impact;
issuance.
AGENCY:
Week of June 26, 2017—Tentative
VerDate Sep<11>2014
notice. For more information or to verify
the status of meetings, contact Denise
McGovern at 301–415–0681 or via email
at Denise.McGovern@nrc.gov.
*
*
*
*
*
The NRC Commission Meeting
Schedule can be found on the Internet
at: https://www.nrc.gov/public-involve/
public-meetings/schedule.html.
*
*
*
*
*
The NRC provides reasonable
accommodation to individuals with
disabilities where appropriate. If you
need a reasonable accommodation to
participate in these public meetings, or
need this meeting notice or the
transcript or other information from the
public meetings in another format (e.g.,
braille, large print), please notify
Kimberly Meyer, NRC Disability
Program Manager, at 301–287–0739, by
videophone at 240–428–3217, or by
email at Kimberly.Meyer-Chambers@
nrc.gov. Determinations on requests for
reasonable accommodation will be
made on a case-by-case basis.
*
*
*
*
*
Members of the public may request to
receive this information electronically.
If you would like to be added to the
distribution, please contact the Nuclear
Regulatory Commission, Office of the
Secretary, Washington, DC 20555 (301–
415–1969), or email
Brenda.Akstulewicz@nrc.gov or
Patricia.Jimenez@nrc.gov.
The U.S. Nuclear Regulatory
Commission (NRC) is considering an
exemption request from Entergy Nuclear
Operations, Inc. (Entergy) to allow the
Vermont Yankee Nuclear Power Station
(VYNPS) to load higher enriched fuel
assemblies with certain lower enriched
fuel assemblies in the same HI–STORM
100 multi-purpose canister (MPC) using
SUMMARY:
PO 00000
Frm 00103
Fmt 4703
Sfmt 4703
Certificate of Compliance (CoC) No.
1014, Amendment No. 10. The NRC
prepared an environmental assessment
(EA) documenting its finding. The NRC
concluded that the proposed action
would have no significant
environmental impact. Accordingly, the
NRC staff is issuing a finding of no
significant impact (FONSI) associated
with the proposed exemption.
DATES: The EA and FONSI referenced in
this document are available on June 6,
2017.
ADDRESSES: Please refer to Docket ID
NRC–2017–0134 when contacting the
NRC about the availability of
information regarding this document.
You may obtain publicly-available
information related to this document
using any of the following methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2017–0134. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–415–3463;
email: Carol.Gallagher@nrc.gov. For
technical questions, contact the
individual listed in the FOR FURTHER
INFORMATION CONTACT section of this
document.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may obtain publiclyavailable documents online in the
ADAMS Public Documents collection at
https://www.nrc.gov/reading-rm/
adams.html. To begin the search, select
‘‘ADAMS Public Documents’’ and then
select ‘‘Begin Web-based ADAMS
Search.’’ For problems with ADAMS,
please contact the NRC’s Public
Document Room (PDR) reference staff at
1–800–397–4209, 301–415–4737, or by
email to pdr.resource@nrc.gov. The
ADAMS accession number for each
document referenced in this document
(if that document is available in
ADAMS) is provided the first time that
a document is referenced.
• 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.
FOR FURTHER INFORMATION CONTACT: YenJu Chen, Office of Nuclear Material
Safety and Safeguards, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555; telephone: 301–415–1018;
email: Yen-ju.Chen@nrc.gov.
SUPPLEMENTARY INFORMATION:
I. Introduction
The NRC is reviewing an exemption
request from Entergy, dated November
9, 2016 (ADAMS Accession No.
ML16319A102), and supplemented by
letter dated January 9, 2017 (ADAMS
E:\FR\FM\06JNN1.SGM
06JNN1
Agencies
[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]
-----------------------------------------------------------------------
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.
-----------------------------------------------------------------------
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 https://www.regulations.gov and search for Docket ID NRC-2017-0131. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: TWFN-8-D36M, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-5411; email: shirley.rohrer@nrc.gov.
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 https://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 https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in this
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2017-0131, 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 https://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
[[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 https://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of the amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission by
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 https://www.nrc.gov/site-help/e-submittals.html. Participants may not submit
paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at hearing.docket@nrc.gov, or by
telephone at 301-415-1677, to (1) request a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign submissions and access the E-Filing
system for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition or
other adjudicatory document (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a
digital ID certificate and a docket has been created, the participant
can then submit adjudicatory documents. Submissions must be in Portable
Document Format (PDF). Additional guidance on PDF submissions is
available on the NRC's public Web site at https://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the
time the document is submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public Web site at https://www.nrc.gov/site-help/e-submittals.html, by
email to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m.
and 6 p.m., Eastern Time, Monday through Friday, excluding government
holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
stating why there is good cause for not filing electronically and
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff. Participants filing adjudicatory documents in this
manner are responsible for serving the document on all other
participants. Filing is considered complete by first-class mail as of
the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service. A presiding officer, having granted an
exemption request from using E-Filing, may require a participant or
party to use E-Filing if the presiding officer subsequently determines
that the reason for granting the exemption from use of E-Filing no
longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the
Commission or the presiding officer. If you do not have an NRC-issued
digital ID certificate as described above, click cancel when the link
requests certificates and you will be automatically directed to the
NRC's electronic hearing dockets where you will be able to access any
publicly-available documents in a particular hearing docket.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or personal phone
numbers in their filings, unless an NRC regulation or other law
requires submission of such information. For example, in some
instances, individuals provide home addresses in order to demonstrate
proximity to a facility or site. With respect to copyrighted works,
except for limited excerpts that serve the purpose of the adjudicatory
filings and would constitute a Fair Use application, participants are
requested not to include copyrighted materials in their submission.
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For 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