[Federal Register Volume 82, Number 58 (Tuesday, March 28, 2017)]
[Notices]
[Pages 15377-15392]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-05990]
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NUCLEAR REGULATORY COMMISSION
[NRC-2017-0080]
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 February 28, 2017 to March 13, 2017. The
last biweekly notice was published on March 14, 2017.
DATES: Comments must be filed by April 27, 2017. A request for a
hearing must be filed by May 30, 2017.
[[Page 15378]]
ADDRESSES: You may submit comments by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0080. 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: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Paula Blechman, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-2242; email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2017-0080, 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-2017-0080.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in the
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2017-0080, 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 posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment submissions into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, 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
[[Page 15379]]
petitioner seeks to have litigated in the proceeding. Each contention
must consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner must provide a
brief explanation of the bases for the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to the specific sources and documents on which the petitioner intends
to rely to support its position on the issue. The petition must include
sufficient information to show that a genuine dispute exists with the
applicant or licensee on a material issue of law or fact. Contentions
must be limited to matters within the scope of the proceeding. The
contention must be one which, if proven, would entitle the petitioner
to relief. A petitioner who fails to satisfy the requirements at 10 CFR
2.309(f) with respect to at least one contention will not be permitted
to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of the amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission by May
30, 2017. The petition must be filed in accordance with the filing
instructions in the ``Electronic Submissions (E-Filing)'' section of
this document, and should meet the requirements for petitions set forth
in this section, except that under 10 CFR 2.309(h)(2) a State, local
governmental body, or federally recognized Indian Tribe, or agency
thereof does not need to address the standing requirements in 10 CFR
2.309(d) if the facility is located within its boundaries.
Alternatively, a State, local governmental body, Federally-recognized
Indian Tribe, or agency thereof may participate as a non-party under 10
CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
the discretion of the presiding officer, be permitted to make a limited
appearance pursuant to the provisions of 10 CFR 2.315(a). A person
making a limited appearance may make an oral or written statement of
his or her position on the issues but may not otherwise participate in
the proceeding. A limited appearance may be made at any session of the
hearing or at any prehearing conference, subject to the limits and
conditions as may be imposed by the presiding officer. Details
regarding the opportunity to make a limited appearance will be provided
by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for leave to intervene (petition), any
motion or other document filed in the proceeding prior to the
submission of a request for hearing or petition to intervene, and
documents filed by interested governmental entities that request to
participate under 10 CFR 2.315(c), must be filed in accordance with the
NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR
46562, August 3, 2012). The E-Filing process requires participants to
submit and serve all adjudicatory documents over the internet, or in
some cases to mail copies on electronic storage media. Detailed
guidance on making electronic submissions may be found in the Guidance
for Electronic Submissions to the NRC and on the NRC's 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
[[Page 15380]]
have advised the Office of the Secretary that they wish to participate
in the proceeding, so that the filer need not serve the document on
those participants separately. Therefore, applicants and other
participants (or their counsel or representative) must apply for and
receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public Web site at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m.
and 6 p.m., Eastern Time, Monday through Friday, excluding government
holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
stating why there is good cause for not filing electronically and
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff. Participants filing adjudicatory documents in this
manner are responsible for serving the document on all other
participants. Filing is considered complete by first-class mail as of
the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service. A presiding officer, having granted an
exemption request from using E-Filing, may require a participant or
party to use E-Filing if the presiding officer subsequently determines
that the reason for granting the exemption from use of E-Filing no
longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the
Commission or the presiding officer. If you do not have an NRC-issued
digital ID certificate as described above, click cancel when the link
requests certificates and you will be automatically directed to the
NRC's electronic hearing dockets where you will be able to access any
publicly available documents in a particular hearing docket.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or personal phone
numbers in their filings, unless an NRC regulation or other law
requires submission of such information. For example, in some
instances, individuals provide home addresses in order to demonstrate
proximity to a facility or site. With respect to copyrighted works,
except for limited excerpts that serve the purpose of the adjudicatory
filings and would constitute a Fair Use application, participants are
requested not to include copyrighted materials in their submission.
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station (PNPS), Plymouth County, Massachusetts
Date of amendment request: February 14, 2017. A publicly available
version is in ADAMS under Accession No. ML17053A468.
Description of amendment request: The amendment would revise
certain staffing and training requirements, reports, programs, and
editorial changes in the Technical Specifications (TS) Table of
Contents; Section 1.0, ``Definitions''; Section 4.0, ``Design
Features''; and Section 5.0, ``Administrative Controls'' that will no
longer be applicable once PNPS is permanently defueled.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would not take effect until PNPS has
permanently ceased operation and entered a permanently defueled
condition and the Certified Fuel Handler Training and Retraining
Program is approved by the NRC. The proposed amendment would modify
the PNPS TS by deleting the portions of the TS that are no longer
applicable to a permanently defueled facility, while modifying the
other sections to correspond to the permanently defueled condition.
The deletion and modification of provisions of the
administrative controls do not directly affect the design of
structures, systems, and components (SSCs) necessary for safe
storage of irradiated fuel or the methods used for handling and
storage of such fuel in the spent fuel pool. The changes to the
administrative controls are administrative in nature and do not
affect any accidents applicable to the safe management of irradiated
fuel or the permanently shutdown and defueled condition of the
reactor. Thus, the consequences of an accident previously evaluated
are not increased.
In a permanently defueled condition, the only credible accidents
are the fuel handling accident (FHA) and those involving radioactive
waste systems remaining in service. The probability of occurrence of
previously evaluated accidents is not increased, because extended
operation in a defueled condition will be the only operation
allowed. This mode of operation is bounded by the existing analyses.
Additionally, the occurrence of postulated accidents associated with
reactor operation is no longer credible in a permanently defueled
reactor. This significantly reduces the scope of applicable
accidents.
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 have no impact on facility SSCs affecting
the safe storage of irradiated fuel, or on the methods of operation
of such SSCs, or on the handling and storage of irradiated fuel
itself. The administrative removal or modifications of the TS that
are related only to administration of the facility cannot result in
different or more adverse failure modes or accidents than previously
evaluated because the reactor will be permanently shutdown and
defueled and PNPS will no longer be authorized to operate the
reactor or retain or place fuel in the reactor vessel.
