[Federal Register Volume 82, Number 98 (Tuesday, May 23, 2017)]
[Notices]
[Pages 23615-23632]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-10570]
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NUCLEAR REGULATORY COMMISSION
[NRC-2017-0120]
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 April 25, 2017, to May 8, 2017. The last
biweekly notice was published on May 9, 2017.
DATES: Comments must be filed by June 22, 2017. A request for a hearing
must be filed by July 24, 2017.
ADDRESSES: You may submit comments by any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0120. 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: T-8-D36M, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: 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-2017-0120, 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-2017-0120.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in this
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2017-0120, facility name, unit
number(s), plant docket number, application date, and subject in your
comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment
[[Page 23616]]
submissions available to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment
period such that failure to act in a timely way would result, for
example in derating or shutdown of the facility. If the Commission
takes action prior to the expiration of either the comment period or
the notice period, it will publish in the Federal Register a notice of
issuance. If the Commission makes a final no significant hazards
consideration determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and petition for leave to intervene
(petition) with respect to the action. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309. The NRC's regulations are accessible
electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place 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 July
24, 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
[[Page 23617]]
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 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
[[Page 23618]]
additional direction on obtaining information related to this document,
see the ``Obtaining Information and Submitting Comments'' section of
this document.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station (CNS), Units 1 and 2, York County, South Carolina
Date of amendment request: December 15, 2016. A publicly-available
version is in ADAMS under Accession No. ML16350A422.
Description of amendment request: The amendments would modify
Technical Specification (TS) 3.9.4, ``Residual Heat Removal (RHR) and
Coolant Circulation--High Water Level,'' and TS 3.9.5, ``Residual Heat
Removal (RHR) and Coolant Circulation--Low Water Level.'' Condition A
of TS 3.9.4 applies when RHR requirements are not met, and includes
four required actions. Required Action A.4 requires, within 4 hours,
the closure of all containment penetrations providing direct access
from containment atmosphere to outside atmosphere. The proposed changes
revise Required Action A.4 and add new Required Actions A.5, A.6.1, and
A.6.2 to clarify that the intent of the required actions is to
establish containment closure. Each of these required actions will have
a completion time of 4 hours. Condition B of TS 3.9.5 applies when no
RHR loop is in operation, and includes three required actions. Required
Action B.3 requires the closure of all containment penetrations
providing direct access from containment atmosphere to outside
atmosphere. The proposed changes are the same as the proposed changes
to TS 3.9.4, consisting of a revision to Required Action B.3 and the
addition of new Required Actions B.4, B.5.1, and B.5.2. These proposed
changes are consistent with Technical Specification Task Force (TSTF)
Traveler TSTF-197-A, Revision 2, ``Require Containment Closure When
Shutdown Cooling Requirements Are Not Met.''
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 revise the CNS TS to ensure that the
appropriate actions are taken to establish containment closure in
the event that Residual Heat Removal requirements are not met during
refueling operations. Containment closure would be appropriate for
mitigation of a loss of shutdown cooling accident, but it does not
affect the initiation of the accident. The containment purge system
isolation valves will be capable of being closed automatically on a
high containment radiation signal, such that there will be no
significant increase in the radiological consequences of a loss of
shutdown cooling.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any 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 physical alteration to the
plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The containment purge system isolation valves will remain
capable of being closed automatically on a high containment
radiation signal.
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 the margin of safety?
Response: No.
Currently the Technical Specifications are vague and overly
restrictive concerning the requirement for containment closure when
shutdown cooling is lost. The proposed changes eliminate unclear
requirements and provide a clear way to establish containment
closure that meets the [TS] Bases description, which is to prevent
radioactive gas from being released from the containment during a
loss of shutdown cooling incident. The containment purge system
isolation valves will remain capable of being closed automatically
on a high containment radiation signal.
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: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: January 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17025A069.
Description of amendment request: The amendments would modify
Technical Specification 3.1.2, ``Core Reactivity,'' to revise the
Completion Times of Required Action A.1 and A.2 from 72 hours to 7
days. This proposed change is consistent with Technical Specification
Task Force (TSTF) Traveler TSTF-142-A, Revision 0, ``Increase the
Completion Time when the Core Reactivity Balance is Not Within Limit.''
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 extend the Completion Time to take the
Required Actions when measured core reactivity is not within the
specified limit of the predicted values. The Completion Time to
respond to a difference between predicted and measured core
reactivity is not an initiator to any accident previously evaluated.
The radiological consequences of an accident during the proposed
Completion Time are no different from the consequences of an
accident during the existing Completion Time.
Therefore, the proposed changes do not involved a significant
increase in the probability or consequences of any 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 physical alteration to the
plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
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 the margin of safety?
Response: No.
The proposed changes provide additional time to investigate and
to implement appropriate operating restrictions when
[[Page 23619]]
measured core reactivity is not within the specified limit of the
predicted values. The additional time will not have a significant
effect on plant safety due to the conservatisms used in designing
the reactor core and performing the safety analyses, and the low
probability of an accident or transient which would approach the
core design limits during the additional time.
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: Kate B. Nolan, Deputy General Counsel, Duke
Energy Corporation, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station (MNS), Units 1 and 2, Mecklenburg County, North
Carolina
Date of amendment request: January 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17025A069.
Description of amendment request: The amendments would modify
Technical Specification (TS) 3.9.5, ``Residual Heat Removal (RHR) and
Coolant Circulation--High Water Level,'' and TS 3.9.6, ``Residual Heat
Removal (RHR) and Coolant Circulation--Low Water Level.'' Condition A
of TS 3.9.5 applies when RHR requirements are not met, and includes
four required actions. Required Action A.4 requires, within 4 hours,
the closure of all containment penetrations providing direct access
from containment atmosphere to outside atmosphere. The proposed changes
revise Required Action A.4 and add new Required Actions A.5, A.6.1, and
A.6.2 to clarify that the intent of the required actions is to
establish containment closure. Each of these required actions will have
a completion time of 4 hours. Condition B of TS 3.9.6 applies when no
RHR loop is in operation, and includes three required actions. Required
Action B.3 requires the closure of all containment penetrations
providing direct access from containment atmosphere to outside
atmosphere. The proposed changes are the same as the proposed changes
to TS 3.9.5, consisting of a revision to Required Action B.3 and the
addition of new Required Actions B.4, B.5.1, and B.5.2. These proposed
changes are consistent with Technical Specification Task Force (TSTF)
Traveler TSTF-197-A, Revision 2, ``Require Containment Closure When
Shutdown Cooling Requirements Are Not Met.''
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 revise the MNS TS to ensure that the
appropriate actions are taken to establish containment closure in
the event that Residual Heat Removal requirements are not met during
refueling operations. Containment closure would be appropriate for
mitigation of a loss of shutdown cooling accident, but it does not
affect the initiation of the accident. The containment purge system
isolation valves will be capable of being closed automatically on a
high containment radiation signal, such that there will be no
significant increase in the radiological consequences of a loss of
shutdown cooling.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any 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 physical alteration to the
plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The containment purge system isolation valves will remain
capable of being closed automatically on a high containment
radiation signal.
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 the margin of safety?
Response: No.
Currently the Technical Specifications are vague and overly
restrictive concerning the requirement for containment closure when
shutdown cooling is lost. The proposed changes eliminate unclear
requirements and provide a clear way to establish containment
closure that meets the [TS] Bases description, which is to prevent
radioactive gas from being released from the containment during a
loss of shutdown cooling incident. The containment purge system
isolation valves will remain capable of being closed automatically
on a high containment radiation signal.
