[Federal Register Volume 82, Number 127 (Wednesday, July 5, 2017)]
[Notices]
[Pages 31089-31106]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-13804]
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NUCLEAR REGULATORY COMMISSION
[NRC-2017-0152]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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SUMMARY: Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954,
as amended (the Act), the U.S. Nuclear Regulatory Commission (NRC) is
publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued, and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person. This biweekly notice includes all notices of
amendments issued, or proposed to be issued, from June 3, 2017 to June
19, 2017. The last biweekly notice was published on June 19, 2017.
DATES: Comments must be filed by August 4, 2017. A request for a
hearing must be filed by September 5, 2017.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0152. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: TWFN-8-D36M, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Kay Goldstein, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-1506, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2017-0152, facility name, unit
number(s), plant docket number, application date, and subject when
contacting the NRC about the availability of information for this
action. You may obtain publicly-available information related to this
action by any of the following methods:
Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0152.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in this
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One
[[Page 31090]]
White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2017-0152, facility name, unit
number(s), plant docket number, application date, and subject in your
comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment
period such that failure to act in a timely way would result, for
example in derating or shutdown of the facility. If the Commission
takes action prior to the expiration of either the comment period or
the notice period, it will publish in the Federal Register a notice of
issuance. If the Commission makes a final no significant hazards
consideration determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and petition for leave to intervene
(petition) with respect to the action. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309. The NRC's regulations are accessible
electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place
[[Page 31091]]
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 no later
than 60 days from the date of publication of this notice. The petition
must be filed in accordance with the filing instructions in the
``Electronic Submissions (E-Filing)'' section of this document, and
should meet the requirements for petitions set forth in this section,
except that under 10 CFR 2.309(h)(2) a State, local governmental body,
or federally recognized Indian Tribe, or agency thereof does not need
to address the standing requirements in 10 CFR 2.309(d) if the facility
is located within its boundaries. Alternatively, a State, local
governmental body, Federally-recognized Indian Tribe, or agency thereof
may participate as a non-party under 10 CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
the discretion of the presiding officer, be permitted to make a limited
appearance pursuant to the provisions of 10 CFR 2.315(a). A person
making a limited appearance may make an oral or written statement of
his or her position on the issues but may not otherwise participate in
the proceeding. A limited appearance may be made at any session of the
hearing or at any prehearing conference, subject to the limits and
conditions as may be imposed by the presiding officer. Details
regarding the opportunity to make a limited appearance will be provided
by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for leave to intervene (petition), any
motion or other document filed in the proceeding prior to the
submission of a request for hearing or petition to intervene, and
documents filed by interested governmental entities that request to
participate under 10 CFR 2.315(c), must be filed in accordance with the
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR
46562, August 3, 2012). The E-Filing process requires participants to
submit and serve all adjudicatory documents over the internet, or in
some cases to mail copies on electronic storage media. Detailed
guidance on making electronic submissions may be found in the Guidance
for Electronic Submissions to the NRC and on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit
paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to (1) request a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign submissions and access the E-Filing
system for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition or
other adjudicatory document (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a
digital ID certificate and a docket has been created, the participant
can then submit adjudicatory documents. Submissions must be in Portable
Document Format (PDF). Additional guidance on PDF submissions is
available on the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the
time the document is submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public Web site at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m.
and 6 p.m., Eastern Time, Monday through Friday, excluding government
holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
stating why there is good cause for not filing electronically and
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff. Participants filing adjudicatory documents in this
manner are responsible for serving the document on all other
participants. Filing is considered complete by first-class mail as of
the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service. A presiding officer, having granted an
exemption request from using E-Filing, may require a participant or
party to use E-Filing if the presiding officer subsequently determines
that the reason for granting the exemption from use of E-Filing no
longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
https://adams.nrc.gov/ehd, unless excluded
[[Page 31092]]
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, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: April 3, 2017, as supplemented by
letters dated April 3, 2017, and May 2, 2017. Publicly-available
versions are in ADAMS under Accession Nos. ML17093A787, ML17093A796,
and ML17122A223, respectively.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to extend the required
frequency of certain 18-month Surveillance Requirements (SRs) to 24
months to accommodate a 24-month refueling cycle. In addition, the
proposed amendment would revise certain programs in TS Section 5.5,
``Programs and Manuals,'' to change 18-month frequencies to 24 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment changes the surveillance frequency from
18 months to 24 months for SRs in the TSs that are normally a
function of the refueling interval. Duke Energy Progress, LLC's
evaluations have shown that the reliability of protective
instrumentation and equipment will be preserved for the maximum
allowable surveillance interval.
The proposed change does not involve any change to the design or
functional requirements of the associated systems. That is, the
proposed TS change neither degrades the performance of, nor
increases the challenges to any safety systems assumed to function
in the plant safety analysis. The proposed change will not give rise
to any increase in operation power level, fuel operating limits or
effluents. The proposed change does not affect any accident
precursors since no accidents previously evaluated relate to the
frequency of surveillance testing and the revision to the frequency
does not introduce any accident initiators. The proposed change does
not impact the usefulness of the SRs in evaluating the operability
of required systems and components or the manner in which the
surveillances are performed.
In addition, evaluation of the proposed TS change demonstrates
that the availability of equipment and systems required to prevent
or mitigate the radiological consequences of an accident is not
significantly affected because of the availability of redundant
systems and equipment or the high reliability of the equipment.
Since the impact on the systems is minimal, it is concluded that the
overall impact on the plant safety analysis is negligible.
Furthermore, an historical review of surveillance test results
and associated maintenance records indicates there is no evidence of
any failure that would invalidate the above conclusions. Therefore,
the proposed TS change 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 amendment does not require a change to the plant
design nor the mode of plant operation. No new or different
equipment is being installed. No installed equipment is being
operated in a different manner. As a result, no new failure modes
are being introduced. In addition, the proposed change does not
impact the usefulness of the SRs in evaluating the operability of
required systems and components or the manner in which the
surveillances are performed. Furthermore, an historical review of
surveillance test results and associated maintenance records
indicates there is no evidence of any failure that would invalidate
the above conclusions. Therefore, the implementation of the proposed
change will not create the possibility for an accident of a new or
different type than previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment changes the surveillance frequency from
18 months to 24 months for SRs in the TSs that are normally a
function of the refueling interval. SR 3.0.2 would allow a maximum
surveillance interval of 30 months for these surveillances. Although
the proposed change will result in an increase in the interval
between surveillance tests, the impact on system availability is
small based on other, more frequent testing that is performed, the
existence of redundant systems and equipment or overall system
reliability. There is no evidence of any time-dependent failures
that would impact the availability of the systems. The proposed
change does not significantly impact the condition or performance of
structures, systems and components relied upon for accident
mitigation. This change does not alter the existing TS allowable
values or analytical limits. The existing operating margin between
plant conditions and actual plant setpoints is not significantly
reduced due to these changes. The assumptions and results in any
safety analyses are not significantly impacted. Therefore, the
proposed change does not involve a significant reduction in margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel,
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A,
Charlotte, NC 28202.
NRC Branch Chief: Undine S. Shoop.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: April 24, 2017. A publicly-available
version is in ADAMS under Accession No. ML17114A398.