The proposed changes to the PNPS TS do not affect systems
credited in the accident analysis for the FHA or radioactive waste
system upsets at PNPS. The proposed TS will continue to require
proper control and monitoring of safety significant parameters and
activities.
The proposed amendment does not result in any new mechanisms
that could initiate damage to the remaining relevant safety barriers
for defueled plants (fuel cladding and spent fuel cooling). Extended
operation in a defueled condition will be the only
[[Page 15381]]
operation allowed, and it is bounded by the existing analyses, such
a condition does not create the possibility of a new or different
kind of accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Since the 10 CFR part 50 license for PNPS will no longer
authorize operation of the reactor or emplacement or retention of
fuel into the reactor vessel once the certifications required by 10
CFR 50.82(a)(1) are docketed, as specified in 10 CFR 50.82(a)(2),
the occurrence of postulated accidents associated with reactor
operation is no longer credible. The only remaining credible
accidents are a FHA and those involving radioactive waste systems
remaining in service. The proposed amendment does not adversely
affect the inputs or assumptions of any of the design basis analyses
that impact these analyzed conditions.
The proposed changes are limited to those portions of the TS
that are not related to the safe storage of irradiated fuel. The
requirements that are proposed to be revised or deleted from the
PNPS TS are not credited in the existing accident analysis for the
remaining applicable postulated accident; and as such, do not
contribute to the margin of safety associated with the accident
analysis. Postulated design basis accidents involving the reactor
are no longer possible because the reactor will be permanently
shutdown and defueled and PNPS will no longer be authorized to
operate the reactor or retain or place fuel in the reactor vessel.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2 (NMP2), Oswego County, New York
Date of amendment request: December 13, 2016, as supplemented by
letter dated February 17, 2017. Publicly-available versions are in
ADAMS under Accession Nos. ML16348A368 and ML17048A034, respectively.
Description of amendment request: The amendment would revise the
NMP2 technical specification (TS) safety limit (SL) to increase the low
pressure isolation setpoint allowable value, which will result in
earlier main steam line isolation. The revised main steam line low
pressure isolation capability and the revised SL are intended to ensure
that NMP2 remains within the TS SLs in the event of a pressure
regulator failure maximum demand transient.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below, with NRC staff edits in square
brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because decreasing the reactor dome pressure in TS SL 2.1.1.1 and TS
SL 2.1.1.2 for reactor RTP [rated thermal power] ranges and
increasing the AV [allowable value] for the Main Steam Line
Pressure-Low on TS Table 3.3.6.1-1, Function b, effectively expands
the range of applicability for GEXL correlation and the calculation
of MCPR [minimum critical power ratio]. The CPR [critical power
ratio] rises during the pressure reduction following the scram that
terminates the PRFO [pressure regulator failure--maximum demand
(open)] transient. The reduction in the reactor dome pressure value
in the SL from 785 psig [pounds per square inch gauge] to 700 psia
[pounds per square inch absolute] and the increase in the AV from
>=746 psig to >=814 psig adequately accommodate the pressure
reduction during the PRFO transient within the revised TS limit
without compromising fuel integrity.
The expanded GEXL correlation range supports NMP2 revised low
pressure safety limit of 700 psia. The proposed TS revision involves
no significant changes to the operation of any systems or components
in normal or accident or transient operating 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 amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the proposed reduction in the reactor dome pressure value in
the SL from 785 psig to 700 psia reflects a wider range of
applicability for the GEXL correlation which is approved by the NRC
for both GE14 currently in NMP2 and GNF2 fuels proposed for NMP2.
The proposed changes do not involve physical changes to the plant or
its operating characteristics. In addition, the increase in the AV
for the MSL [main steam line] low pressure from >=746 psig to >=814
psig will result in the MSIV [main steam isolation valve] closure
signal initiation at a higher temperature. As a result, no new
failure modes are being introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not involve a significant reduction in a
margin of safety because the margin of safety is established through
the design of the plant structures, systems, and components, and
through the parameters for safe operation and setpoints for the
actuation of equipment relied upon to respond to transients and
design basis accidents. The proposed change in reactor dome pressure
SLs and the AV for the MSL low pressure ensures the safety margin is
maintained, which protects the fuel cladding integrity during steady
state operation, normal operational transients, or AOOs [anticipated
operational occurrences] such as a depressurization transient, but
does not change the requirements governing operation or availability
of safety equipment assumed to operate to preserve the margin of
safety. The proposed changes do not involve physical changes to the
plant or its operating characteristics. The reduction in the reactor
dome pressure value in the SL from 785 psig to 700 psia and the
increase to the AV for the MSL low pressure provides added margin to
accommodate the pressure reduction during the PRFO transient within
the revised TS limit without compromising fuel integrity.
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 Acting Branch Chief: Stephen S. Koenick.
Exelon Generation Company, LLC (Exelon), Docket No. 50-219, Oyster
Creek Nuclear Generating Station (OCNGS), Ocean County, New Jersey
Date of amendment request: February 20, 2017. A publicly-available
version is available in ADAMS under Accession No. ML17051A003.
Description of amendment request: The licensee proposes to delete
from the Facility Operating License (FOL) certain license conditions,
which impose specific requirements on the decommissioning trust
agreement. The
[[Page 15382]]
licensee proposes to meet the provisions of 10 CFR 50.75(h) for OCNGS.
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 requested changes delete License Conditions 3.F through 3.K
pertaining to Decommissioning Trust Agreements currently in the
OCNGS FOL. The requested changes are consistent with the types of
license amendments [identified] in 10 CFR 50.75(h)(4).
The regulations of 10 CFR 50.75(h)(4) state ``Unless otherwise
determined by the Commission with regard to a specific application,
the Commission has determined that any amendment to the license of a
utilization facility that does no more than delete specific license
conditions relating to the terms and conditions of decommissioning
trust agreements involves ``no significant hazard considerations.''
This request involves changes that are administrative in nature.