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: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station (MNS), Units 1 and 2, Mecklenburg County, North
Carolina
Date of amendment request: January 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17025A069.
Description of amendment request: The amendments would modify
Technical Specification (TS) 3.6.3, ``Containment Isolation Valves,''
to add a Note to TS Limiting Condition for Operation (LCO) 3.6.3
Required Actions A.2, C.2 and E.2 to allow isolation devices that are
locked, sealed or otherwise secured to be verified by use of
administrative means. This proposed change is consistent with Technical
Specification Task Force (TSTF) Traveler TSTF-269-A, Revision 2,
``Allow Administrative Means of Position Verification for Locked or
Sealed Valves.''
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 modify MNS TS 3.6.3, ``Containment
Isolation Valves''. This TS currently includes actions that require
penetrations to be isolated and periodically verified to be
isolated. A Note is proposed to be added to TS 3.6.3 Required
Actions A.2, C.2, and E.2, to allow isolation devices that are
locked, sealed, or otherwise secured to be verified by use of
administrative means. The proposed changes do not affect any plant
equipment, test methods, or plant operation, and is not an initiator
of any analyzed accident sequence. The inoperable containment
penetrations will continue to be isolated, and hence perform their
isolation
[[Page 23620]]
function. Operation in accordance with the proposed TSs will ensure
that all analyzed accidents will continue to be mitigated as
previously analyzed.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any 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 physical alteration to the
plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
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 the margin of safety?
Response: No.
The proposed changes will not affect the operation of plant
equipment or the function of any equipment assumed in the accident
analysis. Affected containment penetrations will continue to be
isolated as required by the existing TS.
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: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: January 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17025A069.
Description of amendment request: The amendments would modify
Technical Specification (TS) 3.8.1, ``AC [Alternating Current]
Sources--Operating,'' to allow greater flexibility in performing
Surveillance Requirements (SRs) by modifying Mode restriction notes in
TS SRs 3.8.1.8, 3.8.1.11, 3.8.1.16, 3.8.1.17, and 3.8.1.19. This
proposed change is consistent with Technical Specification Task Force
(TSTF) Traveler TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode
Restriction Notes.''
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 modify Mode restriction Notes in TS SRs
3.8.1.8, 3.8.1.11, 3.8.1.16, 3.8.1.17, and 3.8.1.19 to allow
performance of the Surveillance in whole or in part to reestablish
Diesel Generator (DG) Operability, and to allow the crediting of
unplanned events that satisfy the Surveillance(s) [Requirements].
The emergency diesel generators and their associated emergency loads
are accident mitigating features, and are not an initiator of any
accident previously evaluated. As a result, the probability of any
accident previously evaluated is not significantly increased. To
manage any increase in risk, the proposed changes require an
assessment to verify that plant safety will be maintained or
enhanced by performance of the Surveillance in the current
prohibited Modes. The radiological consequences of an accident
previously evaluated during the period that the DG is being tested
to reestablish operability are no different from the radiological
consequences of an accident previously evaluated while the DG is
inoperable. As a result, the consequences of any accident previously
evaluated are not increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any 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 physical alteration to the
plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
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 the margin of safety?
Response: No.
The purpose of Surveillances is to verify that equipment is
capable of performing its assumed safety function. The proposed
changes will only allow the performance of the Surveillances to
reestablish operability, and the proposed changes may not be used to
remove a DG from service. In addition, the proposed changes will
potentially shorten the time that a DG is unavailable because
testing to reestablish operability can be performed without a plant
shutdown. The proposed changes also require an assessment to verify
that plant safety will be maintained or enhanced by performance of
the Surveillance in the current prohibited Modes.
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: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: January 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17025A069.
Description of amendment request: The amendments would modify
Technical Specification (TS) 3.4.12, ``Low Temperature Overpressure
Protection (LTOP) System,'' to increase the time allowed for swapping
charging pumps to 1 hour. Additionally, an existing note in the
Applicability section of TS 3.4.12 is being reworded and relocated to
the Limiting Condition for Operation section of TS 3.4.12 as Note 2.
These proposed changes are consistent with Technical Specification Task
Force (TSTF) Traveler TSTF-285-A, Revision 1, ``Charging Pump Swap LTOP
Allowance.''
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 increase the time allowed for swapping
charging pumps from 15 minutes to one hour, and make several other
associated administrative changes and clarifications to the TS.
These changes do not affect event initiators or precursors. Thus,
the proposed changes do not involve a
[[Page 23621]]
significant increase in the probability of an accident previously
evaluated. In addition, the proposed changes do not alter any
assumptions previously made in the radiological consequence
evaluations nor affect mitigation of the radiological consequences
of an accident described in the Updated Final Safety Analysis Report
(UFSAR). As such, the consequences of accidents previously evaluated
in the UFSAR will not be increased and no additional radiological
source terms are generated. Therefore, there will be no reduction in
the capability of those structures, systems, and components (SSCs)
in limiting the radiological consequences of previously evaluated
accidents, and reasonable assurance that there is no undue risk to
the health and safety of the public will continue to be provided.
Thus, the proposed changes do not involve a significant increase in
the consequences of an accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed 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 physical changes to analyzed
SSCs or changes to the modes of plant operation defined in the
technical specification. The proposed changes do not involve the
addition or modification of plant equipment (no new or different
type of equipment will be installed) nor do they alter the design or
operation of any plant systems. No new accident scenarios, accident
or transient initiators or precursors, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
changes. The proposed changes do not cause the malfunction of
safety-related equipment assumed to be operable in accident
analyses. No new or different mode of failure has been created and
no new or different equipment performance requirements are imposed
for accident mitigation. As such, the proposed changes have no
effect on previously evaluated 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 amendment involve a significant reduction
in the margin of safety?
Response: No.
The proposed changes do not adversely affect any current plant
safety margins or the reliability of the equipment assumed in the
safety analysis. Therefore, there are no changes being made to any
safety analysis assumptions, safety limits or limiting safety system
settings that would adversely affect plant safety as a result of the
proposed changes.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: January 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17025A069.
Description of amendment request: The amendments would modify
Technical Specification (TS) 3.1.8, ``PHYSICS TESTS Exceptions,'' to
allow the numbers of channels required by the Limiting Condition for
Operation (LCO) section of TS 3.3.1, ``Reactor Trip System (RTS)
Instrumentation,'' to be reduced from ``4'' to ``3'' to allow one
nuclear instrumentation channel to be used as an input to the
reactivity computer for physics testing without placing the nuclear
instrumentation channel in a tripped condition. This proposed change is
consistent with Technical Specification Task Force (TSTF) Traveler
TSTF-315-A, Revision 0, ``Reduce Plant Trips Due to Spurious Signals to
the Nuclear Instrumentation System (NIS) During Physics Testing.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes revise TS 3.1.8, ``PHYSICS TESTS
Exceptions,'' to allow the number of channels required by LCO 3.3.1,
``RTS Instrumentation,'' to be reduced from ``4'' to ``3'', to allow
one nuclear instrumentation channel to be used as an input to the
reactivity computer for physics testing without placing the nuclear
instrumentation channel in a tripped condition. A reduction in the
number of required nuclear instrumentation channels is not an
initiator to any accident previously evaluated. With the nuclear
instrumentation channel placed in bypass instead of in trip, reactor
protection is still provided by the nuclear instrumentation system
operating in a two-out-of-three channel logic. As a result, the
ability to mitigate any accident previously evaluated is not
significantly affected. The proposed changes will not affect the
source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of any
accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any 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 physical alteration to the
plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
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 the margin of safety?
Response: No.