Description of amendment request: The amendment would revise
Technical Specification requirements regarding steam generator tube
inspections and reporting as described in Technical Specification Task
Force (TSTF) Traveler TSTF-510, Revision 2, ``Revision to Steam
Generator Program Inspection Frequencies and Tube Sample Selection,''
using the Consolidated Line Item Improvement Process for Arkansas
Nuclear One, Unit No. 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
[[Page 31093]]
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Steam Generator (SG) Program to
modify the frequency of verification of SG tube integrity and SG
tube sample selection. A steam generator tube rupture (SGTR) event
is one of the design basis accidents that are analyzed as part of a
plant's licensing basis. The proposed SG tube inspection frequency
and sample selection criteria will continue to ensure that the SG
tubes are inspected such that the probability of a[n] SGTR is not
increased. The consequences of a[n] SGTR are bounded by the
conservative assumptions in the design basis accident analysis. The
proposed change will not cause the consequences of a[n] SGTR to
exceed those assumptions.
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 changes to the SG Program will not introduce any
adverse changes to the plant design basis or postulated accidents
resulting from potential tube degradation. The proposed change does
not affect the design of the SGs or their method of operation. In
addition, the proposed change does not impact any other plant system
or component.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
also isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a[n]
SG is maintained by ensuring the integrity of its tubes.
SG tube integrity is a function of the design, environment, and
the physical condition of the tube. The proposed change does not
affect tube design or operating environment. The proposed change
will continue to require monitoring of the physical condition of the
SG tubes such that there will not be a reduction in the margin of
safety compared to the current requirements.
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: Ms. Anna Vinson Jones, Senior Counsel,
Entergy Services, Inc., 101 Constitution Avenue NW., Suite 200 East,
Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: April 24, 2017. A publicly-available
version is in ADAMS under Accession No. ML17114A399.
Description of amendment request: The amendment would revise
Technical Specification requirements regarding steam generator tube
inspections and reporting as described in Technical Specifications Task
Force (TSTF) Traveler TSTF-510, Revision 2, ``Revision to Steam
Generator Program Inspection Frequencies and Tube Sample Selection,''
using the Consolidated Line Item Improvement Process for Arkansas
Nuclear One, Unit No. 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Steam Generator (SG) Program to
modify the frequency of verification of SG tube integrity and SG
tube sample selection. A steam generator tube rupture (SGTR) event
is one of the design basis accidents that are analyzed as part of a
plant's licensing basis. The proposed SG tube inspection frequency
and sample selection criteria will continue to ensure that the SG
tubes are inspected such that the probability of a[n] SGTR is not
increased. The consequences of a[n] SGTR are bounded by the
conservative assumptions in the design basis accident analysis. The
proposed change will not cause the consequences of a[n] SGTR to
exceed those assumptions.
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 changes to the SG Program will not introduce any
adverse changes to the plant design basis or postulated accidents
resulting from potential tube degradation. The proposed change does
not affect the design of the SGs or their method of operation. In
addition, the proposed change does not impact any other plant system
or component.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
also isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a[n]
SG is maintained by ensuring the integrity of its tubes.
SG tube integrity is a function of the design, environment, and
the physical condition of the tube. The proposed change does not
affect tube design or operating environment. The proposed change
will continue to require monitoring of the physical condition of the
SG tubes such that there will not be a reduction in the margin of
safety compared to the current requirements.
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: Ms. Anna Vinson Jones, Senior Counsel,
Entergy Services, Inc., 101 Constitution Avenue NW., Suite 200 East,
Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana
Date of amendment request: March 28, 2017. A publicly-available
version is
[[Page 31094]]
in ADAMS under Accession No. ML17087A551.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.8.1.3, ``Diesel Fuel Oil,'' by
relocating the current stored diesel fuel oil numerical volume
requirements from the TS to the TS Bases. In addition, the proposed
amendment would revise TS 3.8.1.1, ``A.C. [Alternating Current]
Sources--Operating,'' and TS 3.8.1.2, ``A.C. Sources--Shutdown,'' to
relocate the specific numerical value for feed tank fuel oil volume to
the TS Bases and replace it with the feed tank time requirement. The
proposed changes are consistent with Technical Specifications Task
Force (TSTF) Traveler TSTF-501, Revision 1, ``Relocate Fuel Oil and
Lube Oil Volume Values to Licensee Control.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes revise [TS] 3.8.1.3 (Diesel Fuel Oil) by
removing the current stored diesel fuel oil numerical volume
requirements from the TS and replacing them with diesel operating
time requirements. The specific volume of fuel oil equivalent to a 7
and 6 day supply is calculated using the NRC approved methodology
described in Regulatory Guide 1.137, Revision 1, ``Fuel-Oil Systems
for Standby Diesel Generators'' and [American Nuclear Standards
Institute (ANSI)] N195-1976, ``Fuel Oil Systems for Standby Diesel-
Generators'' using the time dependent load method as approved in
Waterford 3 License Amendment 157. Because the requirement to
maintain a 7 day supply of diesel fuel oil is not changed and is
consistent with the assumptions in the accident analyses, and the
actions taken when the volume of fuel oil is less than a 6 day
supply have not changed, neither the probability nor the
consequences of any accident previously evaluated will be affected.
The proposed change also removes the TS 3.8.1.1 and TS 3.8.1.2
diesel feed tank fuel oil numerical volume requirements and replaces
them with the diesel one hour diesel generator operation
requirement. The specific volume and time is not changed and is
consistent with the existing plant design basis to support a diesel
generator under accident load conditions.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The change 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. The change
does not alter assumptions made in the safety analysis but ensures
that the diesel generator operates as assumed in the accident
analysis. The proposed change 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 a
margin of safety?
Response: No.
The proposed changes revise [TS] 3.8.1.3 (Diesel Fuel Oil) by
removing the current stored diesel fuel oil numerical volume
requirements from the TS and replacing them with diesel operating
time requirements. As the bases for the existing limits on diesel
fuel oil are not changed, no change is made to the accident analysis
assumptions and no margin of safety is reduced as part of this
change.
The proposed change also removes the TS 3.8.1.1 and TS 3.8.1.2
diesel feed tank fuel oil numerical volume requirements and replaces
them with the diesel one hour diesel generator operation
requirement. As the basis for the existing limits on diesel fuel oil
are not changed, no change is made to the accident analysis
assumptions and no margin of safety is reduced as part of this
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: Ms. Anna Vinson Jones, Senior Counsel,
Entergy Services, Inc., 101 Constitution Avenue NW., Suite 200 East,
Washington, DC 20001.
NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket No. 50-277,
Peach Bottom Atomic Power Station (PBAPS), Unit 2, York and Lancaster
Counties, Pennsylvania
Date of amendment request: May 19, 2017. A publicly-available
version is in ADAMS under Accession No. ML17139D357.
Description of amendment request: The amendment would revise the
Technical Specifications (TSs) to decrease the number of safety relief
valves and safety valves required to be operable when operating at a
power level less than or equal to 3358 megawatts thermal (MWt). This
change would be in effect for the current PBAPS, Unit 2, Cycle 22 that
is scheduled to end in October 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, which is presented below, with NRC staff edits in square
brackets:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change would revise TS Section 3.4.3 to decrease
the required number of Safety Relief Valves (SRVs) and Safety Valves
(SVs) from a total of 13 to 12, under reduced reactor thermal power
operation of 3358 MWt (approximately 85% of Current Licensed Thermal
Power (CLTP)). A compensatory reduction in maximum allowed reactor
power to 3358 MWt has been determined to conservatively offset the
impact/effects of operation with an additional (up to 2) SRVs/SVs
Out-of-Service. The Reactor Pressure Vessel (RPV) overpressure
protection capability of the 12 operable SRVs and SVs is adequate at
the lower power level to ensure the ASME [American Society of
Mechanical Engineers] code allowable peak pressure limits are not
exceeded. With the maximum thermal power limitation condition, the
proposed change has no adverse effect on plant operation, or the
availability or operation of any accident mitigation equipment. The
plant response to the design basis accidents, Anticipated
Operational Occurrence (AOO) events and Special Events remains
bounded by existing analyses. The proposed change does not require
any new or unusual operator actions. The proposed change does not
introduce any new failure modes that could result in a new or
different accident. The SRVs and SVs are not being modified or
operated differently and will continue to operate to meet the design
basis requirements for RPV overpressure protection. The proposed
change does not alter the manner in which the RPV overpressure
protection system is operated and functions and thus, there is no
significant impact on reactor operation. There is no change being
made to safety limits or limiting safety system settings that would
adversely affect plant safety as a result of the proposed change.