No actual plant equipment or accident analyses will be affected by
the proposed changes.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Does the [p]roposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This request involves administrative changes to the license that
will be consistent with the NRC's regulations at 10 CFR 50.75(h).
No actual plant equipment or accident analyses will be affected
by the proposed change and no failure modes not bounded by
previously evaluated accidents will be created.
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 confidence in the ability of
the fission product barriers to limit the level of radiation dose to
the public.
This request involves administrative changes to the license that
will be consistent with the NRC's regulations at 10 CFR 50.75(h).
No actual plant equipment or accident analyses will be affected
by the proposed change. Additionally, the proposed changes will not
relax any criteria used to establish safety limits, will not relax
any safety systems settings, or will not relax the bases for any
limiting conditions of operation.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277
and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and
Lancaster Counties, Pennsylvania
Date of amendment request: January 30, 2017. A publicly-available
version is in ADAMS under Accession No. ML17030A302.
Description of amendment request: The amendments would replace
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 RPV water level to be greater than the top of active
irradiated fuel. The proposed changes are based on TS Task Force (TSTF)
Traveler TSTF-542, Revision 2, ``Reactor Pressure Vessel Water
Inventory 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 replace 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 changes reduce 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 changes reduce 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 subsystems 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 containment and/or
filtration would be available if needed.
The proposed changes reduce or eliminate 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 changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes replace existing TS requirements related to
OPDRVs with new requirements on RPV WIC that will protect Safety
Limit 2.1.1.3. The proposed changes will not alter the design
function of the equipment involved. Under the proposed changes, 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
[[Page 15383]]
requirements. The event of concern under the current requirements
and the proposed changes are an unexpected draining event. The
proposed changes do 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 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 replace 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 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 Rd., Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
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: January 23, 2017. A publicly-available
version is in ADAMS under Accession No. ML17025A399.
Description of amendment request: The amendments would modify the
St. Lucie Plant, Unit Nos. 1 and 2, Technical Specifications (TSs) by
limiting the MODE of applicability for the Reactor Protection System
(RPS), Startup, and Operating Rate of Change of Power--High, functional
unit trip. Additionally, the proposed license amendments add new
Limiting Condition for Operation (LCO) 3.0.5 and relatedly modifies LCO
3.0.2, to provide for placing inoperable equipment under administrative
control for the purpose of conducting testing required to demonstrate
OPERABILITY.
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.
Limiting the MODE 1 applicability for RPS functional unit,
Startup and Operating Rate of Change of Power--High, to Power Range
Neutron Flux Power <=15% of RATED THERMAL POWER, is an
administrative change in nature and does not alter the manner in
which the functional unit is operated or maintained. The proposed
changes do not represent any physical change to plant [structures,
systems, and components (SSC(s))], or to procedures established for
plant operation. The subject RPS functional unit is not an event
initiator nor is it credited in the mitigation of any event or
credited in the [probabilistic risk assessment (PRA)]. As such, the
initial conditions associated with accidents previously evaluated
and plant systems credited for mitigating the consequences of
accidents previously evaluated remain unchanged.
The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1
and Unit 2 TS and related modification to LCO 3.0.2 is consistent
with the guidance provided in NUREG-1432, Volume 1 [ADAMS Accession
No. ML12102A165] (Reference 6.1 [of the amendment request]) and
thereby has been previously evaluated by the Commission with a
determination that the proposed change does not involve a
significant hazards consideration.
Therefore, facility operation in accordance with the proposed
license amendments would 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.
Limiting the MODE 1 applicability for the RPS functional unit,
Startup and Operating Rate of Change of Power--High, to Power Range
Neutron Flux Power <= 5% of RATED THERMAL POWER, is an
administrative change in nature and does not involve the addition of
any plant equipment, methodology or analyses. The proposed changes
do not alter the design, configuration, or method of operation of
the subject RPS functional unit or of any other SSC. More
specifically, the proposed changes neither alter the power rate-of-
change trip function nor its ability to bypass and reset as
required. The subject RPS functional unit remains capable of
performing its design function.
The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1
and Unit 2 TS and related modification to LCO 3.0.2 is consistent
with the guidance provided in NUREG-1432, Volume 1 (Reference 6.1
[of the amendment request]) and thereby has been previously
evaluated by the Commission with a determination that the proposed
change does not involve a significant hazards consideration.
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.
Limiting the MODE 1 applicability for RPS functional unit,
Startup and Operating Rate of Change of Power--High, to Power Range
Neutron Flux Power <=15% of RATED THERMAL POWER is an administrative
change in nature. The proposed changes neither involve changes to
any safety analyses assumptions, safety limits, or limiting safety
system settings nor do they adversely impact plant operating margins
or the reliability of equipment credited in safety analyses.
The proposed addition of new LCO 3.0.5 to the St. Lucie Unit 1
and Unit 2 TS and related modification to LCO 3.0.2 is consistent
with the guidance provided in NUREG-1432, Volume 1 (Reference 6.1
[of the amendment request]) and thereby has been previously
evaluated by the Commission with a determination that the proposed
change does not involve a significant hazards consideration.
Therefore, operation of the facility in accordance with the
proposed amendment will not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Benjamin G. Beasley.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1 (FCS), Washington County, Nebraska
Date of amendment request: December 16, 2016. A publicly-available
version is in ADAMS under Accession No. ML16351A464.
[[Page 15384]]
Description of amendment request: The proposed amendment would
revise the FCS Emergency Plan and Emergency Action Level (EAL) scheme
for the permanently defueled condition. The proposed permanently
defueled Emergency Plan and EAL scheme are commensurate with the
significantly reduced spectrum of credible accidents that can occur in
the permanently defueled condition and are necessary to properly
reflect the conditions of the facility while continuing to preserve the
effectiveness of the emergency plan.
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 the FCS Emergency Plan and EAL scheme do
not impact the function of facility structures, systems, or
components. The proposed changes do not affect accident initiators
or precursors, nor does it alter design assumptions. The proposed
changes do not prevent the ability of the on-shift staff and
emergency response organization to perform their intended functions
to mitigate the consequences of any accident or event that will be
credible in the permanently defueled condition.
