[Federal Register Volume 81, Number 206 (Tuesday, October 25, 2016)]
[Notices]
[Pages 73428-73447]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-25641]
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NUCLEAR REGULATORY COMMISSION
[NRC-2016-0214]
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 September 27, 2016, to October 7, 2016. The
last biweekly notice was published on October 11, 2016.
DATES: Comments must be filed by November 25, 2016. A request for a
hearing must be filed by December 27, 2016.
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-2016-0214. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Lynn Ronewicz, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-1927, email: [email protected].
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2016-0214, 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-2016-0214.
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-2016-0214, 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:/
/
[[Page 73429]]
www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
I. 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 a petition 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, which is available at the NRC's PDR, located at One White Flint
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. 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/. If a petition is filed within 60 days,
the Commission or a presiding officer designated by the Commission or
by the Chief Administrative Judge of the Atomic Safety and Licensing
Board Panel, will rule on the petition; and the Secretary or the Chief
Administrative Judge of the Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition shall set forth with
particularity the interest of the petitioner in the proceeding, and how
that interest may be affected by the results of the proceeding. The
petition should specifically explain the reasons why intervention
should be permitted with particular reference to the following general
requirements: (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. The petition
must also set forth the specific contentions which the petitioner seeks
to have litigated at 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 shall 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 those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion to support
its position on the issue. The petition must include sufficient
information to show that a genuine dispute exists with the applicant on
a material issue of law or fact. Contentions shall 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 these requirements 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,
and have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that person's admitted
contentions consistent with the NRC's regulations, policies, and
procedures.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day 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).
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 decide 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 held
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 any amendment unless the Commission finds an imminent
danger to the health
[[Page 73430]]
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 December 27, 2016. The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document, and should meet the requirements
for petitions set forth in this section, except that under 10 CFR
2.309(h)(2) a State, local governmental body, or Federally-recognized
Indian Tribe, or agency thereof does not need to address the standing
requirements in 10 CFR 2.309(d) if the facility is located within its
boundaries. A State, local governmental body, Federally-recognized
Indian Tribe, or agency thereof may also have the opportunity to
participate under 10 CFR 2.315(c).
If a hearing is granted, any person who does not wish, or is not
qualified, to become a party to the proceeding may, in 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 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, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene (hereinafter
``petition''), and documents filed by interested governmental entities
participating 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. 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 request (1) a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition (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. System requirements for accessing
the E-Submittal server are available on the NRC's public Web site at
http://www.nrc.gov/site-help/e-submittals/adjudicatory-sub.html.
Participants may attempt to use other software not listed on the Web
site, but should note that the NRC's E-Filing system does not support
unlisted software, and the NRC Electronic Filing Help Desk will not be
able to offer assistance in using unlisted software.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a petition.
Submissions should 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 documents are 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 documents on those
participants separately. Therefore, applicants and other participants
(or their counsel or representative) must apply for and receive a
digital ID certificate before a hearing petition to intervene is filed
so that they can obtain access to the document via the E-Filing system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC 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 7 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, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document 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
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal
[[Page 73431]]
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings, unless an NRC regulation or
other law requires submission of such information. However, in some
instances, a petition will require including information on local
residence in order to demonstrate a proximity assertion of interest in
the proceeding. 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.
The Commission will issue a notice or order granting or denying a
hearing request or intervention petition, designating the issues for
any hearing that will be held and designating the Presiding Officer. A
notice granting a hearing will be published in the Federal Register and
served on the parties to the hearing.
For further details with respect to these license amendment
applications, see the application for amendment, which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Dominion Energy Kewanee, Inc. (DEK), Docket No. 50-305, Kewanee Power
Station (KPS), Carlton, Wisconsin
Date of amendment request: September 14, 2015. A publicly available
version is in ADAMS under Accession No. ML15261A238.
Description of amendment request: The amendment would revise the
KPS Permanently Defueled Emergency Plan (PDEP) and the Permanently
Defueled Emergency Action Levels (EAL) Bases Document. DEK requests
revisions of the PDEP and the EAL Bases Document that reflect DEK's
plan to transfer all spent fuel to the independent spent fuel storage
installation (ISFSI).
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 amendment would modify the KPS renewed facility
operating license by revising the emergency plan and revising the
EAL scheme. KPS has permanently ceased operation and is permanently
defueled. The proposed amendment is conditioned on all spent nuclear
fuel being removed from wet storage in the spent fuel pool and
placed in dry storage within the ISFSI. Occurrence of postulated
accidents associated with spent fuel stored in a spent fuel pool is
no longer credible in a spent fuel pool devoid of such fuel. The
proposed amendment has no effect on plant systems, structures, and
components (SSCs) and no effect on the capability of any plant SSC
to perform its design function. The proposed amendment would not
increase the likelihood of the malfunction of any plant SSC. The
proposed amendment would have no effect on any of the previously
evaluated accidents in the KPS Updated Safety Analysis Report
(USAR).
Since KPS has permanently ceased operation, the generation of
fission products has ceased and the remaining source term continues
to decay. This continues to significantly reduce the consequences of
previously postulated accidents.
Therefore, the proposed amendment does not involve a significant
increase in the consequences of a previously evaluated accident.
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 amendment constitutes a revision of the emergency
planning function commensurate with the ongoing and anticipated
reduction in radiological source term at KPS.
The proposed amendment does not involve a physical alteration of
the plant. No new or different types of equipment will be installed
and there are no physical modifications to existing equipment as a
result of the proposed amendment.
Similarly, the proposed amendment would not physically change
any SSCs involved in the mitigation of any postulated accidents.
Thus, no new initiators or precursors of a new or different kind of
accident are created. Furthermore, the proposed amendment does not
create the possibility of a new failure mode associated with any
equipment or personnel failures. The credible events for the ISFSI
remain unchanged.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
Because the 10 CFR part 50 license for KPS no longer authorizes
operation of the reactor or emplacement or retention of fuel into
the reactor vessel, as specified in 10 CFR 50.82(a)(2), the
occurrence of postulated accidents associated with reactor operation
is no longer credible. With all nuclear spent fuel pool transferred
out of wet storage from the spent fuel pool and placed in dry
storage within the ISFSI, a fuel handling accident is no longer
credible. There are no longer credible events that would result in
any releases beyond the site boundary exceeding the EPA PAG
[Environmental Protection Agency protective action guideline]
exposure levels, as detailed in the EPA's ``Protective Action Guide
and Planning Guidance for Radiological Incidents,'' Draft for
Interim Use and Public Comment dated March 2013 (PAG Manual).
The proposed amendment does not involve a change in the plant's
design, configuration, or operation. The proposed amendment does not
affect either the way in which the plant structures, systems, and
components perform their safety function or their design margins.
Because there is no change to the physical design of the plant,
there is no change to any of these margins.
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: Lillian M. Cuoco, Senior Counsel, Dominion
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
NRC Branch Chief: Bruce A. Watson.
Dominion Energy Kewanee, Inc. (DEK), Docket No. 50-305, Kewanee Power
Station (KPS), Carlton, Wisconsin
Date of amendment request: July 28, 2016. A publicly available
version is in ADAMS under Accession No. ML16216A187.
Description of amendment request: The amendment would revise the
KPS Updated Safety Analysis Report (USAR) Section 9.5.2.2.4,
``Auxiliary Building Crane,'' to: (1) Add a description of a non-single
failure proof intermediate lifting device that DEK intends to use
during a specific spent fuel cask handling activity in the auxiliary
building, and (2) incorporate a new load drop analysis applicable to
the use of this intermediate lifting device. The amendment also
includes (for information) a new Technical Requirements Manual section
that governs the use of the non-single failure proof intermediate
lifting device to ensure compliance with the required parameters in the
load drop analysis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
[[Page 73432]]
The probability of a heavy load drop onto fuel is unchanged by
this amendment since the intermediate lift device is not used for
handling loaded or unloaded spent fuel canisters in or around the
spent fuel pool. Heavy load lifts in and around the spent fuel pool
will continue to be performed per the current licensing basis.
The proposed amendment has no effect on the capability of any
plant systems, structures, and components (SSCs) to perform their
design functions. The spent fuel pool is unaffected by the proposed
amendment. The design function of the auxiliary building crane is
not changed. Other lifting devices and interfacing lifting points
associated with spent fuel cask handling are designed in accordance
with applicable NRC guidance pertaining to single failure proof
lifting systems. Therefore, the proposed amendment would not
increase the likelihood of the malfunction of any plant SSC. The
proposed amendment would have no effect on any of the previously
evaluated accidents in the KPS USAR.