For PBAPS, the limiting overpressure AOO event is the main steam
isolation valve closure with scram on high flux (MSIVF). The PBAPS
ATWS [anticipated transients without scram] Special Event evaluation
considered the limiting cases for RPV overpressure and is analyzed
under two cases: (1) Main Steam Isolation Valve Closure (MSIVC) and
(2) Pressure Regulator Failure Open (PRFO). These events were
analyzed under the proposed conditions and it was confirmed that the
existing analyses remain bounding for the condition of adding a
[[Page 31095]]
second SRV/SV Out-of-Service with a limited maximum operating power
level of 3358 MWt.
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 would revise TS Section 3.4.3 to decrease
the required number of SRVs and SVs from a total of 13 to 12, under
reduced reactor thermal power operation of 3358 MWt (approximately
85% of CLTP). A compensatory reduction in maximum allowed reactor
power to 3358 MWt has been determined to conservatively offset the
impact/effects of operation with an additional (up to 2) SRVs/SVs
Out-of-Service. The RPV overpressure protection capability of the 12
operable SRVs and SVs is adequate at the lower power level to ensure
the ASME code allowable peak pressure limits are not exceeded. The
SRVs and SVs are not being modified or operated differently and will
continue to operate to meet the design basis requirements for RPV
overpressure protection. The proposed change does not introduce any
new failure modes that could result in a new or different accident.
The proposed reactor thermal power restriction of 3358 MWt is within
the existing normal operating domain and no new or special operating
actions are necessary to operate at the intermediate power level.
The proposed change does not alter the manner in which the RPV
overpressure protection system is operated and functions and thus,
there is no new failure mechanisms for the overpressure protection
system. The plant response to the design basis accidents, AOO events
and Special Events remains bounded by existing analyses. [These]
events were analyzed under the proposed conditions and it was
confirmed that the existing analyses remain bounding for the
condition of adding a second SRV/SV Out-of-Service with a limited
maximum operating power level of 3358 MWt.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established though the design of the
plant structures, systems and components, the parameters within
which the plant is operated, and the establishment of setpoints for
the actuation of equipment relied upon to respond to an event. The
proposed change does not change the setpoints at which the
protective actions are initiated. The proposed change would revise
TS Section 3.4.3 to decrease the required number of SRVs and SVs
under reduced reactor thermal power operation of 3358 MWt
(approximately 85% of CLTP). A compensatory reduction in maximum
allowed reactor power to 3358 MWt has been determined to
conservatively offset the impact/effects of operation with an
additional (up to 2) SRVs/SVs Out-of-Service. The RPV overpressure
protection capability of the 12 operable SRVs and SVs is adequate at
the lower power level to ensure the ASME code allowable peak
pressure limits are not exceeded. The plant response to the design
basis accidents, AOO events and Special Events remains bounded by
existing analyses. These events were analyzed under the proposed
conditions and it was confirmed that the existing analyses remain
bounding for the condition of adding a second SRV/SV Out-of-Service
with a limited maximum operating power level of 3358 MWt.
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: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Rd., Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station (CPS), Unit No.1, DeWitt County, Illinois
Date of amendment request: May 4, 2017. A publicly-available
version is in ADAMS under Accession No. ML17124A121.
Description of amendment request: The proposed change would delete
a surveillance requirement (SR) Note associated with technical
specification (TS) 3.5.1, ``ECCS [emergency core cooling system]--
Operating,'' TS 3.5.2, ``ECCS--Shutdown,'' and TS 3.6.1.7, ``Residual
Heat Removal (RHR) Containment Spray System,'' to more appropriately
reflect the RHR system design, and ensure the RHR system operation is
consistent with the TS limiting condition for operation (LCO)
requirements. In addition, the proposed amendment would insert a Note
in the LCO for TSs 3.5.1, 3.5.2, 3.6.1.7, 3.6.1.9, ``Feedwater Leakage
Control System,'' and 3.6.2.3, ``Residual Heat Removal (RHR)
Suppression Pool Cooling,'' to clarify that one of the required
subsystems in each of the affected TS sections may be inoperable during
alignment and operation of the RHR system for shutdown cooling (SDC)
with the reactor steam dome pressure less than the RHR cut in
permissive value.
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.
No physical changes to the facility will occur as a result of
this proposed amendment. The proposed changes will not alter the
physical design. The current TS (CTS) Note in SR 3.5.1.4, SR
3.5.2.4, and 3.6.1.7 could make CPS susceptible to potential water
hammer in the RHR system while operating in the SDC mode of RHR in
MODE 3 when swapping from the SDC to LPCI [low-pressure coolant
injection] and RHR containment spray modes of RHR. Deletion of the
Note from SR 3.5.1.2, SR 3.5.2.4, and SR 3.6.1.7.1 will eliminate
the risk for cavitation of the pump and voiding in the suction
piping, thereby avoiding the potential to damage the RHR system,
including water hammer. The addition of proposed TS note to LCO
3.5.1, LCO 3.5.2, LCO 3.6.1.7, LCO 3.6.1.9, and LCO 3.6.2.3 will re-
establish consistency of the CPS RHR system design with the original
TS requirements.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the physical design, safety
limits, or safety analysis assumptions associated with the operation
of the plant. Accordingly, the change does not introduce any new
accident initiators, nor does it reduce or adversely affect the
capabilities of any plant structure, system, or component to perform
their safety function. Deletion of the Note from SR 3.5.1.2, SR
3.5.2.4 and SR 3.6.1.7.1 is appropriate because current TSs could
put the plant at risk for potential cavitation of the pump and
voiding in the suction piping, resulting in potential to damage the
RHR system, including water hammer. The addition of proposed TS note
to LCO 3.5.1, LCO 3.5.2, LCO 3.6.1.7, LCO 3.6.1.9, and LCO 3.6.2.3
will re-establish consistency of the CPS RHR system design with the
original TS requirements.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change conforms to NRC regulatory guidance
regarding the content of plant Technical Specifications. The
proposed change does not alter the physical design, safety limits,
or safety analysis assumptions associated with the operation of the
plant.
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
[[Page 31096]]
review it appears the three standards of 10 CFR 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: David J. Wrona.
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit No.1, DeWitt County, Illinois
Date of amendment request: May 1, 2017. A publicly-available
version is in ADAMS under Accession No. ML17121A517.
Description of amendment request: The proposed change replaces
existing technical specification (TS) requirements related to
operations with a potential for draining the reactor vessel (OPDRVs)
with new requirements on reactor pressure vessel (RPV) water inventory
control (WIC) to protect Safety Limit 2.1.1.3. Safety Limit 2.1.1.3
requires reactor vessel water level to be greater than the top of
active irradiated fuel.
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 replaces existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.1.3. Draining of RPV water inventory in Mode 4 (i.e., cold
shutdown) and Mode 5 (i.e., refueling) is not an accident previously
evaluated and, therefore, replacing the existing TS controls to
prevent or mitigate such an event with a new set of controls has no
effect on any accident previously evaluated. RPV water inventory
control in Mode 4 or Mode 5 is not an initiator of any accident
previously evaluated. The existing OPDRV controls or the proposed
RPV WIC controls are not mitigating actions assumed in any accident
previously evaluated.