The probability of occurrence of previously evaluated accidents
is not increased, because most previously analyzed accidents can no
longer occur and the probability of the few remaining credible
accidents are unaffected by the proposed amendment.
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 reduce the scope of the FCS Emergency Plan
and EAL scheme commensurate with the hazards associated with a
permanently shutdown and defueled facility. The proposed changes do
not involve installation of new equipment or modification of
existing equipment, so that no new equipment failure modes are
introduced. Also, the proposed changes do not result in a change to
the way that the equipment or facility is operated resulting in new
or different kinds of accident initiators or accident mitigation.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed changes are
associated with the FCS Emergency Plan and EAL scheme and do not
impact operation of the facility or its response to transients or
accidents. The change does not affect the Technical Specifications.
The proposed changes do not involve a change in the method of
facility operation, and no accident analyses will be affected by the
proposed changes. Safety analysis acceptance criteria are not
affected by the proposed changes. The revised Emergency Plan will
continue to provide the necessary response staff.
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 A. Repka, Esq., Winston & Strawn, 1700
K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: Douglas A. Broaddus.
PSEG Nuclear LLC, Docket Nos. 50-354, 50-272, and 50-311, Hope Creek
Generating Station (HCGS) and Salem Nuclear Generating Station (SGS),
Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: February 13, 2017. A publicly-available
version is in ADAMS under Package Accession No. ML17044A346.
Description of amendment request: The amendments would revise the
HCGS and SGS, Unit Nos. 1 and 2, emergency action level (EAL) schemes.
Specifically, the licensee proposes to adopt the EAL scheme described
in Nuclear Energy Institute (NEI) 99-01, Revision 6, ``Development of
Emergency Action Levels for Non-Passive Reactors.'' NEI 99-01, Revision
6, has been endorsed by 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the HCGS and SGS EALs do not impact the
physical function of plant structures, systems or components (SSC)
or the manner in which SSCs perform their design function. 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 change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve a physical alteration of the
plant (i.e., no new or different types of equipment will be
installed or removed) or a change in the method of plant operation.
The proposed changes will not introduce failure modes that could
result in a new accident, and the changes do not alter assumptions
made in the safety analysis. The proposed changes to the HCGS and
SGS EALs are not initiators of any accidents. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
3. Does the proposed change 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 do not impact
operation of the plant or its response to transients 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 a 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.
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.
[[Page 15385]]
Attorney for licensee: Jeffrie J. Keenan, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: James G. Danna.
South Carolina Electric & Gas Company, Docket Nos. 52-027 and 52-028,
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield, South
Carolina
Date of amendment request: February 15, 2017. A publicly-available
version is in ADAMS under Accession No. ML17046A660.
Description of amendment request: The amendment request proposes to
revise the licensing basis information to reflect changes to the
locations of the hydrogen venting primary openings in the passive core
cooling system (PXS) valve/accumulator rooms inside containment.
Because this proposed change requires a departure from Tier 1
information in the Westinghouse Electric Company's AP1000 Design
Control Document (DCD), the licensee also requested an exemption from
the requirements of the Generic DCD Tier 1 in accordance with 10 CFR
52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed revision to the hydrogen venting for the Passive
Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and
clarification of the venting path definition for PXS Valve/
Accumulator Room B (Room 11207) do not affect any safety-related
equipment or function. The hydrogen ignition subsystem, including
designed hydrogen venting features, is designed to mitigate beyond
design basis hydrogen generation in the containment. The hydrogen
venting changes do not involve any accident, initiating event or
component failure; thus, the probabilities of the accidents
previously evaluated are not affected. The modified venting
locations and definitions will maintain the hydrogen ignition
subsystem designed and analyzed beyond design basis function to
maintain containment integrity. The maximum allowable containment
leakage rate specified in the Technical Specifications is unchanged,
and radiological material release source terms are not affected;
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 revision to the hydrogen venting for the Passive
Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and
clarification of the venting path definition for PXS Valve/
Accumulator Room B (Room 11207) will maintain the beyond design
basis function of the hydrogen ignition subsystem. The hydrogen
venting changes do not impact the hydrogen ignition subsystem's
function to maintain containment integrity during beyond design
basis accident conditions, and, thus does not introduce any new
failure mode. The proposed changes do not create a new fault or
sequence of events that could result in a radioactive release. The
proposed changes would not affect any safety-related accident
mitigating function.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed revision to the hydrogen venting for the Passive
Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and
clarification of the venting path definition for PXS Valve/
Accumulator Room B (Room 11207) will maintain the beyond design
basis function of the hydrogen ignition subsystem. The proposed
changes do not have any effect on the ability of safety-related
structures, systems, or components to perform their beyond design
basis functions. The proposed changes are a result of a low
probability, severe accident scenario being evaluated. The revision
to this scenario does not result in an increase in the plant risk
(frequency and/or consequences). The frequency is low and there is
no increase to the consequences because containment integrity is
maintained and there is no containment leakage. There is no change
to the maximum allowed containment leakage rate (0.10% of
containment air weight per day) for the containment vessel. The
proposed changes do not affect the ability of the hydrogen igniter
subsystem to maintain containment integrity following a beyond
design basis accident. The hydrogen igniter subsystem continues to
meet the requirements for which it was designed and continues to
meet the regulations.
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 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: February 16, 2017. A publicly-available
version is in ADAMS under Accession No. ML17047A192.
Description of amendment request: The requested amendment proposes
to depart from Tier 2 information in the Updated Final Safety Analysis
Report (UFSAR) and involves changes to related plant-specific Tier 1
information, with corresponding changes to the associated combined
license (COL) Appendix C information, to clarify text that currently
refers to raceways with an electrical classification (i.e., Class 1E/
non-Class 1E). This includes rewording multiple Inspections, Tests,
Analyses, and Acceptance Criteria (ITAAC) and UFSAR material to clarify
that any text referring to Class 1E or non-Class 1E raceways or raceway
systems is referring to raceways or raceway systems that route Class 1E
or non-Class 1E circuits.
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.