Therefore, the proposed amendment does not involve a significant
increase in the consequences of a previously evaluated accident.
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 amendment does not affect cask handling activities
in or around the KPS spent fuel pool. Drops of heavy loads will
continue to be very improbable events. Use of a different type of
equipment to lift spent fuel canisters does not involve any new or
different kind of accident.
The proposed amendment does not involve a physical alteration of
the plant. Similarly, the proposed amendment would not physically
change any SSCs involved in the mitigation of any postulated
accidents. The physical structure of the spent fuel canisters is not
altered by this amendment.
The possibility of a heavy load drop onto fuel remains non-
credible since the intermediate lift device is not used to handle
spent fuel canisters in or around the spent fuel pool. Heavy load
lifts in and around the spent fuel pool will continue to be
performed per the current licensing basis. The proposed amendment
does not impact safe shutdown equipment. The spent fuel pool,
including its cooling and inventory makeup, is unaffected by the
proposed amendment.
The current licensing basis (USAR Section 14.2.1) includes
evaluations of the consequences of a fuel handling accident
involving failure of fuel cladding. Postulation of a canister load
drop creates the possibility of a new initiator of this previously
evaluated accident (failure of fuel cladding) caused by the
postulated non-mechanistic single failure of the intermediate lift
device. The analysis concludes that the postulated drop of a
canister loaded with fuel assemblies would not result in failure of
canister integrity (and therefore there would be no radiological
release). The consequences of a canister drop are bounded by the
current licensing scenario of a fuel handling accident.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
Heavy load handling will continue to be conducted in accordance
with NRC approved methods. Analysis of a postulated load drop of a
loaded spent fuel canister demonstrates satisfactory outcomes.
The proposed amendment does not involve a change in the plant's
design, configuration, or operation. The proposed amendment does not
significantly affect either the way in which the plant structures,
systems, and components perform their safety function or their
design margins. Because there is no change to the physical design of
the plant, there is likewise no significant change to any of these
margins.
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: Lillian M. Cuoco, Senior Counsel, Dominion
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
NRC Branch Chief: Bruce A. Watson.
Duke Energy Florida, Inc. (DEF), et al., Docket No. 50-302, Crystal
River Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida
Date of amendment request: August 31, 2016. A publicly available
version is in ADAMS under Accession No. ML16243A259.
Description of amendment request: The amendment would revise the
Operating License and associated Permanently Defueled Technical
Specifications (PDTS) to reflect removal of all CR-3 spent nuclear fuel
from the spent fuel pools and its transfer to dry cask storage within
the onsite Independent Spent Fuel Storage Installation (ISFSI).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would modify the CR-3 facility operating
license and PDTS by deleting the portions of the license and PDTS
that are no longer applicable to a facility with no spent nuclear
fuel stored in the spent fuel pools, while modifying the remaining
portions to correspond to all nuclear fuel stored within an ISFSI.
This amendment will be implemented within 60 days following DEF's
submittal of written notification to the NRC that all spent fuel
assemblies have been transferred out of the spent fuel pools and
placed in dry storage within the ISFSI.
The definition of safety-related structures, systems, and
components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are
those relied on to remain functional during and following design
basis events to assure:
1. The integrity of the reactor coolant boundary;
2. The capability to shutdown the reactor and maintain it in a
safe shutdown condition; or
3. The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures
comparable to the applicable guideline exposures set forth in 10 CFR
50.43(a)(1) or 100.11.
The first two criteria (integrity of the reactor coolant
pressure boundary and safe shutdown of the reactor) are not
applicable to a plant in a permanently defueled condition. The third
criterion is related to preventing or mitigating the consequences of
accidents that could result in potential offsite exposures exceeding
limits. However, after all nuclear spent fuel assemblies have been
transferred to dry cask storage within an ISFSI, none of the SSCs at
CR-3 are required to be relied on for accident mitigation.
Therefore, none of the SSCs at CR-3 meet the definition of a safety-
related SSC stated in 10 CFR 50.2. The proposed deletion of
requirements in the PDTS does not affect systems credited in any
accident analysis at CR-3.
Section 14 of the CR-3 Final Safety Analysis Report (FSAR)
described the design basis accidents (DBAs) related to the spent
fuel pools. These postulated accidents are predicated on spent fuel
being stored in the spent fuel pools. With the removal of the spent
fuel from the spent fuel pools, there are no remaining spent fuel
assemblies to be monitored and there are no credible accidents that
require the actions of a Certified Fuel Handler, Shift Manager, or a
Non-certified Operator to prevent occurrence or mitigate the
consequences of an accident.
The proposed changes do not have an adverse impact on the
remaining decommissioning activities or any of their postulated
consequences.
The proposed changes related to the relocation of certain
administrative requirements do not affect operating procedures or
administrative controls that have the function of preventing or
mitigating any accidents applicable to the safe management of
irradiated fuel or decommissioning of the facility.
Therefore, the proposed amendment 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?
[[Page 73433]]
Response: No.
The proposed changes eliminate the operational requirements and
certain design requirements associated with the storage of the spent
fuel in the spent fuel pools, and relocate certain administrative
controls to the Quality Assurance Program Description or other
licensee controlled document.
After the removal of the spent fuel from the spent fuel pools
and transfer to the ISFSI, there are no spent fuel assemblies that
remain in the spent fuel pools. Coupled with a prohibition against
storage of fuel in the spent fuel pools, the potential for fuel
related accidents is removed. The proposed changes do not introduce
any new failure modes.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The removal of all spent nuclear fuel from the spent fuel pools
into storage in casks within an ISFSI, coupled with a prohibition
against future storage of fuel within the spent fuel pools, removes
the potential for fuel related accidents.
The design basis and accident assumptions within the CR-3 FSAR
and the PDTS relating to safe management and safety of spent fuel in
the spent fuel pools are no longer applicable. The proposed changes
do not affect remaining plant operations, systems, or components
supporting decommissioning activities.
The requirements for systems, structures, and components (SSCs)
that have been removed from the CR-3 PDTS are not credited in the
existing accident analysis for any applicable postulated accident;
and as such, do not contribute to the margin of safety associated
with the accident analysis.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, 550 South Tryon Street,
Charlotte, NC 28202.
NRC Branch Chief: Bruce A. Watson.
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant (HNP), Unit 1, Wake and Chatham Counties, North Carolina
Date of amendment request: June 29, 2016. A publicly-available
version is in ADAMS under Accession No. ML16182A387.
Description of amendment request: The amendment would revise HNP
Technical Specifications (TSs) to (1) delete the Gaseous Radwaste
Treatment System definition from TSs, (2) relocate Explosive Gas
Mixture TS requirements and Liquid Holdup Tanks TS requirements to a
licensee-controlled program in the Procedures and Programs TSs section,
and (3) modify the Gas Storage Tank Radioactivity Monitoring Program
TSs into an Explosive Gas and Storage Tank Radioactivity Monitoring
Program to include controls for potentially explosive gas mixtures and
the quantity of radioactivity contained in unprotected outdoor liquid
storage tanks.
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 do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes are administrative in nature and alter only the
format and location of programmatic controls and procedural details
relative to explosive gas monitoring and liquid holdup tanks.
Existing TS containing procedural details are being relocated to
licensee control. Compliance with applicable regulatory requirements
will continue to be maintained. In addition, the proposed changes do
not alter the conditions or assumptions in any of the previous
accident analyses. Because the previous accident analyses remain
bounding, the radiological consequences previously evaluated are not
adversely affected by the proposed changes.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve any change to the configuration
or method of operation of any plant equipment. Accordingly, no new
failure modes have been defined for any plant system or component
important to safety nor has any new limiting single failure been
identified as a result of the proposed changes. Also, there will be
no change in types or increase in the amounts of any effluents
released offsite.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not involve a significant reduction in a
margin of safety and are considered administrative in nature. The
proposed changes do not involve any actual change in the methodology
used in the monitoring of explosive gas mixtures contained in the
Gaseous Waste Processing System. HNP does not currently utilize
unprotected outdoor liquid storage tanks; therefore, there are no
associated methodology changes with this request. These changes
provide for the relocation of procedural details outside of the
technical specifications with the addition of appropriate
administrative controls to provide continued assurance of compliance
to applicable regulatory requirements. 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: Lara Nichols, Deputy General Counsel, Duke
Energy Corporation, 550 South Tryon St., M/C DEC45A, Charlotte, NC
28202.