The proposed changes reduce the probability of a spurious
reactor trip during physics testing. The reactor trip system
continues to be capable of protecting the reactor utilizing the
power range neutron flux trips operating in a two-out-of-three trip
logic. As a result, the reactor is protected and the probability of
a spurious reactor trip is significantly reduced.
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: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: January 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17025A069.
Description of amendment request: The amendments would modify
Technical Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW)
System,'' to expand the TS 3.7.5 Limiting
[[Page 23622]]
Condition for Operation, Condition A to include the situation when one
turbine driven AFW pump is operable in MODE 3, immediately following a
refueling outage (if MODE 2 has not been entered), with a 7-day
Completion Time. This proposed change is consistent with Technical
Specification Task Force (TSTF) Traveler TSTF-340-A, Revision 3,
``Allow 7 Day Completion Time for a Turbine-Driven AFW Pump
Inoperable.''
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 revise TS 3.7.5, ``Auxiliary Feedwater
(AFW) System,'' to allow a 7 day Completion Time to restore an
inoperable AFW turbine-driven pump in MODE 3 immediately following a
refueling outage, if MODE 2 has not been entered. An inoperable AFW
turbine-driven pump is not an initiator of any accident previously
evaluated. The ability of the plant to mitigate an accident is no
different while in the extended Completion Time than during the
existing Completion Time. The proposed changes will not affect the
source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of any
accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any 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 physical alteration to the
plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
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 the margin of safety?
Response: No.
The proposed changes revise TS 3.7.5, ``Auxiliary Feedwater
(AFW) System,'' to allow a 7 day Completion Time to restore an
inoperable turbine-driven AFW pump in Mode 3, immediately following
a refueling outage, if Mode 2 has not been entered. In Mode 3
immediately following a refueling outage, core decay heat is low and
the need for AFW is also diminished. The two operable motor driven
AFW pumps are available and there are alternate means of decay heat
removal if needed. As a result, the risk presented by the extended
Completion Time is minimal.
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: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station (MNS), Units 1 and 2, Mecklenburg County, North
Carolina
Date of amendment request: January 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17025A069.
Description of amendment request: The amendments would modify
Technical Specification (TS) 3.4.10, ``Pressurizer Safety Valves,'' TS
3.4.12, ``Low Temperature Overpressure Protection (LTOP) System,'' and
TS 3.7.4, ``Steam Generator Power Operated Relief Valves (SG PORVs),''
to revise the Completion Times for Limiting Condition for Operation
(LCO) 3.4.10 Required Action B.2, and LCO 3.7.4 Required Action C.2
from 12 to 24 hours and LCO 3.4.12 Required Action G.1 from 8 to 12
hours. The proposed changes are consistent with Technical Specification
Task Force (TSTF) Traveler TSTF-352-A, Revision 1, ``Provide Consistent
Completion Time to Reach MODE 4.''
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 allow a more reasonable time to plan and
execute required actions, and will not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained. The proposed changes will not
alter or prevent the ability of structures, systems, and components
(SSCs) from performing their intended functions to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed changes do not physically alter safety-related
systems nor affect the way in which safety-related systems perform
their functions. All accident analysis acceptance criteria will
continue to be met with the proposed changes. The proposed changes
will not affect the source term, containment isolation, or
radiological release assumptions used in evaluating the radiological
consequences of an accident previously evaluated. The proposed
changes will not alter any assumptions or change any mitigation
actions in the radiological consequence evaluations in the MNS
Updated Final Safety Analysis Report (UFSAR). The applicable
radiological dose acceptance criteria will continue to be met.
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.
There are no proposed design changes nor are there any changes
in the method by which any safety-related plant SSC performs its
safety function. The proposed changes will not affect the normal
method of plant operation or change any operating parameters. No
equipment performance requirements will be affected. The proposed
changes will not alter any assumptions made in the safety analyses.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of this amendment. There will be no adverse effect or
challenges imposed on any safety-related system as a result of this
amendment.
Therefore, the proposed changes do not create the possibility of
a new or different accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their intended functions.
These barriers include the fuel cladding, the reactor coolant system
pressure boundary, and the containment barriers. The proposed
changes will not have any impact on these barriers. No accident
mitigating equipment will be adversely impacted.
Therefore, existing safety margins will be preserved. None of
the proposed changes will 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
[[Page 23623]]
amendment request involves no significant hazards consideration.
Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station (MNS), Units 1 and 2, Mecklenburg County, North
Carolina
Date of amendment request: January 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17025A069.
Description of amendment request: The amendments would modify
Technical Specification (TS) 3.9.6, ``Residual Heat Removal (RHR) and
Coolant Circulation--Low Water Level,'' to add Note 1 to the Limiting
Condition for Operation (LCO) Section of TS 3.9.6 to allow the securing
of the operating train of RHR for up to 15 minutes to support switching
operating trains. The allowance is restricted to three conditions: (a)
The core outlet temperature is maintained greater than 10 degrees
Fahrenheit below saturation temperature; (b) no operations are
permitted that would cause an introduction of coolant into the Reactor
Coolant System (RCS) with boron concentration less than that required
to meet the minimum required boron concentration of LCO 3.9.1; and (c)
no draining operations to further reduce RCS water volume are
permitted. Additionally, the amendments would modify the LCO Section of
TS 3.9.6 to add Note 2 which would allow one required RHR loop to be
inoperable for up to 2 hours for surveillance testing, provided that
the other RHR loop is operable and in operation. These proposed changes
are consistent with Technical Specification Task Force (TSTF) Travelers
TSTF-349-A, Revision 1, ``Add Note to LCO 3.9.5 Allowing Shutdown
Cooling Loops Removal from Operation,'' TSTF-361-A, Revision 2, ``Allow
Standby SDC [Shutdown Cooling]/RHR/DHR [Decay Heat Removal] Loop to be
Inoperable to Support Testing,'' and TSTF-438-A, Revision 0, ``Clarify
Exception Notes to be Consistent with the Requirement Being Excepted.''
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 add two notes to MNS TS LCO 3.9.6. Note 1
would allow securing the operating train of Residual Heat Removal
(RHR) for up to 15 minutes to support switching operating trains,
subject to certain restrictions. Note 2 to would allow one RHR loop
to be inoperable for up to 2 hours for surveillance testing provided
the other RHR loop is Operable and in operation. These provisions
are operational allowances. Neither operational allowance is an
initiator to any accident previously evaluated. In addition, the
proposed changes will not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of any accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any 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 physical alteration to the
plant (i.e., no new or different type of equipment will be
installed) or a change to the methods governing normal plant
operation. The changes do not alter the assumptions made in the
safety analysis.
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 the margin of safety?
Response: No.
An operational allowance is proposed which would allow securing
the operating train of RHR for up to 15 minutes to support switching
operating trains, subject to certain restrictions. Considering these
restrictions, combined with the short time frame allowed to swap
operating RHR trains, and the ability to start an operating RHR
train, if needed, the occurrence of an event that would require
immediate operation of an RHR train is extremely remote.
An operational allowance is also proposed which would allow one
RHR loop to be inoperable for up to 2 hours for surveillance testing
provided the other RHR loop is operable and in operation. A similar
allowance currently appears in MNS TS 3.4.7, ``Reactor Coolant
System (RCS) Loops--MODE 5, Loops Filled,'' and MNS TS 3.4.8, ``RCS
Loops--MODE 5, Loops Not Filled,'' and the conditions under which
the operational allowance would be applied in TS 3.9.6 are not
significantly different from those specifications. This operational
allowance provides the flexibility to perform surveillance testing,
while ensuring that there is reasonable time for operators to
respond to and mitigate any expected failures. The purpose of the
RHR System is to remove decay and sensible heat from the Reactor
Coolant System, to provide mixing of borated coolant, and to prevent
boron stratification. Removal of system components from service as
described above, and with limitations in place to maintain the
ability of the RHR System to perform its safety function, does not
significantly impact the margin of safety. Operators will continue
to have adequate time to respond to any off-normal events. Removing
the system from service, for a limited period of time, with other
operational restrictions, limits the consequences to those already
assumed in the Updated Final Safety Analysis Report (UFSAR).