These proposed changes are for clarification and consistency. No
structure, system, or component (SSC) or function is changed within
this activity. There is no change to the application of regulatory
guides or industry standards to raceways or raceway systems, nor is
there a change to how they are designed, fabricated, procured or
installed. Raceway systems that route Class 1E circuits will
continue to be designated and designed as equipment Class C, safety-
related, and seismic Category I structures. The proposal to align
the text in COL Appendix C (and plant-specific Tier 1) Section 3.3
with the associated ITAAC is made for clarification and consistency
to reduce misinterpretation. The proposal to reword multiple ITAAC
in 3.3.00.07 does not change the intent of the ITAAC, nor is the
ITAAC scope or closure method impacted.
The proposed amendment does not affect the prevention and
mitigation of abnormal events; e.g., accidents, anticipated
operation occurrences, earthquakes, floods, turbine missiles, and
fires or their safety or design analyses. This change does not
involve containment of radioactive isotopes or any adverse effect on
a fission product barrier. There is no impact on previously
evaluated accidents.
Therefore, the proposed amendment does not involve a significant
increase in the
[[Page 15386]]
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 involve a new failure mechanism or
malfunction, which affects an SSC accident initiator, or interface
with any SSC accident initiator or initiating sequence of events
considered in the design and licensing bases. There is no adverse
effect on radioisotope barriers or the release of radioactive
materials. The proposed amendment does not adversely affect any
accident, including the possibility of creating a new or different
kind of accident from any accident previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different type of accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
These proposed changes are for clarification and consistency to
reduce misinterpretation. No SSC or function is changed within this
activity. There is no change to the application of regulatory guides
or industry standards to raceways or raceway systems, nor is there a
change to how they are designed, fabricated, procured or installed.
Raceway systems that route Class 1E circuits will continue to be
designated and designed as Equipment Class C, safety-related, and
seismic Category I.
The proposed changes would not affect any safety-related design
code, function, design analysis, safety analysis input or result, or
existing design/safety margin. No safety analysis or design basis
acceptance limit/criterion is challenged or exceeded by the
requested changes.
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 Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke
County, Georgia
Date of amendment request: August 30, 2016. A publicly-available
version is in ADAMS under Accession No. ML16243A373.
Description of amendment request: The amendment request proposes a
change to Updated Final Safety Analysis Report in the form of
departures from the incorporated plant-specific Design Control Document
(DCD) Tier 2 * information and related changes to the VEGP Units 3 and
4 Combined License (COL) Appendix C (and corresponding plant-specific
DCD 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,
a design certification rule is also requested for the plant-specific
Tier 1 material departures. The proposed change is to the thickness of
one floor in the auxiliary building located between Column Lines I to
J-1 and Column Lines 2 to 4 at Elevation 153'-0''. This submittal
requests approval of the license amendment, necessary to implement
these changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with NRC staff edits in square
brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design functions of the nuclear island structures are to
provide support, protection, and separation for the seismic Category
I mechanical and electrical equipment located in the nuclear island.
The nuclear island structures are structurally designed to meet
seismic Category I requirements as defined in Regulatory Guide 1.29.
The change of the thickness of the floor above the [Component
Cooling Water System (CCS)] Valve Room in the auxiliary building
meets criteria and requirements of American Concrete Institute (ACI)
349 and American Institute of Steel Construction (AISC) N690 and
does not have an adverse impact on the response of the nuclear
island structures safe shutdown earthquake ground motions or loads
due to anticipated transient or postulated accident conditions. The
proposed changes do not impact the support, design, or operation of
mechanical and fluid systems. There is no change to plant systems or
the response of systems to postulated accident conditions. There is
no change to the predicted radioactive releases due to normal
operation or postulated accident conditions. The plant response to
previously evaluated accidents or external events is not adversely
affected, nor does the change described create any new accident
precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change is to revise the thickness of the floor
above the CCS Valve Room in the auxiliary building. The proposed
changes do not change the design requirements of the nuclear island
structures. The proposed changes do not change the design function,
support, design, or operation of mechanical and fluid systems. The
proposed changes do not result in a new failure mechanism for the
nuclear island structures or new accident precursors. As a result,
the design function of the nuclear island structures is not
adversely affected by the proposed change.
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.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed changes, thus, no margin of
safety is reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County,
Georgia
Date of amendment request: January 31, 2017. A publicly-available
version is in ADAMS under Accession No. ML17031A446.
Description of amendment request: The requested amendment proposes
to depart from Tier 2 information in the Updated Final Safety Analysis
Report (UFSAR) and to change Combined License Appendix A, Technical
Specifications (TS), to modify engineered safety features logic for
containment vacuum relief actuation.
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?
[[Page 15387]]
Response: No.
The proposed changes to the UFSAR and TS will include the
Containment Pressure--Low automatic reset function for the
containment vacuum relief valves manual initiation logic, such that
the containment vacuum relief manual actuation will be automatically
reset when the containment pressure rises above the Containment
Pressure--Low setpoint. This reset allows a containment isolation
signal to close the valves when necessary. The Containment
Pressure--Low signal is an interlock for the containment vacuum
relief manual actuation such that the valves cannot be opened unless
the Containment Pressure--Low setpoint has been reached in any two-
out-of-four divisions. The modified logic will ensure that the
automatic initiation of containment isolation is made available
following manual initiation of containment vacuum relief actuation.
The analyzed design and function of the Engineered Safety Features
Actuation System and its actuated components is not affected. The
proposed changes do not adversely affect any safety-related
equipment and does not involve any accident, initiating event, or
component failure, thus the probabilities of accidents previously
evaluated are not affected. The proposed changes do not adversely
interface with or adversely affect any system containing
radioactivity or affect any radiological material release source
term; thus the radiological releases in an accident 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 changes to the UFSAR and TS to include the Containment
Pressure--Low manual actuation interlock and automatic reset
function for the containment vacuum relief valves manual initiation
logic will maintain the Engineered Safety Features Actuation System
and Plant Safety and Monitoring System in accordance with the design
objectives as licensed. The design of the Class 1E Containment
Pressure--Low manual actuation interlock and automatic reset
function is required to meet the licensing basis for the Engineered
Safety Features Actuation System and Plant Safety and Monitoring
System. The changes to the manual initiation logic do not adversely
affect the function of any safety-related structure, system, or
component, and thus does not introduce a new failure mode. The
changes to the containment vacuum relief valves manual initiation
logic do not adversely interface with any safety-related equipment
or any equipment associated with radioactive material and, thus, do
not create a new fault or sequence of events that could result in a
new or different kind of accident.