NRC Acting Branch Chief: Jeanne A. Dion.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: July 12, 2016. A publicly-available
version is in ADAMS under Accession No. ML16194A515.
Description of amendment request: The amendment would reduce the
minimum reactor dome pressure associated with the critical power
correlation from 785 pounds per square inch gauge (psig) to 685 psig in
Technical Specification (TS) 2.1.1, ``Reactor Core SLs [Safety
Limits],'' and associated bases.
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 change does not involve a modification of any plant
hardware; the probability and consequence of the Pressure Regulator
Failure Open (PRFO) transient are essentially unchanged. The
reduction in the reactor dome pressure safety limit (SL) from 785
psig to 685 psig provides greater margin to accommodate the pressure
reduction
[[Page 73434]]
during the transient within the revised TS limit.
The proposed change will continue to support the validity range
for the correlations and the calculation of Minimum Core Power Ratio
(MCPR) as approved. The proposed TS revision involves no significant
changes to the operation of any systems or components in normal,
accident or transient operating conditions.
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 previously evaluated?
Response: No.
The proposed reduction in the reactor dome pressure SL from 785
psig to 685 psig is a change based upon previously approved
documents and does not involve changes to the plant hardware or its
operating characteristics. As a result, no new failure modes are
being introduced.
Therefore, the change does not introduce a new or different kind
of accident from those previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through the design of the
plant structures, systems, and components, and through the
parameters for safe operation and setpoints for the actuation of
equipment relied upon to respond to transients and design basis
accidents. The proposed change in reactor dome pressure enhances the
safety margin, which protects the fuel cladding integrity during a
depressurization transient, but does not change the requirements
governing operation or availability of safety equipment assumed to
operate to preserve the margin of safety. The change does not alter
the behavior of plant equipment, which remains unchanged. The
available pressure range is expanded by the change, thus offering
greater margin for pressure reduction during the transient.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: Robert J. Pascarelli.
Entergy Nuclear Operations, Inc. (ENO), Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant (JAF), Oswego County, New York
Date of amendment request: January 15, 2016, as supplemented by
letters dated June 3, 2016, and September 19, 2016. Publicly available
versions are available in ADAMS under Accession Nos. ML16015A456,
ML16155A326, and ML16263A237, respectively.
Description of amendment request: The licensee has provided a
formal notification to the NRC of the intention to permanently cease
power operations of JAF at the end of the current operating cycle. Once
certifications for permanent cessation of operation and permanent
removal of fuel from the reactor are submitted to the NRC, certain
staffing and training Technical Specifications (TSs) administrative
controls will no longer be applicable or appropriate for the
permanently defueled condition. Therefore, ENO is requesting approval
of changes to the staffing and training requirements in Section 5.0,
``Administrative Controls,'' of the JAF TSs. Specifically, the
amendment would revise and remove certain requirements in TS Sections
5.1, ``Responsibility''; 5.2, ``Organization''; and 5.3, ``Plant Staff
Qualifications,'' and add additional definitions to TS Section 1.1,
``Definitions.'' The proposed amendment would not be effective until
the certification of permanent cessation of operation and certification
of permanent removal of fuel from the reactor vessel are submitted to
the NRC.
The license amendment request was originally noticed in the Federal
Register on March 1, 2016 (81 FR 10678). The notice is being reissued
in its entirety to include the revised scope and description of the
amendment request.
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, with NRC staff revisions provided in [brackets], which
is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would not take effect until JAF has
permanently ceased operation and entered a permanently defueled
condition. The proposed amendment would modify the JAF TS by
deleting the portions of the TS that are no longer applicable to a
permanently defueled facility, while modifying the other sections to
correspond to the permanently defueled condition.
The deletion and modification of provisions of the
administrative controls do not directly affect the design of
structures, systems, and components (SSCs) necessary for safe
storage of irradiated fuel or the methods used for handling and
storage of such fuel in the fuel pool. The changes to the
administrative controls are administrative in nature and do not
affect any accidents applicable to the safe management of irradiated
fuel or the permanently shutdown and defueled condition of the
reactor.
In a permanently defueled condition, the only credible accident
is the fuel handling accident.
The probability of occurrence of previously evaluated accidents
is not increased, since extended operation in a defueled condition
will be the only operation allowed, and therefore bounded by the
existing analyses. Additionally, the occurrence of postulated
accidents associated with reactor operation is no longer credible in
a permanently defueled reactor. This significantly reduces the scope
of applicable accidents.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes have no impact on facility SSCs affecting
the safe storage of irradiated fuel, or on the methods of operation
of such SSCs, or on the handling and storage of irradiated fuel
itself. The administrative removal of or modifications of the TS
that are related only to administration of facility cannot result in
different or more adverse failure modes or accidents than previously
evaluated because the reactor will be permanently shutdown and
defueled and JAF will no longer be authorized to operate the
reactor.
The proposed deletion of requirements of the JAF TS do not
affect systems credited in the accident analysis for the fuel
handling accident at JAF. The proposed TS will continue to require
proper control and monitoring of safety significant parameters and
activities.
The proposed amendment does not result in any new mechanisms
that could initiate damage to the remaining relevant safety barriers
for defueled plants (fuel cladding and spent fuel cooling). Since
extended operation in a defueled condition will be the only
operation allowed, and therefore bounded by the existing analyses,
such a condition does not create the possibility of a new or
different kind of accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Because the 10 CFR part 50 license for JAF will no longer
authorize operation of the reactor or emplacement or retention of
fuel into the reactor vessel once the certifications required by 10
CFR 50.82(a)(1) are submitted, as specified in 10 CFR 50.82(a)(2),
the occurrence of postulated accidents associated with reactor
operation is no longer credible. The only remaining credible
accident is a fuel handling accident (FHA). The proposed
[[Page 73435]]
amendment does not adversely affect the inputs or assumptions of any
of the design basis analyses that impact the FHA.
The proposed changes are limited to those portions of the OL
[operating license] and TS that are not related to the safe storage
of irradiated fuel. The requirements that are proposed to be revised
or deleted from the JAF OL and TS are not credited in the existing
accident analysis for the remaining applicable as such, do not
contribute to the margin of safety associated with the accident
analysis. Postulated DBAs [design-basis accidents] involving the
reactor are no longer possible because the reactor will be
permanently shutdown and defueled and JAF will no longer be
authorized to operate the reactor.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: July 26, 2016. A publicly-available
version is in ADAMS under Accession No. ML16210A227.
Description of amendment request: The amendments would revise
technical specification (TS) requirements relating to: (1) The
inservice inspection (ISI) program required by the American Society of
Mechanical Engineers (ASME) Boiler and Pressure Code (Code), and (2)
the inservice testing (IST) program required by the ASME Code for
Operation and Maintenance of Nuclear Power Plants (OM Code). The
proposed changes are based, in part, on Technical Specifications Task
Force (TSTF) Traveler TSTF-545, Revision 3, ``TS Inservice Testing
Program Removal & Clarify SR Usage Rule Application to Section 5.5
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 with NRC staff edits in square
brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS 4.0.5, Surveillance Requirements
for inservice inspection and testing of ASME Code Class 1, 2 & 3
components, by revising the Inservice Testing Program and Inservice
Inspection Program specification.
Most requirements in the IST Program are removed, as they are
duplicative of requirements in the ASME OM Code, as clarified by
Code Case OMN-20, ``Inservice Test Frequency.'' The remaining
requirements in the TS Section 4.0.5, IST Program are eliminated
because the NRC has determined their inclusion in the TS is contrary
to regulations. A new defined term, ``Inservice Testing Program,''
is added to the TS, which references the requirements of 10 CFR
50.55a(f).
Similarly, the requirements in the ISI Program are revised, as
they are [ ] duplicative of requirements in Section XI of the ASME
Boiler and Pressure Vessel Code and applicable Addenda.