Therefore, the proposed changes do 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: Kate B. Nolan, Deputy General Counsel, Duke
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC
28202-1802.
NRC Branch Chief: Michael T. Markley.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant (PNP), Van Buren County, Michigan
Date of amendment request: March 30, 2017. A publicly-available
version is in ADAMS under Accession No. ML17089A380.
Description of amendment request: The proposed amendment would
revise the PNP Cyber Security Plan (CSP) Milestone 8 full
implementation date from December 15, 2017, to May 31, 2020. This
amendment request is in support of PNP's transition, starting on
October 1, 2018, from an operating power plant to a decommissioned
plant.
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 CSP implementation schedule is
administrative in nature. This change does not alter accident
analysis assumptions, add any initiators, or affect the function of
plant systems or the manner in which systems are operated,
maintained, modified, tested, or inspected. The proposed change does
not require any
[[Page 23624]]
plant modifications which affect the performance capability of the
structures, system, and components relied upon to mitigate the
consequences of postulated accidents, and has no impact on the
probability or consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the CSP implementation schedule is
administrative in nature. This proposed change does not alter
accident analysis assumptions, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected. The proposed
change does not require any plant modifications which affect the
performance capability of the structures, systems, and components
relied upon to mitigate the consequences of postulated accidents and
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Plant safety margins are established through limiting conditions
for operation, limiting safety system settings, and safety limits
specified in the technical specifications. The proposed changes to
the CSP implementation schedule is administrative in nature. In
addition, the milestone date delay for full implementation of the
CSP has no substantive impact because other measures have been taken
which provide adequate protection during this period of time.
Because there is no change to established safety margins as a result
of this change, the proposed change does not involve a significant
reduction in a margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeanne Cho, Senior Counsel, Entergy
Services, Inc., 440 Hamilton Ave., White Plains, NY 10601.
NRC Branch Chief: David J. Wrona.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: March 30, 2017. A publicly-available
version is in ADAMS under Accession No. ML17101A608.
Description of amendment request: The amendment would revise the
renewed facility operating license Paragraph 3.G, ``Physical
Protection.'' The amendment would revise the Pilgrim Nuclear Power
Station Cyber Security Plan (CSP) implementation schedule for Milestone
8 full implementation date from December 15, 2017, to December 31,
2020.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the CSP implementation schedule is
administrative in nature. The change does not alter accident
analysis assumptions, add any initiators, or affect the function of
plant systems or the manner in which systems are operated,
maintained, modified, tested, or inspected. The proposed change does
not require any plant modifications which affect the performance
capability of the structures, systems, and components relied upon to
mitigate the consequences of postulated accidents, and has no impact
on the probability or consequences of an accident previously
evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to the CSP implementation schedule is
administrative in nature. The proposed change does not alter
accident analysis assumptions, add any initiators, or affect the
function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected. The proposed
change does not require any plant modifications which affect the
performance capability of the structures, systems, and components
relied upon to mitigate the consequences of postulated accidents,
and does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Plant safety margins are established through limiting conditions
for operation, limiting safety system settings, and safety limits
specified in the technical specifications. The proposed change to
the CSP implementation schedule is administrative in nature. In
addition, the milestone date delay for full implementation of the
CSP has no substantive impact because other measures have been taken
which provide adequate protection during this period of time.
Because there is no change to established safety margins as a result
of this change, the proposed change does not involve a significant
reduction in a margin of safety.
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.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.
(the licensees), Docket Nos. 50-416 and 72-50, Grand Gulf Nuclear
Station, Unit 1 (Grand Gulf), and Independent Spent Fuel Storage
Installation (ISFSI), Claiborne County, Mississippi
Date of amendment request: March 29, 2017. A publicly-available
version is in ADAMS under Accession No. ML17093A729.
Description of amendment request: The proposed amendment would make
an administrative change to the name of South Mississippi Electric
Power Association, one of the licensees for Grand Gulf and its ISFSI.
Effective November 10, 2016, South Mississippi Electric Power
Association changed its corporate name from ``South Mississippi
Electric Power Association'' to ``Cooperative Energy, a Mississippi
Electric Cooperative.'' The corporate name was changed for commercial
reasons. The changes proposed herein to the Grand Gulf operating
license solely reflects the changed licensee name. This name change is
purely administrative in nature. This request does not involve a
transfer of control or of an interest in the license.
[[Page 23625]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes ``involve a significant increase in
the probability or consequences of an accident previously
evaluated''?
Response: No.
The proposed amendments simply change the name of a licensee.
The name change is purely administrative. None of the functions or
responsibility of any of the Grand Gulf licensees will change as a
result of the amendments. The proposed amendments do not alter the
design, function, or operation of any plant equipment. As such, the
accident and transient analyses contained in the facility updated
final safety analysis report will not be affected.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes ``create the possibility of a new or
different kind of accident from any accident previously evaluated''?
Response: No.
The proposed amendments simply change the name of a licensee.
The proposed name change is purely administrative. None of the
functions or responsibility of any of the Grand Gulf licensees will
change as a result of the amendments. The proposed amendments do not
alter the design, function, or operation of any plant equipment. As
such, the accident and transient analyses contained in the facility
updated final safety analysis report will not be affected.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes ``involve a significant reduction in
the margin of safety''?
Response: No.
The proposed amendments simply change the name of a licensee.
The name change is purely administrative. None of the functions or
responsibility of any of the Grand Gulf licensees will change as a
result of the amendments. The proposed amendments do not alter the
design, function, or operation of any plant equipment. As such, the
accident and transient analyses contained in the facility updated
final safety analysis report will not be affected.
Therefore, the proposed changes do 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: William B. Glew, Jr., Associate General
Counsel--Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New
York 10601.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne
County, Mississippi
Date of amendment request: December 29, 2016. A publicly-available
version is in ADAMS under Accession No. ML16364A338.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) for GGNS. The amendment would
allow for a one cycle extension to the 10-year frequency of the GGNS
containment integrated leakage rate test (ILRT) or Type A test and the
drywell bypass leak rate test (DWBT). These tests are required by TS
5.5.12, ``10 CFR part 50, Appendix J [Primary Reactor Containment
Leakage Testing for Water-Cooled Power Reactors], Testing Program,''
and TS Surveillance Requirement 3.6.5.1.1, respectively. The proposed
change would permit the existing ILRT and DWBT frequency to be extended
from 10 years to 11.5 years.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below, with NRC edits in [brackets]:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment to the Technical Specifications (TS)
involves the extension of the Grand Gulf Nuclear Station, Unit 1
(GGNS) Type A integrated leakage rate test and the drywell bypass
leakage rate test intervals to 11.5 years.