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 changes to the UFSAR and TS to include the Containment
Pressure--Low automatic reset function for the containment vacuum
relief valves manual initiation logic will maintain the Engineered
Safety Features Actuation System and Plant Safety and Monitoring
System in accordance with the design objectives as licensed. The
changes to the manual initiation logic do not adversely interface
with any safety-related equipment or adversely affect any safety-
related function. The changes to the containment vacuum relief
manual initiation logic continue to comply with existing design
codes and regulatory criteria, and do not involve a significant
reduction in the margin of safety.
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: March 2, 2017. A publicly-available
version is in ADAMS under Accession No. ML17061A747.
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 material departures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed 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: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
[[Page 15388]]
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: February 22, 2017. A publicly-available
version is in ADAMS under Accession No. ML17053A425.
Description of amendment request: The amendment request proposes to
revise the licensing basis information to reflect changes to the
locations of the hydrogen venting primary openings in the passive core
cooling system (PXS) valve/accumulator rooms inside containment.
Because, this proposed change requires a departure from Tier 1
information in the Westinghouse Electric Company's AP1000 Design
Control Document (DCD), the licensee also requested an exemption from
the requirements of the Generic DCD Tier 1 in accordance with 10 CFR
52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed revision to the hydrogen venting for the Passive
Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and
clarification of the venting path definition for PXS Valve/
Accumulator Room B (Room 11207) do not affect any safety-related
equipment or function. The hydrogen ignition subsystem, including
designed hydrogen venting features, is designed to mitigate beyond
design basis hydrogen generation in the containment. The hydrogen
venting changes do not involve any accident, initiating event or
component failure; thus, the probabilities of the accidents
previously evaluated are not affected. The modified venting
locations and definitions will maintain the hydrogen ignition
subsystem designed and analyzed beyond design basis function to
maintain containment integrity. The maximum allowable containment
leakage rate specified in the Technical Specifications is unchanged,
and radiological material release source terms are not affected;
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 revision to the hydrogen venting for the PXS Valve/
Accumulator Room A (Room 11206) and clarification of the venting
path definition for PXS Valve/Accumulator Room B (Room 11207) will
maintain the beyond design basis function of the hydrogen ignition
subsystem. The hydrogen venting changes do not impact the hydrogen
ignition subsystem's function to maintain containment integrity
during beyond design basis accident conditions, and, thus does not
introduce any new failure mode. The proposed changes do not create a
new fault or sequence of events that could result in a radioactive
release. The proposed changes would not affect any safety-related
accident mitigating function.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed revision to the hydrogen venting for the Passive
Core Cooling System (PXS) Valve/Accumulator Room A (Room 11206) and
clarification of the venting path definition for PXS Valve/
Accumulator Room B (Room 11207) will maintain the beyond design
basis function of the hydrogen ignition subsystem. The proposed
changes do not have any effect on the ability of safety-related
structures, systems, or components to perform their beyond design
basis functions. The proposed changes are a result of a low
probability, severe accident scenario being evaluated. The revision
to this scenario does not result in an increase in the plant risk
(frequency and/or consequences). The frequency is low and there is
no increase to the consequences because containment integrity is
maintained and there is no containment leakage. There is no change
to the maximum allowed containment leakage rate (0.10% of
containment air weight per day) for the containment vessel. The
proposed changes do not affect the ability of the hydrogen igniter
subsystem to maintain containment integrity following a beyond
design basis accident. The hydrogen igniter subsystem continues to
meet the requirements for which it was designed and continues to
meet the regulations.
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.
Tennessee Valley Authority (TVA), Docket No. 50-391, Watts Bar Nuclear
Plant (WBN), Unit 2, Rhea County, Tennessee
Date of amendment request: December 21, 2016. A publicly-available
version is in ADAMS under Accession No. ML16356A673.
Description of amendment request: The amendment would revise the
containment ice mass limits in WBN, Unit 2, Technical Specification
(TS) Surveillance Requirements (SRs) 3.6.11.2 and 3.6.11.3 to be
identical to the ice mass limits in the WBN, Unit 1, TS SRs 3.6.11.2
and 3.6.11.3.
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 primary purpose of the ice bed is to provide a large heat
sink to limit peak containment pressure in the event of a release of
energy from a design basis LOCA [loss-of-coolant accident] or high
energy line break (HELB) in containment. The LOCA requires the
greatest amount of ice compared to other accident scenarios;
therefore, the reduction in ice weight is based on the LOCA
analysis. The amount of ice in the bed has no impact on the
initiation of an accident, but rather on the mitigation of the
accident. The containment integrity analysis shows that the proposed
reduced ice weight is sufficient to maintain the peak containment
pressure below the containment design pressure, and that the
containment heat removal systems function to rapidly reduce the
containment pressure and temperature in the event of a LOCA.