The proposed change reduces the probability of an unexpected
draining event (which is not a previously evaluated accident) by
imposing new requirements on the limiting time in which an
unexpected draining event could result in the reactor vessel water
level dropping to the top of the active fuel (TAF). These controls
require cognizance of the plant configuration and control of
configurations with unacceptably short drain times. These
requirements reduce the probability of an unexpected draining event.
The current TS requirements are only mitigating actions and impose
no requirements that reduce the probability of an unexpected
draining event.
The proposed change reduces the consequences of an unexpected
draining event (which is not a previously evaluated accident) by
requiring an Emergency Core Cooling System (ECCS) subsystem to be
operable at all times in Modes 4 and 5. The current TS requirements
do not require any water injection systems, ECCS or otherwise, to be
operable in certain conditions in Mode 5. The change in requirement
from two ECCS subsystem to one ECCS subsystem in Modes 4 and 5 does
not significantly affect the consequences of an unexpected draining
event because the proposed Actions ensure equipment is available
within the limiting drain time that is as capable of mitigating the
event as the current requirements. The proposed controls provide
escalating compensatory measures to be established as calculated
drain times decrease, such as verification of a second method of
water injection and additional confirmations that secondary
containment and/or filtration would be available if needed.
The proposed change reduces or eliminates some requirements that
were determined to be unnecessary to manage the consequences of an
unexpected draining event, such as automatic initiation of an ECCS
subsystem and control room ventilation. These changes do not affect
the consequences of any accident previously evaluated since a
draining event in Modes 4 and 5 is not a previously evaluated
accident and the requirements are not needed to adequately respond
to a draining event.
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 replaces existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.1.3. The proposed change will not alter the design
function of the equipment involved. Under the proposed change, some
systems that are currently required to be operable during OPDRVs
would be required to be available within the limiting drain time or
to be in service depending on the limiting drain time. Should those
systems be unable to be placed into service, the consequences are no
different than if those systems were unable to perform their
function under the current TS requirements.
The event of concern under the current requirements and the
proposed change is an unexpected draining event. The proposed change
does not create new failure mechanisms, malfunctions, or accident
initiators that would cause a draining event or a new or different
kind of accident not previously evaluated or included in the design
and licensing bases.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change replaces existing TS requirements related to
OPDRVs with new requirements on RPV WIC. The current requirements do
not have a stated safety basis and no margin of safety is
established in the licensing basis. The safety basis for the new
requirements is to protect Safety Limit 2.1.1.3. New requirements
are added to determine the limiting time in which the RPV water
inventory could drain to the top of the fuel in the reactor vessel
should an unexpected draining event occur. Plant configurations that
could result in lowering the RPV water level to the TAF within one
hour are now prohibited. New escalating compensatory measures based
on the limiting drain time replace the current controls. The
proposed TS establish a safety margin by providing defense-in-depth
to ensure that the Safety Limit is protected and to protect the
public health and safety. While some less restrictive requirements
are proposed for plant configurations with long calculated drain
times, the overall effect of the change is to improve plant safety
and to add safety margin.
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 the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: David J. Wrona.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station (LGS), Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: April 24, 2017. A publicly available
version is in ADAMS under Accession No. ML17115A087.
Description of amendment request: The amendments would revise the
LGS, Units 1 and 2, Technical Specifications (TSs) to a set of Improved
Technical Specifications (ITS) based on NUREG-1433, Revision 4,
``Standard Technical Specifications--General Electric Plants, BWR/4,''
published April 2012. Specifically, the amendments would relocate TS
Section 3.3.7.12, ``Offgas Gas Monitoring Instrumentation''; TS
3.11.2.5, ``Explosive Gas Mixture''; and Surveillance Requirement (SR)
4.11.2.6.1, which requires continuously monitoring the main condenser
gaseous effluent to the LGS Offsite Dose Calculation Manual or to the
LGS Technical Requirements Manual. In
[[Page 31097]]
addition, associated with the relocation of the main condenser offgas
noble gas activity monitor, (1) SR 4.11.2.6.2.b will be changed to
account for the relocated instrument's requirements, and (2) associated
with the relocation of the explosive gas mixture instrumentation and
gaseous effluent TS sections, a new TS Program Section, 6.8.4.l,
``Explosive Gas Monitoring Program,'' will be added to TS Section 6.8,
``Procedures and Programs.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes relocate certain operability and
surveillance requirements for the Main Condenser Offgas Monitoring
Instrumentation and Gaseous Effluents limits from the Limerick
Generating Station (LGS) Technical Specifications (TS) to a
licensee-controlled document under the control of 10 CFR 50.59 or
under the control of regulatory requirements applicable to the
licensee-controlled document. A new TS Administrative Program is
proposed to be added to ensure the limit for Main Condenser Offgas
hydrogen concentration is maintained.
The proposed changes do not alter the physical design of any
plant structure, system, or component; therefore, the proposed
changes have no adverse effect on plant operation, or the
availability or operation of any accident mitigation equipment. The
plant response to the design basis accidents does not change.
Operation or failure of the Main Condenser Offgas Radioactivity and
Hydrogen Monitors capability are not assumed to be an initiator of
any analyzed event in the Updated Final Safety Analysis Report
(UFSAR) and cannot cause an accident. Whether the requirements for
the Main Condenser Offgas Radioactivity and Hydrogen Monitor
capability are located in TS or another licensee-controlled document
has no effect on the probability or consequences of any accident
previously evaluated.
The proposed changes conform to NRC regulatory requirements
regarding the content of plant TS as identified in 10 CFR 50.36, and
also the guidance as approved by the NRC in NUREG-1433, ``Standard
Technical Specifications--General Electric BWR/4 Plants.''
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes relocate certain operability and
surveillance requirements for the Main Condenser Offgas Monitoring
Instrumentation and Gaseous Effluents limits from the LGS TS to a
licensee-controlled document under the control of 10 CFR 50.59 or
under the control of regulatory requirements applicable to the
licensee-controlled document. A new TS Administrative Program is
proposed to be added to ensure the limit for Main Condenser Offgas
hydrogen concentration is maintained.
The proposed changes do not alter the plant configuration (no
new or different type of equipment is being installed) or require
any new or unusual operator actions. The proposed changes do not
alter the safety limits or safety analysis assumptions associated
with the operation of the plant. The proposed changes do not
introduce any new failure modes that could result in a new accident.
The proposed changes do not reduce or adversely affect the
capabilities of any plant structure, system, or component in the
performance of their safety function. Also, the response of the
plant and the operators following the design basis accidents is
unaffected by the proposed changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes relocate certain operability and
surveillance requirements for the Main Condenser Offgas Monitoring
Instrumentation and Gaseous Effluents limits from the LGS TS to a
licensee-controlled document under the control of 10 CFR 50.59 or
under the control of regulatory requirements applicable to the
licensee-controlled document. A new TS Administrative Program is
proposed to be added to ensure the limit for the Main Condenser
Offgas hydrogen concentration is maintained. The relocated TS
requirements do not meet any of the 10 CFR 50.36c(2)(ii) criteria on
items for which a TS must be established.
The proposed changes have no adverse effect on plant operation,
or the availability or operation of any accident mitigation
equipment. The plant response to the design basis accidents does not
change. The proposed changes do not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analyses. There is no change being made to
safety analysis assumptions, safety limits or limiting safety system
settings that would adversely affect plant safety as a result of the
proposed changes.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant (PNPP), Unit No. 1, Lake County, Ohio
Date of amendment request: April 26, 2017. A publicly-available
version is in ADAMS under Accession No. ML17116A575.