Performance of inservice testing or inservice inspection is not
an initiator to any accident previously evaluated. As a result, the
probability of occurrence of an accident is not significantly
affected by the proposed change. Inservice test frequencies under
Code Case OMN-20 are equivalent to the current testing period
allowed by the TS with the exception that testing frequencies
greater than two years may be extended by up to six months to
facilitate test scheduling and consideration of plant operating
conditions that may not be suitable for performance of the required
testing. The testing frequency extension will not affect the ability
of the components to mitigate any accident previously evaluated as
the components are required to be operable during the testing period
extension. Performance of inservice tests utilizing the allowances
in OMN-20 will not significantly affect the reliability of the
tested components. As a result, the availability of the affected
components, as well as their ability to mitigate the consequences of
accidents previously evaluated, is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not alter the design or configuration
of the plant. The proposed change does not involve a physical
alteration of the plant; no new or different kind of equipment will
be installed. The proposed change does not alter the types of
inservice testing or inservice inspection performed. In most cases,
the frequency of inservice testing and inservice inspection is
unchanged. However, the frequency of testing or inspection would not
result in a new or different kind of accident from any previously
evaluated since the testing methods are not altered.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change eliminates some provisions from the TS in
lieu of provisions in the ASME Code, as modified by use of Code Case
OMN-20 (IST) or ASME Boiler and Pressure Vessel Code (ISI).
Compliance with the ASME Code is required by 10 CFR 50.55a. The
proposed change also allows inservice tests with frequencies greater
than two years to be extended by six months to facilitate test
scheduling and consideration of plant operating conditions that may
not be suitable for performance of the required testing. The testing
frequency extension will not affect the ability of the components to
respond to an accident as the components are required to be operable
during the testing period extension. The proposed change will
eliminate the existing TS SR 4.0.2 allowance to perform a specified
surveillance time interval with a maximum allowable extension not to
exceed 25% of the surveillance interval, unless there is a specific
SR referencing usage of the INSERVICE TESTING PROGRAM and TS SR
4.0.3 allowance to defer performance of missed inservice tests up to
the duration of the specified testing frequency, and instead will
require an assessment of the missed test on equipment operability.
This assessment will consider the effect on a margin of safety
(equipment operability). Should the component be inoperable, the
Technical Specifications provide actions to ensure that the margin
of safety is protected. The proposed change also eliminates a
statement that nothing in the ASME Code should be construed to
supersede the requirements of any TS. However, elimination of the
statement will have no effect on plant operation or 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: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: July 15, 2016. A publicly-available
version is in ADAMS under Package Accession No. ML16201A306.
Description of amendment request: The amendment would revise the
[[Page 73436]]
Radiological Emergency Plan Annex for TMI-1. The proposed changes would
decrease the radiation protection technician staffing from three to two
technicians, remove two maintenance technicians currently assigned to
the repair and corrective action function, and eliminate the on-shift
Operations Support Center director position.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the TMI Emergency Plan do not increase
the probability or consequences of an accident. The proposed changes
do not impact the function of plant Structures, Systems, or
Components (SSCs). The proposed changes do not affect accident
initiators or accident precursors, nor do the changes alter design
assumptions. The proposed changes do not alter or prevent the
ability of the onsite ERO [emergency response organization] to
perform their intended functions to mitigate the consequences of an
accident or event. The proposed changes remove onsite ERO positions
no longer credited or considered necessary in support of Emergency
Plan implementation.
Therefore, the proposed changes to the Emergency Plan 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 previously evaluated?
Response: No.
The proposed changes have no impact on the design, function, or
operation of any plant SSCs. The proposed changes do not affect
plant equipment or accident analyses. The proposed changes do not
involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed), a change in the
method of plant operation, or new operator actions. The proposed
changes do not introduce failure modes that could result in a new
accident, and the proposed changes do not alter assumptions made in
the safety analysis. The proposed changes remove onsite ERO
positions no longer credited or considered necessary in support of
Emergency Plan implementation.
Therefore, the proposed changes to the Emergency Plan do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation dose to the public.
The proposed changes do not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analyses. There are no changes being made to
safety analysis assumptions, safety limits, or limiting safety
system settings that would adversely affect plant safety as a result
of the proposed changes. Margins of safety are unaffected by the
proposed changes to the ERO minimum on-shift staffing.
The proposed changes are associated with the Emergency Plan
staffing and do not impact operation of the plant or its response to
transients or accidents. The proposed changes do not affect the
Technical Specifications. The proposed changes do not involve a
change in the method of plant operation, and no accident analyses
will be affected by the proposed changes. Safety analysis acceptance
criteria are not affected by these proposed changes. The proposed
changes to the Emergency Plan will continue to provide the necessary
onsite ERO response staff.
Therefore, the proposed changes to the Emergency Plan do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: Douglas A. Broaddus.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue
County, Minnesota
Date of amendment request: August 31, 2016. A publicly-available
version is in ADAMS under Accession No. ML16244A493.
Brief description of amendment request: The amendments would revise
the Required Actions and associated Completion Times to Technical
Specification (TS) 3.8.7, ``Inverters--Operating.'' Specifically,
Condition B would be deleted and current Condition C would be re-
lettered to Condition B. Additionally, the Required Actions and
associated Completion Times for Condition A would be modified to
require restoration of one inoperable inverter to operability within 24
hours. These changes conform to Improved Standard Technical
Specification TS 3.8.7.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the TS for the purpose of
eliminating a non-conservative Required Action. The proposed TS
change does not introduce new equipment or new equipment operating
modes, nor does the proposed change alter existing system
relationships. The proposed change does not affect normal plant
operation. Further, the proposed change does not increase the
likelihood of the malfunction of any SSC [structure, system and
component] or impact any analyzed accident. Consequently, the
probability of an accident previously evaluated is not affected and
there is no significant increase in the consequences of any 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 revises the TS for the purpose of
eliminating a non-conservative Required Action. The change does not
involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operations. The proposed change does
not alter assumptions made in the safety analysis. Further, the
proposed change does not introduce new accident initiators.
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 a
margin of safety?
Response: No.
The proposed change revises the TS for the purpose of
eliminating a non-conservative Required Action. The proposed change
does not alter the manner in which safety limits, limiting safety
system settings, or limiting conditions for operation are
determined. The safety analysis assumptions and acceptance criteria
are not affected by this change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy
[[Page 73437]]
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: July 20, 2016. A publicly-available
version is in ADAMS under Accession No. ML16203A006.
Description of amendment request: The amendment would revise the
Hope Creek Generating Station (Hope Creek) Technical Specifications
(TS), Section 6.8.4.i, ``Inservice Testing Program,'' to remove
requirements duplicated in the American Society of Mechanical Engineers
(ASME) Code for Operations and Maintenance of Nuclear Power Plants Case
OMN-20, ``Inservice Test Frequency.'' A new defined term, ``Inservice
Testing Program,'' will be added to the TS 1.0, ``Definitions,''
section. The licensee stated that the proposed change to the TS is
consistent with Technical Specifications Task Force (TSTF) Traveler
TSTF-545, Revision 3, ``TS Inservice Testing Program Removal & Clarity
SR Usage Rule Application to Section 5.5 Testing'' (ADAMS Accession No.
ML15294A555), with no proposed variations or deviations. However, the
Hope Creek TS uses different numbering for surveillance requirements
than the Standard Technical Specifications on which TSTF-545 was based,
so the licensee changed the TSTF-545 numbering to be consistent with
the Hope Creek TS numbering.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with NRC staff edits in square
brackets:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises TS Chapter 6, ``Administrative
Controls,'' Section 6.8, ``Procedures and Programs,'' by eliminating
the ``Inservice Testing Program'' specification. Most requirements
in the Inservice Testing Program are removed, as they are
duplicative of requirements in the ASME OM Code, as clarified by
Code Case OMN-20, ``Inservice Test Frequency.'' The remaining
requirements in the Section 6.8 IST Program are eliminated [. . .].
A new defined term, ``Inservice Testing Program,'' is added to the
TS, which references the requirements of 10 CFR 50.55a(f).
Performance of inservice testing is not an initiator to any
accident previously evaluated. As a result, the probability of
occurrence of an accident is not significantly affected by the
proposed change. Inservice test frequencies under Code Case OMN-20
are equivalent to the current testing period allowed by the TS with
the exception that testing frequencies greater than 2 years may be
extended by up to 6 months to facilitate test scheduling and
consideration of plant operating conditions that may not be suitable
for performance of the required testing. The testing frequency
extension will not affect the ability of the components to mitigate
any accident previously evaluated as the components are required to
be operable during the testing period extension. Performance of
inservice tests utilizing the allowances in OMN-20 will not
significantly affect the reliability of the tested components. As a
result, the availability of the affected components, as well as
their ability to mitigate the consequences of accidents previously
evaluated, is not affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the design or configuration
of the plant. The proposed change does not involve a physical
alteration of the plant; no new or different kind of equipment will
be installed. The proposed change does not alter the types of
inservice testing performed. In most cases, the frequency of
inservice testing is unchanged. However, the frequency of testing
would not result in a new or different kind of accident from any
previously evaluated since the testing methods are not altered.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change eliminates some requirements from the TS in
lieu of requirements in the ASME Code, as modified by use of Code
Case OMN-20. Compliance with the ASME Code is required by 10 CFR
50.55a. The proposed change also allows inservice tests with
frequencies greater than 2 years to be extended by 6 months to
facilitate test scheduling and consideration of plant operating
conditions that may not be suitable for performance of the required
testing. The testing frequency extension will not affect the ability
of the components to respond to an accident as the components are
required to be operable during the testing period extension. The
proposed change will eliminate the existing TS 4.0.3 allowance to
defer performance of missed inservice tests up to the duration of
the specified testing frequency, and instead will require an
assessment of the missed test on equipment operability. This
assessment will consider the effect on a margin of safety (equipment
operability). Should the component be inoperable, the TS provide
actions to ensure that the margin of safety is protected. The
proposed change also eliminates a statement that nothing in the ASME
Code should be construed to supersede the requirements of any TS. [.