Therefore, containment integrity is maintained and the consequences
of an accident previously evaluated in the WBN dual-unit Updated
Final Safety Analysis Report (UFSAR) are not significantly
increased. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The ice condenser serves to limit the peak pressure inside
containment following a LOCA. TVA has evaluated the revised
containment pressure analysis and determined that sufficient ice
would be present to maintain the peak containment pressure below the
containment design pressure. Therefore, the reduced ice weight does
not create the possibility of an accident that is different than any
already evaluated in the WBN dual-unit UFSAR. No new accident
scenarios, failure mechanisms, or limiting single failures are
introduced as a result of this proposed change.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed TS ice weight SR limit is based on the conservatism
of the WBN Unit
[[Page 15389]]
1 WCOBRA/TRAC LOCA M&E [mass and energy] methodology in comparison
to the WBN Unit 2 operating conditions. The WBN Unit 1 WCOBRA/TRAC
LOCA M&E methodology is modeled on the WBN Unit 1 RSGs [replacement
steam generators], which have a greater mass, volume, and stored
metal energy than the WBN Unit 2 original model D3 SGs [steam
generators]. Additionally, the containment pressure calculations in
Section 6.2.1.3.3 of the WBN Unit 1 portion of the WBN dual-unit
UFSAR state that the analytical limit for the mass of ice assumed in
the WBN Unit 1 ice condenser, in order to limit the maximum
containment peak pressure from a LOCA to below the containment
design pressure, is 2,260,000 lb. The proposed revised TS SR ice
mass limit of 2,404,500 lb [pound] includes additional ice mass to
conservatively bound ice bed sublimation effects. Based on TVA's
evaluation and the revised containment analysis, TVA considers the
reduction of the ice mass limit to be acceptable for satisfying the
safety function of the ice condenser for the current SR interval.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Benjamin G. Beasley.
III. Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses and Combined Licenses,
Proposed No Significant Hazards Consideration Determination, and
Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Tennessee Valley Authority, Docket No. 50-391 Watts Bar Nuclear Plant,
Unit 2, Rhea County, Tennessee
Date of amendment request: November 14, 2016. A publicly-available
version is in ADAMS under Accession No. ML16320A161.
Brief description of amendment request: The proposed amendment
would revise the Watts Bar Nuclear Plant, Unit 2, Cyber Security Plan
Implementation Schedule for Milestone 8 and would revise the associated
license condition in the Facility Operating License.
Date of publication of individual notice in Federal Register:
January 5, 2017 (82 FR 1370).
Expiration date of individual notice: February 6, 2017 (public
comments); March 6, 2017 (hearing requests).
IV. Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in the
``Obtaining Information and Submitting Comments'' section of this
document.
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: March 22, 2016, as supplemented by
letter dated August 11, 2016.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.5.12, ``Primary Containment Leakage Rate Testing
Program,'' for the permanent extension of the Type A test interval up
to one test in 15 years, as stipulated in Nuclear Energy Institute
(NEI) 94-01, Revision 2-A, ``Industry Guideline for Implementing
Performance-Based Option of 10 CFR part 50, Appendix J,'' October 2008
(ADAMS Accession No. ML100620847). The license amendment request also
proposes to increase the containment isolation valves leakage test
intervals (i.e., Type C tests) from their current 60 months to 75
months by replacing TS 5.5.12.a. reference to Regulatory Guide 1.163,
``Performance-Based Containment Leak-Test Program'' (ADAMS Accession
No. ML003740058), with a reference to NEI 94-01, Revision 3-A (ADAMS
Accession No. ML12221A202), and the conditions and limitations
specified in NEI 94-01, Revision 2-A, to implement the performance-
based leakage testing program in accordance with title 10 of the Code
of Federal Regulations part 50, Appendix J, Option B. The amendment
also deletes from TS 5.5.12, text that authorized a one-time extension
of the Type A test interval to 2007 and revised paragraph 2.D of the
renewed facility operating license to reflect removal of a reference to
an exemption from 10 CFR part 50, Appendix J, requirements for testing
of containment air locks.
Date of issuance: March 9, 2017.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 205. A publicly-available version is in ADAMS under
Accession No. ML16351A460; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-43: Amendment revised
the renewed facility operating license and TSs.
Date of initial notice in Federal Register: June 7, 2016 (81 FR
36616). The August 11, 2016 supplement provided additional information
that clarified the application, did not expand the scope of the
application as originally
[[Page 15390]]
noticed, and did not change the staff's original proposed no
significant hazard consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 9, 2017.
No significant hazards consideration comments received: No.
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: September 26, 2016.
Brief description of amendments: The amendments revised Technical
Specification Section 2.1.1.2 to change the minimum critical power
ratio safety limit.
Date of issuance: March 10, 2017.
Effective date: As of date of issuance and shall be implemented for
Unit 1 prior to start-up from the 2018 refueling outage (March 2018)
and for Unit 2 prior to start-up from the 2017 refueling outage.
Amendment Nos.: 272 (Unit 1) and 300 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML17059D146; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-71 and DPR-62:
Amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: December 20, 2016 (81
FR 92866).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 10, 2017.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: May 5, 2016, as supplemented by
letter dated June 16, 2016.
Brief description of amendments: The amendments would modify the
McGuire Nuclear Station, Units 1 and 2, Technical Specifications (TS)
by removing footnote (c) from TS Table 3.3.2-1, ``Engineered Safety
Feature Actuation System Instrumentation,'' which is no longer
applicable, and by removing an expired footnote from TS 3.8.1, ``AC
Sources--Operating.''
Date of issuance: March 8, 2017.
Effective date: As of its date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 293 and 272. A publicly-available version is in
ADAMS under Accession No. ML17003A019; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and technical specifications.
Date of initial notice in Federal Register: July 5, 2016 (81 FR
43649). The supplemental letter dated June 16, 2016, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 8, 2017.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: May 10, 2016, as supplemented by
letters dated May 18, 2016, and January 31, 2017.
Brief description of amendment: The amendment revised the safety
function lift and lower setpoint tolerances of the safety/relief valves
that are listed in Surveillance Requirements 3.4.3.1 and 3.4.4.1 of the
Technical Specifications.
Date of issuance: March 9, 2017.
Effective date: As of its date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 240. A publicly-available version is in ADAMS under
Accession No. ML17052A125; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-21: The amendment
revised the Renewed Facility Operating License and Technical
Specifications.
Date of initial notice in Federal Register: July 19, 2016 (81 FR
46961). The supplemental letter January 31, 2017, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 9, 2017.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
1, Pope County, Arkansas
Date of amendment request: March 25, 2016.
Brief description of amendment: The amendment deleted Technical
Specification (TS) 5.5.8, ``Inservice Testing Program.'' A new defined
term, ``Inservice Testing Program,'' is added to TS Section 1.1,
``Definitions.'' Also, existing uses of the term ``Inservice Testing
Program'' in the TSs are capitalized throughout to indicate that it is
now a defined term. The NRC staff has concluded that the amendment is
consistent with Technical Specifications Task Force Traveler TSTF-545,
Revision 3, which was made available to the TSTF via NRC letter dated
December 11, 2015 (ADAMS Accession No. ML15317A071).