Description of amendment request: The proposed amendment would
revise the PNPP Environmental Protection Plan (nonradiological) to
clarify and enhance wording, to remove duplicative or outdated program
information, and to relieve the burden of submitting unnecessary or
duplicative information to the NRC.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment involves changes to the Environmental
Protection Plan (EPP), which provides for protection of
nonradiological environmental values during operation of the nuclear
facility. The proposed amendment does not change the objectives of
the EPP, does not change the way the plant is maintained or
operated, and does not affect any accident mitigating feature or
increase the likelihood of malfunction for plant structures, systems
and components.
The proposed amendment will not change any of the analyses
associated with the PNPP Updated Safety Analysis Report Chapter 15
accidents because plant operation, plant structures, systems,
components, accident initiators, and accident mitigation functions
remain unchanged.
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 amendment involves changes to the EPP, which
provides for protection of nonradiological environmental values
during operation of the nuclear facility. The proposed amendment
does not involve a physical alteration of the plant. No new or
different type of equipment will be installed, and there are no
physical modifications to existing installed equipment associated
with the proposed changes. The
[[Page 31098]]
proposed amendment does not change the way the plant is operated or
maintained and does not create a credible failure mechanism,
malfunction or accident initiator not already considered in the
design and licensing basis.
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.
Safety margins are applied to design and licensing basis
functions and to the controlling values of parameters to account for
various uncertainties and to avoid exceeding regulatory or licensing
limits. The proposed amendment involves changes to the EPP, which
provides for protection of nonradiological environmental values
during operation of the nuclear facility. The proposed amendment
does not involve a physical change to the plant, does not change
methods of plant operation within prescribed limits, or affect
design and licensing basis functions or controlling values of
parameters for plant systems, structures, and components.
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: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: David J. Wrona.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: May 2, 2017. A publicly-available
version is in ADAMS under Accession No. ML17144A294.
Description of amendment request: The amendments would revise the
St. Lucie Plant Unit Nos. 1 and 2 Renewed Facility Operating Licenses,
Nos. DPR-67 and NPF-16, respectively, fire protection license
conditions. The revisions would incorporate new references into these
license conditions that propose and approve a revision to plant
modifications previously approved in the March 31, 2016, NRC issuance
of amendments regarding transition to a risk-informed, performance-
based fire protection program in accordance with 10 CFR 50.48(c), dated
March 21, 2016 (ADAMS Accession No. ML15344A346) (known as the National
Fire Protection Association Standard 805 (NFPA 805)).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes are clarifications to methods applied to
ensure compliance with NFPA 30, section 2348. The revised methods
comply with NFPA 30, section 2348. This LAR [license amendment
request] is essentially an administrative change to revise the
letter referenced by the Fire Protection Transition License
Conditions. The actual design changes and any related procedural
changes are being managed separately from this LAR per 10 CFR 50.59.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained. The proposed changes do not
adversely affect the ability of structures, systems and components
(SSCs) to perform their intended safety function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change does not increase the probability or
consequence of an accident.
Therefore, 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 changes are clarifications to methods applied to
ensure compliance with NFPA 30, section 2348. The revised methods of
compliance align with NFPA 30, section 2348, and will not result in
new or different kinds of accidents. This LAR is essentially an
administrative change to revise the letter referenced by the Fire
Protection Transition License Conditions. The actual design changes
and any related procedural changes are being managed separately from
this LAR per 10 CFR 50.59.
The requirements in NFPA 30 address only fire protection. The
impacts of fire effects on the plant have been evaluated. The
proposed amendment does not involve new failure mechanisms or
malfunctions that could initiate a new or different kind of accident
beyond those already analyzed in the Unit 1 and Unit 2 UFSARs
[updated final safety analysis reports].
Therefore, this change does not create the possibility of a new
or different kind of accident from an accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Operation of Plant St. Lucie (PSL) in accordance with the
proposed amendment does not involve a reduction in the margin of
safety. The proposed amendment does not alter the manner in which
safety limits, limiting safety system settings or limiting
conditions for operation are determined. The safety analysis
acceptance criteria are not affected by this change. The proposed
amendment does not adversely affect existing plant safety margins or
the reliability of equipment assumed to mitigate accidents in the
UFSAR. The proposed amendment does not adversely affect the ability
of SSCs to perform their design function. SSCs required to safely
shut down the reactor and to maintain it in a safe shutdown
condition remain capable of performing their design function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
NRC Branch Chief: Undine S. Shoop.
Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant (CNP), Units Nos. 1 and 2, Berrien County,
Michigan
Date of amendment request: May 23, 2017. A publicly-available
version is in ADAMS under Accession No. ML17146A073.
Description of amendment request: The proposed changes update the
emergency action levels (EALs) used at CNP, Unit Nos. 1 and 2, from the
current scheme based on Nuclear Management and Resources Council
(NUMARC) and National Environmental Studies Project (NESP) NUMARC/NESP-
007, ``Methodology for Development of Emergency Action Levels'' dated
January 1992, to a scheme based on Nuclear Energy Institute 99-01,
Revision 6, ``Development of Emergency Action Levels for Non-Passive
Reactors.''
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.
[[Page 31099]]
The proposed changes to the CNP EALs do not impact the physical
function of plant structures, systems, or components (SSC) or the
manner in which SSCs perform their design function. EALs are used as
criteria for determining the need for notification and participation
of local and State agencies, and for determining when and what type
of protective measures should be considered within and outside the
site boundary to protect health and safety. The proposed changes
neither adversely affect accident initiators or precursors, nor
alter design assumptions. The proposed changes do not alter or
prevent the ability of SSCs to perform their intended function to
mitigate the consequences of an initiating event within assumed
acceptance limits. No operating procedures or administrative
controls that function to prevent or mitigate accidents are affected
by the proposed changes.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to the CNP EALs do not involve any physical
changes to plant systems or equipment. The proposed changes do not
involve the addition of any new equipment. EALs are based on plant
conditions, so the proposed changes will not alter the design
configuration or the method of plant operation. The proposed changes
will not introduce failure modes that could result in a new or
different type of accident, and the change does not alter
assumptions made in the safety analysis. The proposed changes to the
CNP Emergency Plan are not initiators of any accidents.
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.
Margin of safety is associated with the ability of the fission
product barriers (i.e., fuel cladding, reactor coolant system
pressure boundary, and containment structure) to limit the level of
radiation dose to the public. The proposed changes to the CNP EALs
do not impact operation of the plant or its response to transient or
accidents. The changes do not affect the Technical Specifications or
the operating license. The proposed changes do not involve a change
in the method of plant operation, and no accident analyses will be
affected by the proposed changes.
Additionally, the proposed changes will not relax any criteria
used to establish safety limits and will not relax any safety system
settings. The safety analysis acceptance criteria are not affected
by these changes. The proposed changes will not result in plant
operation in configuration outside the design basis. The proposed
changes do not adversely affect systems that respond to safely shut
down the plant and to maintain the plant in a safe shutdown
condition. The emergency plan will continue to activate an emergency
response commensurate with the extent of degradation of plant
safety.
Plant safety margins are established through limiting conditions
for operation, limiting safety system settings, and safety limits
specified in the technical specifications. The proposed changes
involve references to available plant indications to assess
conditions for determination of entry into an emergency action
level. There is no change to these established safety margins as a
result of this 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: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
NRC Branch Chief: David J. Wrona.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: May 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17135A225.
Description of amendment request: The requested amendment proposes
to depart from combined license (COL) Appendix C information (with
corresponding changes to the associated plant-specific Tier 1
information) and involves associated Tier 2 information in the Updated
Final Safety Analysis Report (UFSAR). Specifically, proposed changes
clarify that there is more than one turbine building main sump and adds
a second sump pump for each of the two turbine building main sumps into
UFSAR Tier 2 and COL Appendix C (and associated plant-specific Tier 1)
information.