. .] However, elimination of the statement will have no effect on
plant operation or 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: Jeffrie J. Keenan, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Douglas A. Broaddus.
South Carolina Electric & Gas Company and South Carolina Public Service
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear
Station, Units 2 and 3, Fairfield, South Carolina
Date of amendment request: September 15, 2016. A publicly-available
version is in ADAMS under Accession No. ML16259A310.
Description of amendment request: The amendments would revise
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer
Nuclear Station, Units 2 and 3. The amendments propose changes to the
Updated Final Safety Analysis Report (UFSAR) in the form of departures
from the incorporated plant-specific Design Control Document Tier 2*
information. Specifically, the proposed changes would revise the
Combined Licenses to clarify information in WCAP-17179,
``AP1000[supreg] Component Interface Module Technical Report,'' which
demonstrates design compliance with licensing bases requirements. WCAP-
17179 is incorporated by reference into the UFSAR to provide additional
details regarding the component interface module (CIM) system design.
The requested amendments also propose a change to the CIM internal
power supply that will enable proper functioning of the field
programmable gate arrays.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
[[Page 73438]]
Response: No.
The proposed change to the CIM internal power supply enables the
field programmable gate array (FPGA) to function properly. The
proposed change to the FPGA core power has no adverse effect on the
operation of the output actuation relays. The function of the
internal power supply has no input to plant safety analysis. The
change to the CIM internal power supply has a negligible effect on
the 24 Vdc [volts direct current] supplies and ultimately the plant
electrical system load and has no adverse effect on the CIM
functionality.
The proposed changes to clarify how licensing basis design
documentation reflects compliance with license basis requirements,
and the proposed change to the ownership of safety remote node
controller (SRNC) and CIM intellectual property, are not technical
changes. The proposed changes do not affect any accident initiator
in the UFSAR, or affect the radioactive material releases in the
UFSAR accident analyses. The proposed change does not alter the
ability of the facility to prevent and mitigate abnormal events,
e.g., accidents, anticipated operational occurrences, earthquakes,
floods and turbine missiles, or their safety or design analyses. No
safety-related structure, system, or component (SSC) or function is
adversely affected. The change does not involve or interface with
any SSC accident initiator or initiating sequence of events, and
thus, the probabilities of the accidents evaluated in the UFSAR are
not affected. This activity does not involve a new fission product
release path, nor a new fission product barrier failure mode, nor
create a new sequence of events that would result in significant
fuel cladding failures. Because the proposed changes do not change
any safety-related SSC or function credited in the mitigation of an
accident, the consequences of the accidents evaluated in the UFSAR
are not affected.
Therefore, the requested amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change to the CIM internal power supply enables the
FPGA to function properly and does not involve accident initiators.
The change to the CIM internal power supply has a negligible effect
on the 24 Vdc supplies and ultimately the plant electrical system
load and has no adverse effect on CIM functionality.
The proposed clarified descriptions and the proposed change to
the ownership of SRNC and CIM intellectual property are not
technical changes. The proposed changes do not affect other plant
equipment or adversely affect the design of the CIM. Therefore, the
proposed changes do not affect any safety-related equipment itself,
nor do they affect equipment whose failure could initiate an
accident or a failure of a fission product barrier. No analysis is
adversely affected by the proposed changes. No system or design
function or equipment qualification would be adversely affected by
the proposed changes. Furthermore, the proposed changes do not
result in a new failure mode, malfunction or sequence of events that
could affect safety or safety-related equipment.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to the CIM internal power supply enables the
FPGA to function properly. The function of the internal power supply
has no input to plant safety analysis. The change to the CIM
internal power supplies has a negligible effect on the 24 Vdc
supplies and ultimately the plant electrical system load and has no
adverse effect on the CIM functionality.
The proposed clarified descriptions and the proposed change to
the ownership of SRNC and CIM intellectual property are not
technical changes. The proposed changes do not adversely affect the
design, construction, or operation of any plant SSCs, including any
equipment whose failure could initiate an accident or a failure of a
fission product barrier. No analysis is adversely affected by the
proposed changes. Furthermore, no system function, design function,
or equipment qualification will be adversely affected by the
changes. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed changes, thus no
margin of safety is reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius, LLC, 1111 Pennsylvania Ave. NW., Washington, DC 20004-2514.
NRC Branch Chief: Michael T. Markley.
South Carolina Electric & Gas Company, Inc., Docket Nos. 52-027 and 52-
028, Virgil C. Summer Nuclear Station Units 2 and 3, Fairfield, South
Carolina
Date of amendment request: September 28, 2016. A publicly-available
version is in ADAMS under Accession No. ML16272A373.
Description of amendment request: The amendment request proposes
changes to revise plant-specific Tier 1, plant-specific Tier 2, and
Combined License (COL) Appendix C information concerning the details of
the Class 1E direct current and uninterruptible power supply system
(IDS), specifically adding seven Class 1E fuse panels to the IDS
design. These proposed changes provide electrical isolation between the
non-Class 1E IDS battery monitors and their respective Class 1E battery
banks. Because, this proposed change requires a departure from Tier 1
information in the Westinghouse Electric Company's AP1000 Design
Control Document (DCD), the licensee also requested an exemption from
the requirements of the Generic DCD Tier 1 in accordance with 10 CFR
52.63(b)(1).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with NRC staff edits in square
brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to revise plant-specific Tier 1, COL
Appendix C, and [Updated Final Safety Analysis Report (UFSAR)]
information concerning details of the IDS, specifically the addition
of seven Class 1E fuse isolation panels at the interconnection of
the non-Class 1E IDS battery monitors and Class 1E IDS circuits, are
necessary to conform to Regulatory Guide 1.75 Rev. 2 (consistent
with UFSAR Appendix 1A exceptions) and IEEE 384-1981 to prevent a
fault on non-Class 1E circuits or equipment from degrading the
operation of Class 1E IDS circuits and equipment below an acceptable
level. The proposed changes do not adversely affect the design
functions of the IDS, including the Class 1E battery banks and the
battery monitors.
These proposed changes to revise plant-specific Tier 1, COL
Appendix C, and UFSAR information concerning details of the IDS,
specifically the addition of seven Class 1E fuse isolation panels at
the interconnection of the non-Class 1E IDS battery monitors and
Class 1E IDS circuits as described in the current licensing basis do
not have an adverse effect on any of the design functions of any
plant systems. The proposed changes do not adversely affect any
plant electrical system and do not affect the support, design, or
operation of mechanical and fluid systems required to mitigate the
consequences of an accident. There is no change to plant systems or
the response of systems to postulated accident conditions. There is
no change to the predicted radioactive releases due to postulated
accident conditions. The plant response to previously evaluated
accidents or external events is not adversely affected, nor do the
proposed changes create any new accident precursors.
Therefore, the requested amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
[[Page 73439]]
The proposed changes to revise plant-specific Tier 1, COL
Appendix C, and UFSAR information concerning details of the IDS,
specifically the addition of seven Class 1E fuse isolation panels at
the interconnection of the non-Class 1E IDS battery monitors and
Class 1E IDS circuits, are necessary to conform to Regulatory Guide
1.75 Rev. 2 (consistent with UFSAR Appendix 1A exceptions) and IEEE
384-1981 to prevent a fault on non-Class 1E circuits or equipment
from degrading the operation of Class 1E IDS circuits and equipment
below an acceptable level. The proposed changes do not adversely
affect any plant electrical system and do not adversely affect the
design function, support, design, or operation of mechanical and
fluid systems. The proposed changes do not result in a new failure
mechanism or introduce any new accident precursors. No design
function described in the UFSAR is adversely affected by the
proposed changes.