The proposed extension does not involve either a physical change
to the plant or a change in the manner in which the plant is
operated or controlled. The containment is designed to provide an
essentially leak tight barrier against the uncontrolled release of
radioactivity to the environment for postulated accidents. Type B
and C testing ensures that individual containment isolation valves
are essentially leak tight. In addition, aggregate Type B and C
leakage rates support the leakage tightness of primary containment
by minimizing potential leakage paths. The assessment of the [leak-
tightness] of the drywell will continue to be performed at least
once each operating cycle. The proposed amendment will not change
the leakage rate acceptance requirements. As such, the containment
will continue to perform its design function as a barrier to fission
product releases. In addition, the containment and the testing
requirements invoked to periodically demonstrate the integrity of
the containment and the assessment of the [leak-tightness] of the
drywell exist to ensure the plant's ability to mitigate the
consequences of an accident, and do not involve the prevention or
identification of any precursors of an accident. Therefore, this
proposed extension does not involve a significant increase in the
probability of an accident previously evaluated.
Therefore, the proposed change does not result in a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment to the Technical Specifications (TS)
involves the extension of the Grand Gulf Nuclear Station, Unit 1
(GGNS) Type A integrated leakage rate test and the drywell bypass
leakage rate test intervals to 11.5 years. The containment and the
testing requirements to periodically demonstrate the integrity of
the containment exist to ensure the plant's ability to mitigate the
consequences of an accident do not involve any accident precursors
or initiators. The proposed change does not involve a physical
change to the plant (i.e., no new or different type of equipment
will be installed) or a change to the manner in which the plant is
operated or controlled.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The proposed amendment to the Technical Specifications (TS)
involves the extension of the Grand Gulf Nuclear Station, Unit 1
(GGNS) Type A integrated leakage rate test and the drywell bypass
leakage rate test intervals to 11.5 years. This amendment does not
alter the manner in which safety limits, limiting safety system set
points, or limiting conditions for operation are determined. The
specific requirements and conditions of the TS 10 CFR part 50,
Appendix J, Testing Program for containment leak rate testing exist
to ensure that the degree of containment structural integrity and
leak-tightness that is considered in the plant safety analysis is
maintained. The overall containment leak rate limit specified by TS
is maintained.
The proposed change involves the extension of the interval for
only the Type A containment leakage rate test and the drywell bypass
leakage rate test for GGNS. The proposed surveillance interval
extension is bounded by the 15-year Type A test interval currently
authorized within NEI 94-01, Revision 3-A. The design, operation,
testing methods, and acceptance criteria for Types A, B, and C
containment leakage tests specified in applicable codes and
standards would continue to be met with the
[[Page 23626]]
acceptance of this proposed change, since these are not affected by
the proposed changes to the Type A test interval. In addition to the
scheduled performance of DWBT GGNS will continue to monitor the
drywell for significant leakage during operation.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William B. Glew, Jr., Associate General
Counsel--Entergy Services, Inc., 440 Hamilton Avenue, White Plains, New
York 10601.
NRC Branch Chief: Robert J. Pascarelli.
Exelon Generation Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station (OCNGS), Ocean County, New Jersey
Date of amendment request: April 10, 2017. A publicly-available
version is available in ADAMS under Accession No. ML17100A844.
Description of amendment request: The amendment would revise the
OCNGS Cyber Security Plan (CSP) Milestone 8 (MS8) full implementation
completion date, as set forth in the CSP implementation schedule, and
revise the physical protection license condition in the renewed
facility operating license. The licensee proposes to revise the CSP MS8
completion date from December 31, 2017, to August 31, 2021.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The amendment request proposes a change to the OCNGS CSP MS8
completion date as set forth in the CSP implementation schedule and
associated regulatory commitments. The NRC staff has concluded that
the proposed change: (1) Does not alter accident analysis
assumptions, add any initiators, or affect the function of plant
systems or the manner in which systems are operated, maintained,
modified, tested, or inspected; (2) does not require any plant
modifications which affect the performance capability of the
structures, systems, and components relied upon to mitigate the
consequences of postulated accidents; and (3) has no impact on the
probability or consequences of an accident previously evaluated. In
addition, the NRC staff has concluded that the proposed change to
the CSP implementation schedule is administrative in nature.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The NRC staff has concluded the proposed change: (1) Does not
alter accident analysis assumptions, add any initiators, or affect
the function of plant systems or the manner in which systems are
operated, maintained, modified, tested, or inspected; and (2) does
not require any plant modifications which affect the performance
capability of the structures, systems, and components relied upon to
mitigate the consequences of postulated accidents and does not
create the possibility of a new or different kind of accident from
any accident previously evaluated. In addition, the NRC staff has
concluded that the proposed change to the OCNGS CSP MS8
implementation schedule is administrative in nature.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Plant safety margins are established through limiting conditions
for operation, limiting safety system settings, and safety limits
specified in the technical specifications. The delay of the full
implementation date for the OCNGS CSP MS8 has no substantive impact
because other measures have been taken which provide adequate
protection for the plant during this period of time. Therefore, the
NRC staff has concluded that there is no significant reduction in a
margin of safety. In addition, the NRC staff has concluded that the
proposed change to the OCNGS CSP MS8 implementation schedule is
administrative in nature.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
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.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: March 24, 2017. A publicly-available
version is in ADAMS under Accession No. ML17087A012.
Description of amendment request: The proposed changes would modify
Technical Specifications (TS) Section 3.7.2, ``Steam Generator Stop
Valves (SGSVs),'' to incorporate the SGSV actuator trains into the
Limiting Condition for Operation and provide associated Conditions and
Required Actions. In addition, Surveillance Requirement (SR) 3.7.2.2
would be revised to clearly identify that the SGSV actuator trains are
required to be tested in accordance with the SR.
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 provide requirements for SGSVs that have
dual actuators which receive signals from separate instrumentation
trains. The design and functional performance requirements,
operational characteristics, and reliability of the SGSVs and
actuator trains are unchanged. There is no impact on the design
safety function of the SGSVs to close (as an accident mitigator),
nor is there any change with respect to inadvertent closure of an
SGSV (as a potential transient initiator). Since no failure mode or
initiating condition that could cause an accident (including any
plant transient) is created or affected, the change cannot involve a
significant increase in the probability of an accident previously
evaluated.
With regard to the consequences of an accident and the equipment
required for mitigation of the accident, the proposed changes
involve no design or physical changes to the SGSVs or any other
equipment required for accident mitigation. With respect to SGSV
actuator train Completion Times, the consequences of an accident are
independent of equipment Completion Times as long as adequate
equipment availability is maintained. The proposed SGSV actuator
Completion Times take into account the redundancy of the actuator
trains and are limited in extent consistent with other Completion
Times specified in the TS. Adequate equipment availability would
therefore continue to be required by the TS. On this basis, the
consequences of applicable, analyzed accidents are not significantly
affected by the proposed changes.
[[Page 23627]]
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to incorporate requirements for the SGSV
actuator trains in TS 3.7.2 do not involve any design or physical
changes to the facility, including the SGSVs and actuator trains
themselves. No physical alteration of the plant is involved, as no
new or different type of equipment is to be installed. The proposed
changes do not alter any assumptions made in the safety analyses,
nor do they involve any changes to plant procedures for ensuring
that the plant is operated within analyzed limits. As such, no new
failure modes or mechanisms that could cause a new or different kind
of accident from any previously evaluated are being introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes to incorporate requirements for the SGSV
actuator trains do not alter the manner in which safety limits or
limiting safety system settings are determined. No changes to
instrument/system actuation setpoints are involved. The safety
analysis acceptance criteria are not affected by this change and the
proposed changes will not permit plant operation in a configuration
outside the design basis.
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: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
NRC Branch Chief: David J. Wrona.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center (DAEC), Linn County, Iowa
Date of amendment request: March 24, 2017. A publicly-available
version is in ADAMS under Accession No. ML17086A442.