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 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 activity adds a second pump to each of the turbine building
main sumps, and identifies that there is more than one turbine
building sump. The reason for the additional pumps is to account for
an increase in volume due to the changes to the [condensate
polishing system (CPS)] rinse effluent flowpath from [component
cooling water system (CCW)] CCW to [waste water system (WWS)] WWS
via the Turbine Building sumps. The extra sump pumps will prevent
potential overflowing and flooding of the sumps during CPS rinse
operations. The CPS serves no safety-related function. By directing
the effluent to the turbine building sumps it is subject to
radiation monitoring. Under normal operating conditions, there are
no significant amounts of radioactive contamination within the CPS.
However, radioactive contamination of the CPS can occur as a result
of a primary to secondary leakage in the steam generator should a
steam generator tube leak develop while the CPS is in operation and
radioactive condensate is processed by the CPS. Radiation monitors
associated with the steam generator blowdown, steam generator, and
turbine island vents, drains and relief systems provide the means to
determine if the secondary side is radioactively contaminated. The
main turbine building sumps and sump pumps are not safety-related
components and do not interface with any systems, structures, or
components (SSC) accident initiator or initiating sequence of
events; thus, the probability of accidents evaluated within the
plant-specific UFSAR are not affected. The proposed changes do not
involve a change to the predicted radiological releases due to
accident conditions, thus the consequences of 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 to the non-safety waste water system (WWS)
do not affect any safety-related equipment, nor does it add any new
interface to safety-related SSCs. No system or design function or
equipment qualification is affected by this change. The changes do
not introduce a new failure mode, malfunction, or sequence of events
that could affect safety or safety-related equipment.
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 WWS is a nonsafety-related system that does not interface
with any safety-related equipment. The proposed changes to identify
that there is more than one turbine building sump and to add two
turbine building sump pumps do not affect any design code,
[[Page 31100]]
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 change.
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: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Jennifer Dixon-Herrity.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: May 16, 2017. A publicly-available
version is in ADAMS under Accession No. ML17137A107.
Description of amendment request: The requested amendment consist
of changes to inspections, tests, analyses, and acceptance criteria
(ITAAC) in combined license (COL) Appendix C, with corresponding
changes to the associated plant-specific Tier 1 information, to
consolidate a number of ITAAC to improve efficiency of the ITAAC
completion and closure process.
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 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 non-technical change to COL Appendix C will
consolidate, relocate and subsume redundant ITAAC in order to
improve and create a more efficient process for the ITAAC Closure
Notification submittals. No structure, system, or component (SSC)
design or function is affected. No design or safety analysis is
affected. The proposed changes do not affect any accident initiating
event or component failure, thus the probabilities of the accidents
previously evaluated are not affected. No function used to mitigate
a radioactive material release and no radioactive material release
source term is involved, thus the radiological releases in the
accident analyses 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 change to COL Appendix C does not affect the design
or function of any SSC, but will consolidate, relocate and subsume
redundant ITAAC in order to improve efficiency of the ITAAC
completion and closure process. The proposed changes would not
introduce a new failure mode, fault or sequence of events that could
result in a radioactive material release.
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 change to COL Appendix C to consolidate, relocate
and subsume redundant ITAAC in order to improve efficiency of the
ITAAC completion and closure process is considered non-technical and
would not affect any design parameter, function or analysis. There
would be no change to an existing design basis, design function,
regulatory criterion, or analysis. No safety analysis or design
basis acceptance limit/criterion is involved.
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: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment request: May 16, 2017. A publicly-available
version is in ADAMS under Accession No. ML17142A315.
Description of amendment request: The proposed amendment would
revise the Facility Operating Licenses for the San Onofre Nuclear
Generating Station (SONGS), Units 2 and 3, to reflect deletion of the
Cyber Security Plan from License Condition 2.E. This will allow
Southern California Edison (SCE) to terminate the SONGS Cyber Security
Plan and associated activities at the site. These changes will more
fully reflect the permanently shutdown and defueled status of the
facility, as well as the reduced scope of potential radiological
accidents and security concerns that exist during the decommissioning
process.
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 remove the San Onofre Nuclear Generating
Station (SONGS) Cyber Security Plan requirement does not alter
accident analysis assumptions, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected. The proposed
change does not require any plant modifications which affect the
performance capability of the structures, systems, and components
(SSCs) relied upon to mitigate the consequences of postulated
accidents, and has no impact on the probability or consequences of
an accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to remove the SONGS Cyber Security Plan
requirement does not alter accident analysis assumptions, add any
initiators, or affect the function of plant systems or the manner in
which systems are operated, maintained, modified, tested, or
inspected. The proposed change does not require any plant
modifications which affect the performance capability of the SSCs
relied upon to mitigate the consequences of postulated accidents,
and does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Plant safety margins are established through limiting conditions
for operation,
[[Page 31101]]
limiting safety system settings, and safety limits specified in the
technical specifications. The proposed change to the SONGS Cyber
Security Plan does not change these established safety margins.
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: Walker A. Matthews, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: Bruce Watson, CHP.
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 5, 2017. A publicly-available
version is in ADAMS under Accession No. ML17125A331.
Description of amendment request: The amendment request proposes to
depart from plant-specific Tier 1 emergency planning inspection, test,
analysis, and acceptance criteria (ITAAC) information and associated
combined license (COL) Appendix C information. The proposed changes do
not involve changes to the approved emergency plan or the plant-
specific Tier 2 Design Control Document (DCD). Specifically, the
requested amendment proposes to revise plant-specific emergency
planning inspections (ITAAC) in Appendix C of the VEGP Units 3 and 4
COLs. Also, proposed changes to COL Appendix C information also include
changes to the list of acronyms and abbreviations. Because, this
proposed change requires a departure from Tier 1 information in the
Westinghouse Electric Company's AP1000 Design 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 VEGP 3 and 4 emergency planning inspections, tests,
analyses, and acceptance criteria (ITAAC) provide assurance that the
facility has been constructed and will be operated in conformity
with the license, the provisions of the Act, and the Commission's
rules and regulations. The proposed changes do not affect the design
of a system, structure, or component (SSC) use to meet the design
bases of the nuclear plant. Nor do the changes affect the
construction or operation of the nuclear plant itself, so there is
no change to the probability or consequences of an accident
previously evaluated. Changing the VEGP 3 and 4 emergency planning
ITAAC and COL, Appendix C, list of acronyms and abbreviations do not
affect prevention and mitigation of abnormal events (e.g.,
accidents, anticipated operational occurrences, earthquakes, floods,
or turbine missiles) or their safety or design analyses. No safety-
related structure, system, component (SSC) or function is adversely
affected. The changes neither involve nor interface with any SSC
accident initiator or initiating sequence of events, so the
probabilities of the accidents evaluated in the Updated Final Safety
Analysis Report (UFSAR) are not affected. Because the changes do not
involve any safety-related SSC or function used to mitigate an
accident, the consequences of the accidents evaluated in the UFSAR
are not affected.
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 VEGP 3 and 4 emergency planning ITAAC provide assurance that
the facility has been constructed and will be operated in conformity
with the license, the provisions of the Act, and the Commissioner's
rules and regulations. The changes do not affect the design of an
SSC used to meet the design bases of the nuclear plant. Nor do the
changes affect the construction or operation of the nuclear plant.
Consequently, there is no new or different kind of accident from any
accident previously evaluated. The changes do not affect safety-
related equipment, nor do they affect equipment that, if it failed,
could initiate an accident or a failure of a fission product
barrier. In addition, the changes do not result in a new failure
mode, malfunction, or sequence of events that could affect safety or
safety-related equipment.
No analysis is adversely affected. No system or design function
or equipment qualification is adversely affected by the changes.