Therefore, the 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.
There is no safety-related [structure, system, and component
(SSC)] or function adversely affected by the proposed change to add
IDS fuse isolation panels to non-Class 1E IDS battery monitors and
Class 1E IDS circuits. No safety analysis or design basis acceptance
limit/criterion is challenged or exceeded by the proposed changes
and no margin or safety is reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius, LLC, 1111 Pennsylvania Ave. NW., Washington, DC 20004-2514.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
Date of amendment request: December 22, 2015, as supplemented by
letter dated July 27, 2016. Publicly-available versions are in ADAMS
under Accession Nos. ML15356A655 and ML16209A477, respectively.
Description of amendment request: The proposed changes would revise
the Combined License (COL) Appendix C and corresponding plant-specific
Tier 1 information to add two turbine building sump pumps to
accommodate the increased flow that will be experienced during
condensate polishing system rinsing operations, for each unit,
respectively. The proposed changes include information in the combined
license, Appendix C. An exemption request relating to the proposed
changes to the AP1000 DCD Tier 1 is included with the request.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below with NRC staff edits in square
brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to identify that there is more than one
turbine building sump and to add two turbine building sump pumps
(WWS-MP-07A and WWS-MP-07B) to COL Appendix C Subsection 2.3.29 and
corresponding Table 2.3.29-1 will provide consistency within the
current licensing basis. The main turbine building sumps and sump
pumps are not safety-related components and do not interface with
any systems, structures, or components (SSCs) accident initiator or
initiating sequence of events; thus, the probability of accidents
evaluated within the [Updated Final Safety Analysis Report (UFSAR)]
are not affected. The proposed changes do not involve a change to
the predicted radiological releases due to accident conditions, thus
the consequences of accidents evaluated in the UFSAR are not
affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to identify that there is more than one
turbine building sump and to add two turbine building sump pumps to
the non-safety waste water system (WWS) do not affect any safety-
related equipment, nor do they add any new interface to safety-
related SSCs. No system or design function or equipment
qualification is affected by these changes. The changes do not
introduce a new failure mode, malfunction, or sequence of events
that could affect safety or safety-related equipment.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The WWS is a non-safety-related system that does not interface
with any safety-related equipment. The proposed changes to identify
that there is more than one turbine building sump and to add two
turbine building sump pumps do not affect any design code, function,
design analysis, safety analysis input or result, or design/safety
margin. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed change.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County,
Georgia
Date of amendment request: August 23, 2016. A publicly-available
version is in ADAMS under Accession No. ML16236A266.
Description of amendment request: The proposed changes would amend
Combined License Nos. NPF-91 and NPF-92 for the Vogtle Electric
Generating Plant, Units 3 and 4. The amendments propose changes to the
Updated Final Safety Analysis Report (UFSAR) in the form of departures
from the incorporated plant-specific Design Control Document Tier 2
information and involve related changes to the Combined Operating
License Appendix C (and corresponding plant-specific design control
document Tier 1) information. Specifically, the proposed departures
consist of changes to the design reliability assurance program (D-RAP)
to identify the covers for the in-containment refueling water storage
tank vents and overflow weirs as the risk-significant components
included in the D-RAP and to differentiate between the rod drive motor-
generator (MG) sets field control relays and the rod drive power supply
control cabinets in which the relays are located.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
[[Page 73440]]
Response: No.
The in-containment refueling water storage tank (IRWST) provides
flooding of the refueling cavity for normal refueling. The tank also
serves as a heat sink during Passive Residual Heat Removal (PRHR)
Heat Exchanger (HX) operation and in the event of a loss-of-coolant-
accident (LOCA) provides injection in support of long-term RCS
[reactor coolant system] cooling. This activity adds normally closed
covers to the IRWST vents and overflow weirs to prevent debris from
entering the tank, prevent over-pressurization and accommodate
volume and mass increases in the tank. The vent and overflow weir
covers open upon differential pressures between the IRWST and
containment.
The rod drive MG sets provide the power to the control rod drive
mechanisms through the reactor trip switchgear. This activity
revises the equipment description and equipment tag associated with
the risk-significant control relays which open to de-energize the
rod drive MG sets and permit rods to drop.
The proposed changes to add the IRWST vent and overflow weir
covers and to change the description of the equipment and equipment
tag related to the rod drive MG sets does not inhibit the SSCs from
performing their safety-related function. The design bases of the
IRWST vents and overflow weirs are not modified as a result of the
addition of the covers to the vents and overflow weirs and the
change to the control cabinet relay description and equipment tag.
This proposed amendment does not have an adverse impact on the
response to anticipated transients or postulated accident conditions
because the functions of the SSCs are not changed. Required IRWST
venting is not affected for any accident conditions. Required DAS
functions are not affected for any accident conditions. Safety-
related structure, system, component (SSC) or function is not
adversely affected by this change. The changes to include the IRWST
covers and to change the control cabinet relay description and tag
number do not involve an interface with any SSC accident initiator
or initiating sequence of events, and thus, the probabilities of the
accidents evaluated in the UFSAR are not affected. The proposed
changes do not involve a change to the predicted radiological
releases due to postulated accident conditions, thus, the
consequences of the accidents evaluated in the UFSAR are not
affected. Probabilistic Risk Assessment (PRA) modeling and analyses
associated with the SSCs are not impacted by this change.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to the design of the IRWST vent and
overflow weir covers do not adversely affect any safety-related
equipment, and do not add any new interfaces to safety-related SSCs.
No system or design function or equipment qualification is affected
by these changes. The changes do not introduce a new failure mode,
malfunction or sequence of events that could affect plant safety or
safety-related equipment as the simplistic design of the cover
louvers and hinged flappers are not considered unique designs. No
new credible failure modes are introduced by the addition of the
covers.
The proposed changes to the description and equipment tag
associated with the risk-significant control relays for the rod
drive MG sets do not adversely affect any safety-related equipment,
and do not add any new interfaces to safety-related SSCs. No system
or design function or equipment qualification is affected by these
changes. The changes do not introduce a new failure mode,
malfunction or sequence of events that could affect plant safety or
safety-related equipment because the design function of the control
relays, control cabinets, or rod drive MG sets is not changed.
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 compliance with the applicable
Codes and Standards, thereby maintaining the margin of safety
associated with these SSCs. The proposed changes do not alter any
applicable design codes, code compliance, design function, or safety
analysis. Consequently, no safety analysis or design basis
acceptance limit/criterion is challenged or exceeded by the proposed
change, thus the margin of safety is not reduced. Because no safety
analysis or design basis acceptance limit/criterion is challenged or
exceeded by these changes, no margin of safety is reduced.
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: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Inc., Docket Nos. 52-025 and 52-
026, Vogtle Electric Generating Plant, Units 3 and 4, Burke County,
Georgia
Date of amendment request: August 31, 2016. A publicly-available
version is in ADAMS under Accession No. ML16244A836.
Description of amendment request: The amendments propose changes to
the Updated Final Safety Analysis Report (UFSAR) in the form of
departures from the incorporated plant-specific Design Control Document
Tier 2* information. Specifically, the proposed changes would revise
the Combined Licenses for the Vogtle Electric Generating Plant, Units 3
and 4, to clarify information in WCAP-17179, ``AP1000[supreg] Component
Interface Module Technical Report,'' which demonstrates design
compliance with licensing bases requirements. WCAP-17179 is
incorporated by reference into the UFSAR to provide additional details
regarding the component interface module (CIM) system design. The
requested amendments also propose a change to the CIM internal power
supply that will enable proper functioning of the field programmable
gate arrays.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the CIM internal power supply enables the
field programmable gate array (FPGA) to function properly. The
proposed change to the FPGA core power has no adverse effect on the
operation of the output actuation relays. The function of the
internal power supply has no input to plant safety analysis. The
change to the CIM internal power supply has a negligible effect on
the 24 Vdc [volts direct current] supplies and ultimately the plant
electrical system load and has no adverse effect on the CIM
functionality.