Description of amendment request: The proposed change would
relocate cycle specific minimum critical power ratio (MCPR) values to
the DAEC core operating limits report (COLR). The proposed amendment
would revise the DAEC technical specifications (TS) to modify TS Table
3.3.2.1-1, ``Control Rod Block Instrumentation,'' Footnotes (a) through
(e), and would relocate cycle specific MCPR values previously specified
in TS Table 3.3.2.1-1, Footnotes (a) through (e) to TS 5.6.5(a)(4) by
reference to the DAEC COLR.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change is an administrative change that does not
affect any plant systems, structures, or components designed for the
prevention or mitigation of previously evaluated accidents. No new
equipment is added nor is installed equipment being changed or
operated in a different manner.
Relocation of the Control Rod Block Instrumentation MCPR values
to the COLR has no influence or impact on, nor does it contribute in
any way to the probability or consequences of transients or
accidents. The COLR will continue to be controlled by the NextEra
programs and procedures that comply with TS 5.6.5. Transient
analyses addressed in the Final Safety Analysis Report will continue
to be performed in the same manner with respect to changes in the
cycle-dependent parameters obtained from the use of NRC-approved
reload design methodologies, which ensures that the transient
evaluation of new reloads are bounded by previously accepted
analyses.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of a previously evaluated
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed administrative change does not involve any changes
to the operation, testing, or maintenance of any safety-related, or
otherwise important to safety systems. All systems important to
safety will continue to be operated and maintained within their
design bases. Relocation of the Control Rod Block Instrumentation
MCPR values to the COLR has no influence or impact on new or
different kind of accidents.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is not affected by the relocation of cycle-
specific Control Rod Block Instrumentation MCPR values from the TS
to the COLR. Appropriate measures exist to control the values of
these cycle-specific limits since it is required by TS that only
NRC-approved methods be used to determine the limits. The proposed
change continues to require operation within the core thermal limits
as obtained from NRC-approved reload design methodologies and the
actions to be taken if a limit is exceeded remain unchanged, again,
in accordance with existing TS.
Therefore, the proposed change has no impact to 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: William Blair, P.O. Box 14000, Juno Beach,
FL 33408-0420.
NRC Branch Chief: David J. Wrona.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
No. 1, Rockingham County, New Hampshire
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: March 30, 2017. A publicly-available
version is in ADAMS under Accession No. ML17093A688.
Description of amendment request: The amendments would revise
technical specification requirements to operate ventilation systems
with charcoal filters from 10 hours to 15 minutes in accordance with
Technical Specifications Task Force (TSTF) Traveler TSTF-522, Revision
0, ``Revise Ventilation System Surveillance Requirements to Operate for
10 hours per Month.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change replaces an existing Surveillance
Requirement to operate the CREMAFS [Control Room Emergency Makeup
Air and Filtration System], FSBEACS [Fuel Storage Building Emergency
Air Cleaning System], and SBVS [Shield Building Ventilation System]
equipped with electric heaters for at least a continuous 10-hour
period in accordance with the SFCP [Surveillance Frequency Control
Program] with a requirement to operate the systems for 15 continuous
minutes with heaters operating.
[[Page 23628]]
These systems are not accident initiators and therefore, these
changes do not involve a significant increase in the probability of
an accident. The proposed system and filter testing changes are
consistent with current regulatory guidance for these systems and
will continue to assure that these systems perform their design
function which may include mitigating accidents. Thus, the change
does not involve a significant increase in the consequences of an
accident.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change replaces an existing Surveillance
Requirement to operate the CREMAFS, FSBEACS, and SBVS equipped with
electric heaters for at least a continuous 10-hour period in
accordance with the SFCP with a requirement to operate the systems
for 15 continuous minutes with heaters operating.
The change proposed for these ventilation systems does not
change any system operations or maintenance activities. Testing
requirements will be revised and will continue to demonstrate that
the Limiting Conditions for Operation are met and the system
components are capable of performing their intended safety
functions. The change does not create new failure modes or
mechanisms and no new accident precursors are generated.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change replaces an existing Surveillance
Requirement to operate the CREMAFS, FSBEACS, and SBVS equipped with
electric heaters for at least a continuous 10-hour period in
accordance with the SFCP with a requirement to operate the systems
for 15 continuous minutes with heaters operating.
The design basis for the ventilation systems' heaters is to heat
the incoming air which reduces the relative humidity. The heater
testing change proposed will continue to demonstrate that the
heaters are capable of heating the air and will perform their design
function. The proposed change is consistent with regulatory
guidance.
Therefore, it is concluded that this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William Blair, Managing Attorney--Nuclear,
Florida Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-
0420.
NRC Branch Chief: James G. Danna.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2 (PINGP),
Goodhue County, Minnesota
Date of amendment request: March 29, 2017. A publicly-available
version is in ADAMS under Accession No. ML17094A565.
Brief description of amendment request: The proposed amendments
would revise the current emergency action levels (EAL) scheme used at
PINGP to the EAL scheme contained in NEI 99-01, Revision 6,
``Development of Emergency Action Levels.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the PINGP EAL scheme does not impact the
physical function of plant structures, systems or components (SSC)
or the manner in which the SSCs perform their design function. The
proposed change neither adversely affects accident initiators or
precursors, nor alters design assumptions. Therefore, the proposed
change does not alter or prevent the ability of SSCs to perform
their intended function to mitigate the consequences of an event.
The Emergency Plan, including the associated EALs, is implemented
when an event occurs and cannot increase the probability of an
accident. Further, the proposed change does not reduce the
effectiveness of the Emergency Plan to meet the emergency planning
requirements established in 10 CFR 50.47 and 10 CFR part 50,
Appendix E.
Therefore, the proposed EAL scheme change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve any physical alteration to
the plant, that is, no new or different type of equipment will be
installed. The proposed change also does not change the method of
plant operation and does not alter assumptions made in the safety
analysis. Therefore, the proposed change will not create new failure
modes or mechanisms that could result in a new or different kind of
accident. The emergency plan, including the associated EAL scheme,
is implemented when an event occurs and is not an accident
initiator.
Therefore, the proposed EAL scheme change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is provided by the ability of accident
mitigation SCCs to perform at their analyzed capability. The change
proposed in this license amendment request does not modify any plant
equipment and there is no impact to the capability of the equipment
to perform its intended accident mitigation function. The proposed
change does not impact operation of the plant or its response to
transients or accidents. Additionally, the proposed changes will not
change any criteria used to establish safety limits or any safety
system settings. The applicable requirements of 10 CFR 50.47 and 10
CFR part 50, Appendix E will continue to be met.
Therefore, the proposed EAL scheme change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: David J. Wrona.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: March 27, 2017, as supplemented by
letter dated April 28, 2017. Publicly-available versions are in ADAMS
under Accession Nos. ML17086A364 and ML17118A092, respectively.
Description of amendment request: The amendment would amend the
Hope Creek Generating Station (Hope Creek) Technical Specifications
(TSs) to revise and relocate the pressure-temperature (P-T) limit
curves to a licensee-controlled pressure and temperature limits report
(PTLR). The request was submitted in accordance with guidance provided
in NRC Generic Letter 96-03, ``Relocation of the Pressure Temperature
Limit Curves and Low Temperature Overpressure Protections System
Limits,'' dated January 31, 1996.
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:
[[Page 23629]]
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed license amendment adopts the NRC approved
methodology described in Boiling Water Reactor Owner's Group (BWROG)
Licensing Topical Report (LTR) (BWROG-TP-11-022-A, SIR-05-044),
``Pressure Temperature Limits Report Methodology for Boiling Water
Reactors.'' The Hope Creek PTLR was developed based on the
methodology and template provided in the BWROG LTR.