This activity will not allow for a new fission product release path,
nor will it result in a new fission product barrier failure mode,
nor create a new sequence of events that would 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.
2. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The VEGP 3 and 4 emergency planning ITAAC provide assurance that
the facility has been constructed and will be operated in conformity
with the license, the provisions of the Act, and the Commissioner's
rules and regulations. The changes do not affect the assessments or
the plant itself. The changes do not adversely affect the safety-
related equipment or fission product barriers. No safety analysis or
design basis acceptance limit or criterion is challenged or exceeded
by the proposed change.
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: Mr. 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: May 19, 2017. A publicly-available
version is in ADAMS under Accession No. ML17139D394.
Description of amendment request: The requested amendment proposes
to depart from combined license (COL) Appendix C information (with
corresponding changes to the associated plant-specific Tier 1
information) and involves associated Tier 2 information in the Updated
Final Safety Analysis Report (UFSAR). Specifically, proposed changes
clarify that there is more than one turbine building main sump and adds
a second sump pump for each of the two turbine building main sumps into
the UFSAR Tier 2 and COL Appendix C (and associated plant-specific Tier
1) information.
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 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
[[Page 31102]]
consequences of an accident previously evaluated?
Response: No.
The activity adds a second pump to each of the turbine building
main sumps, and identifies that there is more than one turbine
building sump. The reason for the additional pumps is to account for
an increase in volume due to the changes to the condensate polishing
system (CPS) rinse effluent flowpath from CPS to waste water system
(WWS) via the turbine building sumps. The extra sump pumps will
prevent potential overflowing and flooding of the sumps during CPS
rinse operations. The CPS serves no safety-related function. By
directing the effluent to the turbine building sumps it is subject
to radiation monitoring. Under normal operating conditions, there
are is no significant amount of radioactive contamination within the
CPS. However, radioactive contamination of the CPS can occur as a
result of a primary-to-secondary leakage in the steam generator
should a steam generator tube leak develop while the CPS is in
operation and radioactive condensate is processed by the CPS.
Radiation monitors associated with the steam generator blowdown,
steam generator, and turbine island vents, drains and relief systems
provide the means to determine if the secondary side is
radioactively contaminated. The main turbine building sumps and sump
pumps are not safety-related components and do not interface with
any systems, structures, or components (SSC) accident initiator or
initiating sequence of events; thus, the probability of accidents
evaluated within the plant-specific UFSAR are not affected. The
proposed changes do not involve a change to the predicted
radioactive releases due to accident conditions, thus the
consequences of 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 to the nonsafety-related WWS do not affect
any safety-related equipment, nor do they add any new interface to
safety-related SSCs. No system or design function or equipment
qualification is affected by this change. The changes do not
introduce a new failure mode, malfunction, or sequence of events
that could affect safety or safety-related equipment. 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 WWS is a nonsafety-related system that does not interface
with any safety-related equipment. The proposed changes to identify
that there is more than one turbine building sump and to add two
turbine building sump pumps do not 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 change.
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.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: March 13, 2017. A publicly available
version is in ADAMS under Accession No. ML17073A018.
Description of amendment request: The amendments would modify the
Surveillance Requirement (SR) 3.8.1.17 of the Technical Specification
(TS) 3.8.1, ``AC [Alternating Current] Sources--Operating,'' to delete
the note to allow the performance of the SR in Modes 1 through 4 when
the associated load is out of service for maintenance or testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequence of an accident previously evaluated?
Response: No.
The proposal does not alter the function of any structure,
system or component functions, does not modify the manner in which
the plant is operated, and does not alter equipment out-of-service
time. This request does not degrade the ability of the emergency
diesel generator or equipment downstream of the load sequencers to
perform their intended function.
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 change does not involve any physical changes to
plant safety related structure, system or component or alter the
modes of plant operation in a manner that is outside the bounds of
the current emergency diesel generator system design analyses. The
proposed change to revise the note modifying SR 3.8.1.17 to allow
the performance of the SR in Modes 1 through 4 when the associated
equipment is out of service for maintenance or testing does not
create the possibility for an accident or malfunction of a different
type than any evaluated previously in SQN's Updated Final Safety
Analysis Report. The proposal does not alter the way any structure,
system or component function and does not modify the manner in which
the plant is operated. 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.
The proposed change to TS 3.8.1, ``AC Sources--Operating'' to
revise the note modifying SR 3.8.1.17 to allow the performance of
the SR in Modes 1 through 4 when the associated equipment is out of
service for maintenance or testing does not reduce the margin of
safety because the test methodologies are not being changed and LCO
[limiting condition for operation] allowed outage times are not
being changed. The results of accident analyses remain unchanged by
this request. 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Undine S. Shoop.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: March 31, 2017. A publicly available
version is in ADAMS under Accession No. ML17093A854.
Description of amendment request: The amendment would revise
Technical Specification (TS) 5.7.2.14, ``Ventilation Filter Testing
Program (VFTP),'' to delete references to the reactor building (RB)
purge filters. A previous amendment deleted the reactor building purge
air cleanup system from the TSs based on partial implementation of the
alternate source term methodology; however, references to the RB purge
filters were not removed from TS 5.7.2.14 at that time due to an
administrative oversight. The proposed change corrects the
administrative
[[Page 31103]]
oversight by deleting references to the RB purge filters in TS
5.7.2.14.
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 revision to WBN TS 5.7.2.1.14 is administrative in
nature. Nuclear Regulatory Commission (NRC) Amendment Number 92
(ML13141A564) deleted TS 3.9.8, ``Reactor Building Purge Air Cleanup
Units,'' based on implementation of the alternate source term (AST)
methodology because no credit is taken for the operation of reactor
building air cleanup units for the dose analysis during a fuel
handling accident (FHA). However, TVA neglected to remove the
references to the RB purge filters in TS 5.7.2.14. The proposed
change corrects this oversight by deleting the references to the RB
purge filters in TS 5.7.2.14a. through d.
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 would not require any new or different
accidents to be postulated and subsequently evaluated because no
changes are being made to the plant that would introduce any new
accident causal mechanisms. This license amendment request does not
impact any plant systems that are potential accident initiators, nor
does it have any significantly adverse impact on any accident
mitigating systems. No new or different accident scenarios,
transient precursors, failure mechanisms, or limiting single
failures will be introduced as a result of these changes.
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.
The proposed change does not alter the permanent plant design,
including instrument setpoints, nor does it change the assumptions
contained in the safety analyses. Margin of safety is related to the
ability of the fission product barriers to perform their design
functions during and following accident conditions. These barriers
include the fuel cladding, the reactor coolant system, and the
containment system. The performance of these barriers will not be
significantly degraded by the proposed changes.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Undine S. Shoop.
Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant,
Unit 2, Rhea County, Tennessee
Date of amendment request: March 28, 2017. A publicly-available
version is in ADAMS under Accession No. ML17093A608.
Description of amendment request: The amendment would revise the
Facility Operating License (OL) to extend the completion date for
Condition 2.C.(5) regarding the reporting of actions taken to resolve
issues identified in Nuclear Regulatory Commission Bulletin 2012-01,
``Design Vulnerability in Electric Power System,'' dated July 27, 2012
(ADAMS Accession No. ML12074A115).
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 to revise the completion date for OL
Condition 2.C(5) for WBN Unit 2 regarding the reporting of actions
taken to resolve issues identified in NRC Bulletin 2012-01 from
December 31, 2017 to December 31, 2018 do not affect the structures,
systems, or components (SSCs) of the plant, affect plant operations,
or any design function or any analysis that verifies the capability
of an SSC to perform a design function. No change is being made to
any of the previously evaluated accidents in the WBN Updated Final
Safety Analysis Report (UFSAR).