The proposed changes to clarify how licensing basis design
documentation reflects compliance with license basis requirements,
and the proposed change to the ownership of safety remote node
controller (SRNC) and CIM intellectual property, are not technical
changes. The proposed changes do not affect any accident initiator
in the UFSAR, or affect the radioactive material releases in the
UFSAR accident analyses. The proposed change does not alter the
ability of the facility to prevent and mitigate abnormal events,
e.g., accidents, anticipated operational occurrences, earthquakes,
floods and turbine missiles, or their safety or design analyses. No
safety-related structure, system, or component (SSC) or function is
adversely affected. The change does not involve or interface with
any SSC accident initiator or initiating sequence of events, and
thus, the probabilities of the accidents evaluated in the UFSAR are
not affected. This activity does not involve a new fission product
release path, nor a new fission product barrier failure mode, nor
create a new sequence of events that would result in significant
fuel cladding failures. Because the proposed changes do
[[Page 73441]]
not change any safety-related SSC or function credited in the
mitigation of an accident, the consequences of the accidents
evaluated in the UFSAR are not affected.
Therefore, the requested amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change to the CIM internal power supply enables the
FPGA to function properly and does not involve accident initiators.
The change to the CIM internal power supply has a negligible effect
on the 24 Vdc supplies and ultimately the plant electrical system
load and has no adverse effect on CIM functionality.
The proposed clarified descriptions and the proposed change to
the ownership of SRNC and CIM intellectual property are not
technical changes. The proposed changes do not affect other plant
equipment or adversely affect the design of the CIM. Therefore, the
proposed changes do not affect any safety-related equipment itself,
nor do they affect equipment whose failure could initiate an
accident or a failure of a fission product barrier. No analysis is
adversely affected by the proposed changes. No system or design
function or equipment qualification would be adversely affected by
the proposed changes. Furthermore, the proposed changes do not
result in a new failure mode, malfunction or sequence of events that
could affect safety or safety-related equipment.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to the CIM internal power supply enables the
FPGA to function properly. The function of the internal power supply
has no input to plant safety analysis. The change to the CIM
internal power supplies has a negligible effect on the 24 Vdc
supplies and ultimately the plant electrical system load and has no
adverse effect on the CIM functionality.
The proposed clarified descriptions and the proposed change to
the ownership of SRNC and CIM intellectual property are not
technical changes. The proposed changes do not adversely affect the
design, construction, or operation of any plant SSCs, including any
equipment whose failure could initiate an accident or a failure of a
fission product barrier. No analysis is adversely affected by the
proposed changes. Furthermore, no system function, design function,
or equipment qualification will be adversely affected by the
changes. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed changes, thus no
margin of safety is reduced.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Michael T. Markley.
Southern Nuclear Operating Company, Docket No. 50-364, Joseph M. Farley
Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendment request: September 8, 2016. A publicly-available
version is in ADAMS under Accession No. ML16256A135.
Description of amendment request: The amendment would correct an
error in the Joseph M. Farley Nuclear Plant, Unit 2, Renewed Facility
Operating License No. NPF-8, for Condition 2.C.(23). Specifically, the
Unit 2 referenced date prior to the period of extended operation was
incorrectly entered as June 25, 2017. This date corresponds to the Unit
1 period of extended operation. The Unit 2 correct date for this
license condition is March 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, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment contains no technical changes; all
proposed changes are administrative. These changes are consistent
with the intent of what has already been approved by the Nuclear
Regulatory Commission (NRC). There are no accidents affected by this
change, and therefore no increase in the probability or consequences
of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment contains no technical changes; all
proposed changes are administrative. These changes are consistent
with the intent of what has already been approved by the Nuclear
Regulatory Commission (NRC). There are no accidents affected by this
change, and therefore no 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 amendment contains no technical changes; all
proposed changes are administrative. These changes are consistent
with the intent of what has already been approved by the Nuclear
Regulatory Commission (NRC). There are no accidents affected by this
change, and therefore no 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: Jennifer M. Buettner, Associate General
Counsel, Southern Nuclear Operating Company, Inc., 40 Inverness Center
Parkway, Birmingham, AL 35242.
NRC Branch Chief: Michael T. Markley.
Susquehanna Nuclear, LLC, Docket Nos. 50-387 and 50-388, Susquehanna
Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: July 27, 2016, as supplemented by letter
dated September 13, 2016. Publicly-available versions are in ADAMS
under Accession Nos. ML16210A001 and ML16257A598, respectively.
Description of amendment request: The amendments would revise
Technical Specification 3.6.4.1, ``Secondary Containment,''
Surveillance Requirement (SR) 3.6.4.1.3 to provide an allowance for
brief, inadvertent, simultaneous opening of redundant secondary
containment access doors during normal entry and exit conditions.
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, along with NRC edits in square
brackets:
1. Do the proposed changes involve a significant increase in the
probability or consequences of any accident previously evaluated?
Response: No.
The proposed change does not involve any physical change to
structures, systems, or components (SSCs) and do not alter the
method of operation of any SSCs. The proposed change addresses a
temporary condition during which Secondary Containment SRs are not
met. The Secondary Containment is not an initiator of any accident
previously evaluated. As a result, the probability of any accident
previously evaluated is not increased. [Two accidents credit the
Secondary Containment from a dose consequence perspective. They are
the
[[Page 73442]]
loss-of-coolant accident (LOCA) and fuel/equipment handling
accident. Each accident requires time to drawdown the secondary
containment to less than atmospheric pressure. The brief,
inadvertent, simultaneous opening of both an inner and outer
personnel access door during normal entry and exit conditions
followed by prompt closure does not challenge the design basis
drawdown time and does not result in an increase in any on-site or
offsite dose for the LOCA dose analysis. All dose consequences are
within the regulatory limits established for the fuel handling
accident and bound the case in which airlock doors are briefly,
inadvertently opened.] As a result, the consequences of any accident
previously evaluated is not significantly increased.
Therefore, the proposed change does 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 change does not involve a physical alteration of
any plant equipment. No new equipment is being introduced, and
installed equipment is not being operated in a new or different
manner. There are not setpoints, at which protective or mitigative
actions are initiated, affected by the proposed change. The proposed
change does not alter the manner in which equipment operation is
initiated, nor will the function of credited equipment be changed.
No alterations in the procedures that ensure the plant remains
within analyzed limits are being proposed, and no changes are being
made to the procedures relied upon to respond to an off-normal event
described in the FSAR [Final Safety Analysis Report]. As such, no
new failure modes are being introduced. The change does not alter
the assumptions made in the safety analysis and licensing basis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed change addresses temporary conditions
during which the Secondary Containment SR is not met. The allowance
for both an inner and outer Secondary Containment access door to be
open simultaneously for entry and exit does not affect the safety
function of the reactor enclosure and refuel area Secondary
Containments as the doors are promptly closed after entry of exit,
thereby restoring the Secondary Containment boundary. In addition,
brief, inadvertent simultaneous opening and closing of redundant
Secondary Containment personnel access doors during normal entry and
exit conditions does not affect the ability of the SGTS to establish
the required Secondary Containment vacuum. Therefore, the safety
function of the Secondary Containment is not affected.
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: Damon D. Obie, Esquire, Associate General
Counsel, Talen Energy Supply, LLC, 835 Hamilton St., Suite 150,
Allentown, PA 18101.
NRC Branch Chief: Douglas A. Broaddus.
Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant
(WBN), Unit 2, Rhea County, Tennessee
Date of amendment request: September 30, 2016. A publicly-available
version is in ADAMS under Accession No. ML16277A477.
Description of amendment request: The amendment would revise the
Technical Specifications (TSs) to allow a one-time extension of the
frequency for performing TS Surveillance Requirements (SRs) related to
verifying the operability of the containment ice bed.
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 requested action is a one-time extension to the performance
interval for TS SRs [3.6.11.2] and 3.6.11.3. The performance of
these surveillances, or the extension of these surveillances, is not
a precursor to an accident. Performing these surveillances or
failing to perform these surveillances does not affect the
probability of an accident.
Therefore, the proposed delay in performance of the SRs in this
amendment request does not increase the probability of an accident
previously evaluated.
A delay in performing these surveillances does not result in a
system being unable to perform its required function. In the case of
this one-time extension request, the short period of additional time
that the systems and components will be in service before the next
performance of the surveillance will not affect the ability of those
systems to operate as designed. Therefore, the systems required to
mitigate accidents will remain capable of performing their required
function. No new failure modes have been introduced because of this
action and the consequences remain consistent with previously
evaluated accidents. On this basis, the proposed delay in
performance of the SRs in this amendment request does not involve a
significant increase in the consequences of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not involve a physical alteration of
any system, structure, or component (SSC) or a change in the way any
SSC is operated. The proposed amendment does not involve operation
of any SSCs in a manner or configuration different from those
previously recognized or evaluated. No new failure mechanisms will
be introduced by the one-time SR extensions being requested.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment is a one-time extension of the
performance interval of TS SRs [3.6.11.2] and 3.6.11.3. Extending
these surveillance requirements does not involve a modification of
any TS limiting conditions for operation. Extending these SRs does
not involve a change to any limit on accident consequences specified
in the license or regulations. Extending these SRs does not involve
a change in how accidents are mitigated or a significant increase in
the consequences of an accident. Extending these SRs does not
involve a change in a methodology used to evaluate consequences of
an accident. Extending these SRs does not involve a change in any
operating procedure or process.