10 CFR part 50, Appendix G establishes requirements to protect
the integrity of the reactor coolant pressure boundary (RCPB) in
nuclear power plants.
Implementing this NRC approved methodology does not reduce the
ability to protect the RCPB as specified in Appendix G, nor will
this change increase the probability of malfunction of plant
equipment, or the failure of plant structures, systems, or
components. Incorporation of the new methodology for calculating P-T
curves, and the relocation of the P-T curves from the TS to the PTLR
provides an equivalent level of assurance that the RCPB is capable
of performing its intended safety functions.
The proposed changes do not adversely affect accident initiators
or precursors, and do not alter the design assumptions, conditions,
or configuration of the plant or the manner in which the plant is
operated and maintained. The ability of structures, systems, and
components to perform their intended safety functions is not altered
or prevented by the proposed changes, and the assumptions used in
determining the radiological consequences of previously evaluated
accidents are not affected.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The change in methodology for calculating P-T limits and the
relocation of those limits to the PTLR do not alter or involve any
design basis accident initiators. RCPB integrity will continue to be
maintained in accordance with 10 CFR part 50, Appendix G, and the
assumed accident performance of plant structures, systems and
components will not be affected. The proposed changes do not involve
a physical alteration of the plant (i.e., no new or different type
of equipment will be installed), and the installed equipment is not
being operated in a new or different manner.
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 changes involve a significant reduction in
a margin of safety?
Response: No.
The proposed changes do not affect the function of the RCPB or
its response during plant transients. Calculating the Hope Creek P-T
limits using the NRC approved SI methodology ensures adequate
margins of safety relating to RCPB integrity are maintained. The
proposed changes do not alter the manner in which the Limiting
Conditions for Operation P-T limits for the RCPB are determined.
There are no changes to the setpoints at which protective actions
are initiated, and the operability requirements for equipment
assumed to operate for accident mitigation are not affected.
Therefore, it is concluded that the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: 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 County,
South Carolina
Date of amendment request: April 12, 2017. A publicly-available
version is in ADAMS under Accession No. ML17102B032.
Description of amendment request: The requested amendment proposes
changes to combined license (COL) Appendix C (and plant-specific Tier
1) and Updated Final Safety Analysis Report (UFSAR) Tier 2 that
describe: (1) The inspection and analysis of, and specifies the maximum
calculated flow resistance acceptance criteria for, the fourth-stage
automatic depressurization system loops; (2) revises licensing basis
text in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2
that describes the testing of, and specifies the allowable flow
resistance acceptance criteria for, the in-containment refueling water
storage tank (IRWST) injection line; (3) revises licensing basis text
in COL Appendix C (and plant-specific Tier 1) and UFSAR Tier 2 that
describes the testing of, and specifies the maximum flow resistance
acceptance criteria for, the containment recirculation line; (4)
revises licensing basis text in COL Appendix C (and plant-specific Tier
1) and UFSAR Tier 2 that specifies acceptance criteria for the maximum
flow resistance between the IRWST drain line and the containment; and
5) removes licensing basis text from UFSAR Tier 2 that discusses the
operation of swing check valves in current operating plants.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not adversely affect the operation of
any systems or equipment that initiate an analyzed accident or alter
any structures, systems, and components (SSCs) accident initiator or
initiating sequence of events. The proposed changes do not adversely
affect the physical design and operation of the in-containment
refueling water storage tank (IRWST) injection, drain, containment
recirculation, or fourth-stage automatic depressurization system
(ADS) valves, including as-installed inspections and maintenance
requirements as described in the Updated Final Safety Analysis
Report (UFSAR). Inadvertent operation or failure of the fourth-stage
ADS valves are considered as an accident initiator or part of an
initiating sequence of events for an accident previously evaluated.
However, the proposed change to the test methodology and calculated
flow resistance for the fourth-stage ADS lines does not adversely
affect the probability of inadvertent operation or failure.
Therefore, the probabilities of the accidents previously evaluated
in the UFSAR are not affected.
The proposed changes do not adversely affect the ability of
IRWST injection, drain, containment recirculation, and fourth-stage
ADS valves to perform their design functions. The designs of the
IRWST injection, drain, containment recirculation, and fourth-stage
ADS valves continue to meet the same regulatory acceptance criteria,
codes, and standards as required by the UFSAR. In addition, the
proposed changes maintain the capabilities of the IRWST injection,
drain, containment recirculation, and fourth-stage ADS valves to
mitigate the consequences of an accident and to meet the applicable
regulatory acceptance criteria.
The proposed changes do not adversely affect the prevention and
mitigation of other abnormal events, e.g., anticipated operational
occurrences, earthquakes, floods and turbine missiles, or their
safety or design analyses. Therefore, the consequences of the
accidents evaluated in the UFSAR are not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not affect the operation of any systems
or equipment that might initiate a new or different kind of
accident, or alter any SSC such that a new accident initiator or
initiating sequence of events is created. The proposed changes do
[[Page 23630]]
not adversely affect the physical design and operation of the IRWST
injection, drain, containment recirculation, and fourth-stage ADS
valves, including as-installed inspections, and maintenance
requirements, as described in the UFSAR. Therefore, the operation of
the IRWST injection, drain, containment recirculation, and fourth-
stage ADS valves is not adversely affected. These proposed changes
do not adversely affect any other SSC design functions or methods of
operation in a manner that results in a new failure mode,
malfunction, or sequence of events that affect safety-related or
nonsafety-related equipment. Therefore, this activity does not allow
for a new fission product release path, result in a new fission
product barrier failure mode, or create a new sequence of events
that result in significant fuel cladding failures.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes maintain existing safety margins. The
proposed changes verify and maintain the capabilities of the IRWST
injection, drain, containment recirculation, and fourth-stage ADS
valves to perform their design functions. The proposed changes
maintain existing safety margin through continued application of the
existing requirements of the UFSAR, while updating the acceptance
criteria for verifying the design features necessary to ensure the
IRWST injection, drain, containment recirculation, and fourth-stage
ADS valves perform the design functions required to meet the
existing safety margins in the safety analyses. Therefore, the
proposed changes satisfy the same design functions in accordance
with the same codes and standards as stated in the UFSAR.
These changes do not adversely affect any design code function,
design analysis, safety analysis input or result, or design/safety
margin.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
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.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the 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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3 (PVNGS), Maricopa County, Arizona
Date of amendment request: April 1, 2016, as supplemented by
letters dated July 21, September 9, and October 26, 2016.
Description of amendment request: The amendments revised the
Technical Specifications (TSs) for PVNGS, by modifying the requirements
regarding the degraded and loss of voltage relays that are planned to
be modified to be more aligned with designs generally implemented in
the industry. Specifically, the licensing basis for degraded voltage
protection will be changed from reliance on a TS initial condition that
ensures adequate post-trip voltage support of accident mitigation
equipment to crediting automatic actuation of the degraded and loss of
voltage relays to ensure proper equipment performance.
Date of issuance: April 27, 2017.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: Unit 1-201, Unit 2-201, and Unit 3-201. A publicly-
available version is in ADAMS under Accession No. ML17090A164;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74:
The amendment revised the Operating Licenses and TSs.
Date of initial notice in Federal Register: May 24, 2016 (81 FR
32803). The supplements dated July 21, September 9, and October 26,
2016, 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 amendments is contained
in a Safety Evaluation dated April 27, 2017.
No significant hazards consideration comments received: No.