The proposed changes do not (1) require physical changes to
plant SSCs; (2) prevent the safety function of any safety-related
system, structure, or component during a design basis event; (3)
alter, degrade, or prevent action described or assumed in any
accident described in the WBN UFSAR from being performed because the
safety-related SSCs are not modified; (4) alter any assumptions
previously made in evaluating radiological consequences; or (5)
affect the integrity of any fission product barrier.
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 do not introduce any new accident causal
mechanisms, because no physical changes are being made to the plant,
nor do they affect any plant systems that are potential accident
initiators.
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.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed changes will have no effect
on the availability, operability, or performance of safety-related
systems and components. The proposed change will not adversely
affect the operation of plant equipment or the function of equipment
assumed in the accident analysis.
The proposed amendment does not involve changes to any safety
analyses assumptions, safety limits, or limiting safety system
settings. The changes do not adversely affect plant-operating
margins or the reliability of equipment credited in the safety
analyses.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Undine S. Shoop.
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
[[Page 31104]]
applicable, proposed no significant hazards consideration
determination, and opportunity for a hearing in connection with these
actions, was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
Duke Energy Progress, LLC, Docket Nos. 50-325 and 50-324, Brunswick
Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina
Date of amendment request: November 18, 2016.
Brief description of amendments: The amendments adopted the
approved Technical Specification Task Force (TSTF) Improved Standard
Technical Specifications Change Traveler TSTF-535, revising the
Technical Specification definition of Shutdown Margin (SDM) to require
calculation of the SDM at a reactor moderator temperature of 68 degrees
Fahrenheit, or a higher temperature that represents the most reactive
state throughout the operating cycle.
Date of issuance: June 7, 2017.
Effective date: As of date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 277 and 305. A publicly-available version is in
ADAMS under Accession No. ML17088A396; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: January 17, 2017 (82 FR
4929).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 7, 2017.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of amendment request: November 9, 2016.
Brief description of amendment: The amendment revises Technical
Specification (TS) 5.5.10, ``Ventilation Filter Testing Program,'' to
correct and modify the description of the control room ventilation and
fuel handling area ventilation systems. In addition, the amendment
corrects an editorial omission in TS Limiting Condition for Operation
3.0.9.
Date of issuance: June 8, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 263. A publicly-available version is in ADAMS under
Accession No. ML17121A510; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-20: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: February 14, 2017 (82
FR 10596).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 8, 2017.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: October 26, 2016.
Brief description of amendment: The amendment changed the Technical
Specifications (TS) to revise requirements for unavailable barriers by
adding new Limiting Condition for Operation (LCO) 3.0.9. This LCO
establishes conditions under which systems would remain operable when
required physical barriers are not capable of providing their related
support function. This amendment is consistent with NRC-approved
Technical Specification Task Force (TSTF) Improved Standard Technical
Specifications Change Traveler, TSTF-427, Revision 2, ``Allowance for
Non Technical Specification Barrier Degradation on Supported System
OPERABILITLY.'' The Notice of Availability of this TS improvement and
the model application was published in the Federal Register on October
3, 2006 (71 FR 58444), as part of the consolidated line item
improvement process.
Date of issuance: June 7, 2017.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No: 212. A publicly-available version is in ADAMS under
Accession No. ML17116A032; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-29: The amendment revised the
Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 20, 2016 (81
FR 92866).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 7, 2017.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: November 1, 2016.
Brief description of amendment: The amendment revised Technical
Specification (TS) 2.1.1, ``Reactor Core Safety Limits,'' to reduce the
reactor steam dome pressure value specified in TS 2.1.1.1 and TS
2.1.1.2 from 785 pounds per square inch gauge (psig) to 686 psig.
Date of issuance: June 19, 2017.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 176. A publicly-available version is in ADAMS under
Accession No. ML17139C372; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-58: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 20, 2016 (81
FR 92868).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 19, 2017.
No significant hazards consideration comments received: No.
[[Page 31105]]
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant (CNP), Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment request: October 18, 2016, as supplemented by
letter dated February 27, 2017.
Brief description of amendments: The amendments revised the CNP,
Unit Nos. 1 and 2, Technical Specification 5.5.14, ``Containment
Leakage Rate Testing Program,'' to clarify the containment leakage rate
testing pressure criteria.
Date of issuance: June 7, 2017.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 336 for Unit No. 1 and 318 for Unit No. 2. A
publicly-available version is in ADAMS under Accession No. ML17131A277;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-58 and DPR-74:
Amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: December 6, 2016 (81 FR
87972). The supplemental letter dated February 27, 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 amendments is contained
in a Safety Evaluation dated June 7, 2017.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota (NSPM), Docket No. 50-263,
Monticello Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: July 28, 2016.
Brief description of amendment: The amendment adopts TSTF-545,
Revision 3, ``TS [technical specification] Inservice Testing Program
Removal & Clarify SR [surveillance requirements] Usage Rule Application
to Section 5.5 Testing.''
Date of issuance: June 16, 2017.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 194. A publicly-available version is in ADAMS under
Accession No. ML17123A321; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-22: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: October 11, 2016 (81 FR
70181).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 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, Units 2 and 3, Fairfield County, South Carolina
Date of amendment request: October 9, 2015, as supplemented on
December 1, 2015, August 11, 2016, and December 21, 2016.
Description of amendment: This amendment revises License Condition
(LC) 2.D(12)(c)1. related to initial Emergency Action Levels (EALs).
The LC will require the licensee 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: April 10, 2017.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment Nos.: 68 (Unit 2) and 68 (Unit 3). A publicly-available
version is in ADAMS under Accession Package No. ML16214A135; documents
related to this amendment are listed in the Safety Evaluation enclosed
with the amendment.
Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendment
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: January 19, 2016 (81 FR
2919). The supplemental letters dated December 1, 2015, August 11,
2016, and December 21, 2016, provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original 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 April 10, 2017.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield,
South Carolina
Date of amendment request: January 20, 2017, and supplemented by
letter dated March 8, 2017.
Description of amendment: The amendment consists of changes to the
VCSNS Units 2 and 3 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 31, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 74. A publicly-available version is in ADAMS under
Accession No. ML17130A903; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendment
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: February 28, 2017 (82
FR 12130). The supplemental letter dated March 8, 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 31, 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: February 15, 2016, as supplemented by
letters dated August 19, 2016, August 26, 2016, September 13, 2016,
December 16, 2016, and March 17, 2017.
Description of amendment: The amendment authorizes 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 and involves related changes to the
associated plant-specific Tier 2* information. Specifically, the
departures
[[Page 31106]]
consist of changes to UFSAR text and tables, and information
incorporated by reference into the UFSAR related to updates to WCAP-
16096, ``Software Program Manual for Common Q\TM\ Systems,'' and WCAP-
16097, ``Common Qualified Platform Topical Report.''
Date of issuance: June 8, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 79 (Unit 3) and 78 (Unit 4). A publicly-available
version is in ADAMS under Accession No. ML17104A109; documents related
to this amendment are listed in the Safety Evaluation enclosed with the
amendment.
Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment
revised the Facility Combined License.
Date of initial notice in Federal Register: April 12, 2016 (81 FR
21602). The supplemental letters dated August 19, 2016, August 26,
2016, September 13, 2016, December 16, 2016, and March 17, 2017,
provided additional information that clarified the application, did not
expand the scope of the application request as noticed on February 15,
2016, and did not change the staff's proposed no significant hazards
consideration determination as published in the Federal Register on
April 12, 2016.
The Commission's related evaluation of the amendment is contained
in the Safety Evaluation dated June 8, 2017.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 23rd day of June 2017.
For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2017-13804 Filed 7-3-17; 8:45 am]
BILLING CODE 7590-01-P