Date of issuance: March 10, 2017.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 257. A publicly-available version is in ADAMS under
Accession No. ML16165A423; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-51: The amendment
revised the Renewed Facility Operating License and Technical
Specifications.
Date of initial notice in Federal Register: June 7, 2016 (81 FR
36619).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 10, 2017.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: April 4, 2016.
Brief description of amendments: The amendments revised the
technical specification (TS) requirements for the high pressure coolant
injection (HPCI) and reactor core isolation cooling (RCIC) system
actuation instrumentation. Specifically, the amendments add a footnote
to the TSs indicating that the injection functions of drywell pressure-
high (HPCI only) and manual initiation (HPCI and RCIC) are not required
to be operable under low reactor pressure conditions.
Date of issuance: February 28, 2017.
[[Page 15391]]
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 224 (Unit 1) and 185 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16356A272; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. NPF-39 and NPF-85:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: June 7, 2016 (81 FR
36620).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 28, 2017.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-220, Nine Mile Point
Nuclear Station, Unit 1, (NMP1), Oswego County, New York
Date of amendment request: January 3, 2017.
Brief description of amendment: The amendment revised the NMP1
licensing basis related to alternative source term analysis in the
updated final safety analysis report (UFSAR) to allow the use of the
release fractions listed in Tables 1 and 3 of NRC Regulatory Guide
1.183, ``Alternative Radiological Source Terms for Evaluating Design
Basis Accidents at Nuclear Power Reactors,'' July 2000 (ADAMS Accession
No. ML003716792), for partial length fuel rods (PLRs) that are
operating above the peak burnup limit for the remainder of the current
operating cycle. In addition, the proposed change revised the NMP1
licensing basis to allow movement of irradiated fuel bundles containing
PLRs that have been in operation above 62,000 megawatt days per metric
tons of uranium (MWD/MTU).
Date of issuance: March 9, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 226. A publicly-available version is in ADAMS under
Accession No. ML17055A451; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-63: Amendment revised
the licensing basis related to alternative source term analysis in the
UFSAR.
Date of initial notice in Federal Register: January 31, 2017 (82 FR
8871).
The Commission's related evaluation of the amendment and final no
significant hazards consideration determination are contained in a
Safety Evaluation dated March 9, 2017.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date amendment request: May 17, 2016, as supplemented by letters
dated November 2, 2016, and March 1, 2017.
Brief description of amendment: The amendment revised and removed
certain requirements from the Section 6, ``Administrative Controls,''
portions of the Oyster Creek Nuclear Generating Station Technical
Specifications (TSs) that are not applicable to the facility in a
permanently defueled condition. In addition, the amendment added
definitions to TS Section 1, ``Definitions.'' Also, the amendment made
additions to, deletions from, and conforming administrative changes to
the TSs.
Date of issuance: March 7, 2017.
Effective date: Effective upon the licensee's submittal of the
certifications required by 10 CFR 50.82(a)(1)(i) and 50.82(a)(1)(ii),
and shall be implemented within 60 days of the effective date of the
amendment, but may not exceed March 29, 2020.
Amendment No.: 290. A publicly-available version is in ADAMS under
Accession No. ML16235A413; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-16: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 19, 2016 (81 FR
46963). On July 19, 2016, the NRC staff published a proposed no
significant hazards consideration (NSHC) determination regarding the
amendment request in the Federal Register (81 FR 46963). Subsequently,
by letter dated November 2, 2016, the licensee provided additional
information that expanded the scope of the amendment request as
originally noticed in the Federal Register. Accordingly, the NRC staff
published a second proposed NSHC determination regarding the amendment
request in the Federal Register on November 22, 2016 (81 FR 83876),
which superseded the original Federal Register notice in its entirety.
The supplemental letter dated March 1, 2017, provided additional
information that clarified the application, did not expand the scope of
the application as noticed, and did not change the NRC staff's second
proposed NSHC determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 7, 2017.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 50-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County,
Georgia
Date of amendment request: June 16, 2016.
Brief description of amendments: The amendments changed Combined
License Nos. NPF-91 and NPF-92 for the Vogtle Electric Generating Plant
Units 3 and 4. The amendments authorized changes to the Updated Final
Safety Analysis Report (UFSAR) in the form of departures from the
incorporated plant-specific Design Control Document Tier 2 information.
Specifically, the changes to the Technical Specifications (TS) and
information in the UFSAR revised the AP1000 protection and safety
monitoring system functional logic to comply with the requirements on
operating bypasses in Clause 6.6, ``Operating Bypasses'' of the
Institute of Electrical and Electronics Engineers (IEEE) Std. 603-1991,
``IEEE Standard Criteria for Safety Systems for Nuclear Power
Generating Stations.''
Date of issuance: February 24, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 71/70. A publicly-available version is in ADAMS
under Accession No. ML16320A097; documents related to these amendments
are listed in the Safety Evaluation enclosed with the amendments.
Facility Operating License Nos. NPF-91 and NPF-92: Amendment
revised the Facility Combined License and TS.
Date of initial notice in Federal Register: August 16, 2016 (81 FR
54610).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24, 2017.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1 (VCSNS), Fairfield County, South Carolina
Date of amendment request: June 30, 2016, as supplemented by letter
dated August 4, 2016.
Brief description of amendment: This amendment revised the date of
the
[[Page 15392]]
Cyber Security Plan implementation schedule for Milestone 8. Milestone
8 requires full implementation of the VCSNS Cyber Security Plan.
Date of issuance: March 9, 2017.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 208. A publicly-available version is in ADAMS under
Accession No. ML17011A050; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-12: Amendment revised
the Renewed Facility Operating License.
Date of initial notice in Federal Register: October 4, 2016 (81 FR
68472).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 9, 2017.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 16th day of March 2017.
For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2017-05990 Filed 3-27-17; 8:45 am]
BILLING CODE 7590-01-P