Duke Energy Progress, Inc., Docket Nos. 50-325 and 50-324; Brunswick
Steam Electric Plant, Units 1 and 2 (BSEP), Brunswick County, North
Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2 (CNS), York County, South Carolina
Duke Energy Progress, Inc., Docket No. 50-400; Shearon Harris Nuclear
Power Plant, Unit 1 (HNP), Wake County, North Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2 (MNS), Mecklenburg County, North
Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3 (ONS), Oconee County, South
Carolina
Duke Energy Progress, Inc., Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2 (RNP), Darlington County, South Carolina
Date of amendment request: June 23, 2016.
Brief description of amendments: The amendments modified the
technical specification (TS) requirements for unavailable barriers by
adding Limiting
[[Page 23631]]
Condition for Operation (LCO) 3.0.9 to the TS for BSEP, ONS, and RNP.
The same changes were added as LCO 3.0.10 to the TS for CNS and MNS.
For HNP, TS requirements for unavailable barriers were modified by
adding LCO 3.0.6 to the TS. The changes are consistent with Technical
Specification Task Force Traveler (TSTF)-427, Revision 2, ``Allowance
for Non-Technical Specification Barrier Degradation on Supported System
OPERABILITY,'' subject to stated variations.
Date of issuance: April 26, 2017.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment Nos: 274/302 (BSEP), 288/284 (CNS), 155 (HNP), 295/274
(MNS), 402/404/403 (ONS), and 251 (RNP). A publicly-available version
is in ADAMS under Accession No. ML17066A374; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Facility Operating License Nos. DPR-71 and DPR-62 (BSEP), NPF-35
and NPF-52 (CNS), NPF-63 (HNP), NPF-9 and NPF-17 (MNS), DPR-38, DPR-47,
DPR-55 (ONS), and DPR-23 (RNP): Amendments revised the Facility
Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: August 16, 2016 (81 FR
54614).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 26, 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
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, Wake County, North Carolina
Duke Energy Progress, LLC, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: September 27, 2016, as supplemented by
letters dated November 22, 2016, and April 20, 2017.
Description of amendment request: The amendments revised Technical
Specification Surveillance Requirements to require operating
ventilation systems with charcoal filters for 15 continuous minutes
every 31 days or at a frequency controlled in accordance with the
Surveillance Frequency Control Program. The amendments are consistent
with NRC-approved Technical Specifications Task Force (TSTF) Traveler
TSTF-522, Revision 0, ``Revise Ventilation System Surveillance
Requirements to Operate for 10 hours per Month,'' as published in the
Federal Register on September 20, 2012 (77 FR 58428), with variations
due to plant-specific differences.
Date of issuance: May 8, 2017.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 275 (Unit 1) and 303 (Unit 2) for the Brunswick
Steam Electric Plant; 289 (Unit 1) and 285 (Unit 2) for the Catawba
Nuclear Station; 296 (Unit 1) and 275 (Unit 2) for the McGuire Nuclear
Station; 156 (Unit 1) for the Shearon Harris Nuclear Power Plant; and
252 (Unit No. 2) for the H. B. Robinson Steam Electric Plant. A
publicly-available version is in ADAMS under Accession No. ML17055A647;
documents related to these amendments are listed in the Safety
Evaluations enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-71 and DPR-62, for the
Brunswick Steam Electric Plant, Units 1 and 2; NPF-35 and NPF-52, for
the Catawba Nuclear Station, Units 1 and 2; NPF-9 and NPF-17, for the
McGuire Nuclear Station, Units 1 and 2; NPF-63, for the Shearon Harris
Nuclear Power Plant, Unit 1; and DPR-23, for the H. B. Robinson Steam
Electric Plant, Unit No. 2: The amendments revised the Renewed Facility
Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: January 17, 2017 (82 FR
4929). The supplemental letter dated April 20, 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 evaluations of the amendments are
contained in Safety Evaluations dated May 8, 2017.
No significant hazards consideration comments received: No.
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: February 10, 2016, as supplemented by
letters dated October 10 and December 16, 2016; and January 31,
February 7, February 16, and March 29, 2017.
Brief description of amendment: The amendment revised Technical
Specification 5.5.11, ``Primary Containment Leakage Rate Testing
Program,'' to increase the containment integrated leakage rate test
program Test A interval from 10 to 15 years.
Date of issuance: April 25, 2017.
Effective date: As of the date of issuance and shall be implemented
prior to the startup from the 2017 refueling outage.
Amendment No.: 193. A publicly-available version is in ADAMS under
Accession No. ML17103A235; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-22: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 26, 2016 (81 FR
24663). The supplemental letters dated October 10 and December 16,
2016; and January 31, February 7, February 16, and March 29, 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 April 25, 2017.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant (DCPP), Units 1 and 2, San Luis Obispo
County, California
Date of application for amendments: June 17, 2015, as supplemented
by letters dated August 31, October 22, November 2, November 6, and
December 17, 2015; and February 1, February 10, April 21, June 9,
September 15, October 6, and December 27, 2016.
Brief description of amendments: The amendments revised the
licensing bases to adopt the alternative source term (AST) as allowed
by 10 CFR 50.67, ``Accident source term.'' The AST
[[Page 23632]]
methodology, as established in NRC Regulatory Guide 1.183,
``Alternative Radiological Source Terms for Evaluating Design Basis
Accidents at Nuclear Power Reactors,'' July 2000 (ADAMS Accession No.
ML003716792), is used to calculate the offsite and control room
radiological consequences of postulated accidents for DCPP, Units 1 and
2. The amendments revised Technical Specification (TS) 1.1,
``Definitions,'' for the definition of Dose Equivalent I-131; TS
3.4.16, ``RCS [Reactor Coolant System] Specific Activity,'' to revise
the noble gas activity limit; TS 3.6.3, ``Containment Isolation
Valves,'' to require the 48-inch containment purge supply and exhaust
valves to be sealed closed during Modes 1, 2, 3, and 4; TS 5.5.11,
``Ventilation Filter Testing Program (VFTP),'' to change the allowable
methyl iodide penetration testing criteria for the auxiliary building
system charcoal filter; TS 5.5.19, ``Control Room Habitability
Program,'' to replace ``whole body or its equivalent to any part of the
body,'' with ``Total Effective Dose Equivalent,'' which is the dose
criteria specified in 10 CFR 50.67, and Appendix D, ``Additional
Conditions,'' for Facility Operating License Nos. DPR-80 and DPR-82 for
DCPP, Units 1 and 2, to add additional license conditions.
Date of issuance: April 27, 2017.
Effective date: As of its date of issuance and shall be implemented
within 365 days from the date of issuance.
Amendment Nos.: Unit 1-230; Unit 2-232. A publicly-available
version is in ADAMS under Accession No. ML17012A246; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: The license amendment
request was originally noticed in the Federal Register on October 13,
2015 (80 FR 61486). As a result of the supplemental letters dated
October 22, November 2, November 6, and December 17, 2015; and February
1, February 10, April 21, June 9, and September 15, 2016, the notice
was reissued in its entirety to include the revised scope, description
of the amendment request, and proposed no significant hazards
consideration determination on November 8, 2016 (81 FR 78664).
The supplemental letters dated October 6 and December 27, 2016,
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 amendments is contained
in a Safety Evaluation dated April 27, 2017.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: June 7, 2016.
Brief description of amendment: The amendment modified the
Technical Specifications to allow the use of Component Cooling System
(CCS) pump 2B-B to support Train 1B operability when the normally
aligned CCS pump C-S is removed from service.
Date of issuance: April 27, 2017.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 113. A publicly-available version is in ADAMS under
Accession No. ML17081A263; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-90: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 13, 2016 (81
FR 62932).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 27, 2017.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 11th day of May 2017.
For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2017-10570 Filed 5-22-17; 8:45 am]
BILLING CODE 7590-01-P