Based on the limited additional period of time that the systems
and components will be in service before the surveillances are next
performed, as well as the operating experience that these
surveillances are typically successful when performed, it is
reasonable to conclude that the margins of safety associated with
these SRs will not be affected by the requested extension.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Sherry A. Quirk, Executive Vice
President and General Counsel, Tennessee Valley Authority, 400 West
Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
NRC Acting Branch Chief: Jeanne A. Dion.
[[Page 73443]]
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: May 18, 2016. A publicly-available
version is in ADAMS under Accession No. ML16146A540.
Description of amendment request: The amendments would revise the
Surry Power Station, Unit Nos. 1 and 2, Technical Specification (TS)
3.14, ``Circulating and Service Water Systems,'' to extend the allowed
outage time (AOT) for only one operable service water (SW) flow path to
the charging pump service water (CPSW) subsystem and to the main
control room/emergency switchgear room (MCR/ESGR) air conditioning (AC)
subsystem. TS 3.14.A.5 and TS 3.14.A.7 require two SW flow paths to the
CPSW subsystem and to the MCR/ESGR AC subsystem, respectively, to be
operable. Currently, the TS 3.14.C AOT for only one operable CPSW or
MCR/ESGR AC flow path is 24 hours. The proposed revision will extend
the AOT for only one operable CPSW or MCR/ESGR AC flow path from 24
hours to 72 hours.
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 extends the AOT for only one operable CPSW
or MCR/ESGR AC flow path from 24 hours to 72 hours. The CPSW
subsystem is a support system for the Charging/High Head Safety
Injection (HHSI) pumps; the proposed CPSW AOT extension aligns the
CPSW support system AOT with the AOT for the supported components
(i.e., the Charging/HHSI pumps). The proposed MCR/ESGR AC AOT
extension revises the AOT to be the same as the CPSW AOT since both
subsystems share common piping. The design function of the CPSW
system, which is to provide cooling to the charging pump
intermediate seal coolers and the charging pump lubricating oil
coolers, is not impacted by the proposed revision, nor is the design
function of the Charging/HHSI pumps impacted. Furthermore, the
design functions of the MCR/ESGR AC subsystem and the MCR/ESGR
chillers are not impacted by the proposed revision. In addition, the
proposed change deletes the now expired and no longer necessary
requirements for the temporary SW jumper to the CCHXs [component
cooling heat exchangers]. The deletion of these temporary
requirements is administrative in nature. As a result, 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 extends the AOT for only one operable CPSW
or MCR/ESGRAC flow path from 24 hours to 72 hours. In addition, the
proposed change deletes the now expired and no longer necessary
requirements for the temporary SW jumper to the CCHXs. The proposed
change does not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) and does not
impact plant operation. Furthermore, the proposed change does not
impose any new or different requirements that could initiate an
accident. The proposed change does not alter assumptions made in the
safety analysis and is consistent with the safety analysis
assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change extends the AOT for only one operable CPSW
or MCR/ESGR AC flow path from 24 hours to 72 hours. The proposed
change does not adversely affect any current plant safety margins or
the reliability of the equipment assumed in the safety analysis.
There are no changes being made to any safety analysis assumptions,
safety limits, or limiting safety system settings that would
adversely affect plant safety as a result of the proposed change.
Furthermore, as noted above, a supporting PRA [probabilistic risk
assessment] was performed for the proposed AOT changes. The PRA
concluded that the increase in risk associated with the proposed
changes is consistent with the RG [Regulatory Guide] 1.174 and RG
1.177 acceptance guidelines for a permanent TS AOT change. This PRA
evaluation demonstrates that defense-in-depth will not be
significantly impacted by changing the AOTs for only one operable SW
flow path to the CPSW subsystem and to the MCR/ESGR AC subsystem
from 24 to 72 hours. In addition, the proposed change deletes the
now expired and no longer necessary requirements for the temporary
SW jumper to the CCHXs. The deletion of these temporary requirements
is administrative in nature. 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: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
NRC Branch Chief: Michael T. Markley.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: July 14, 2016. A publicly-available
version is in ADAMS under Accession No. ML16202A068.
Description of amendment request: The amendments would revise the
Surry Power Station, Unit Nos. 1 and 2, Technical Specification (TS)
3.14, ``Circulating and Service Water Systems,'' to extend the allowed
outage time (AOT) for emergency service water (ESW) pump inoperability.
The proposed revision would extend the TS 3.14.B AOT for one inoperable
ESW pump from 7 days to 14 days to provide operational flexibility for
ESW pump maintenance and repairs.
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 design function of the ESW pumps is to ensure that water can
be provided to the intake canal (i.e., the ultimate heat sink) when
power is not available to the Circulating Water (CW) pumps. The
proposed extension of the AOT for one inoperable ESW pump from 7 to
14 days does not impact the design function of the ESW pumps. In
addition, the number of ESW pumps required to be operable for the
specified plant operating conditions is not impacted by the proposed
AOT extension. As a result, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (i.e., no new or different type of equipment will be
installed) and does not impact plant operation. Furthermore, the
proposed change does not impose any new or different requirements
that could initiate an accident. The proposed change does not alter
assumptions made in the safety analysis and is consistent with the
safety analysis assumptions.
Therefore, the proposed change does not create the possibility
of a new or different
[[Page 73444]]
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 does not adversely affect any current plant
safety margins or the reliability of the equipment assumed in the
safety analysis. There are no changes being made to any safety
analysis assumptions, safety limits, or limiting safety system
settings that would adversely affect plant safety as a result of the
proposed change. Furthermore, as noted above, a supporting PRA
[probabilistic risk assessment] was performed for the proposed AOT
change. The PRA concluded that the increase in risk associated with
the proposed change is consistent with the RG [Regulatory Guide]
1.174 and RG 1.177 acceptance guidelines for a permanent TS AOT
change. This PRA evaluation demonstrates that defense-in-depth will
not be significantly impacted by changing the AOT for one inoperable
ESW pump from 7 to 14 days.
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: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., 120 Tredegar St., RS-2, Richmond, VA 23219.
NRC Branch Chief: Michael T. Markley.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, 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.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2 (MPS2), New London County, Connecticut
Date of amendment request: September 1, 2015, as supplemented by
letter dated March 24, 2016.
Brief description of amendment: The amendment revised the MPS2
Technical Specifications (TSs) to add the evaluation model EMF-
2328(P)(A), Supplement 1, ``PWR [pressurized water reactor] Small Break
LOCA [loss-of coolant accident] Evaluation Model S-RELAP5 Based,'' and
EMF-92-116(P)(A), Supplement 1, ``Generic Mechanical Design Criteria
for PWR Fuel Designs,'' to the TS Section 6.9.1.8.b list of analytical
methods used to determine core operating limits as a result of
reanalyzing the small break LOCA.
Date of issuance: September 30, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 329. A publicly-available version is in ADAMS under
Accession No. ML16249A001; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-65: Amendment revised
the Renewed Operating License and TSs.
Date of initial notice in Federal Register: December 8, 2015 (80 FR
76318). The supplemental letter dated March 24, 2016, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 30, 2016.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
1 (ANO-1), Pope County, Arkansas
Date of amendment request: January 29, 2014, as supplemented by
letters dated May 19, June 16, July 21, August 12, September 22,
November 4, and November 17, 2015; and January 15, March 25, April 7,
May 19, and August 29, 2016.
Brief description of amendment: The amendment authorized the
transition of the ANO-1 fire protection program to a risk-informed,
performance-based program based on National Fire Protection Association
(NFPA) 805, in accordance with 10 CFR 50.48(c). NFPA 805 allows the use
of performance-based methods such as fire modeling and risk-informed
methods such as fire probabilistic risk assessment to demonstrate
compliance with the nuclear safety performance criteria.
Date of issuance: October 7, 2016.
Effective date: As of the date of issuance and shall be implemented
as described in the transition license conditions.
Amendment No.: 256. A publicly-available version is in ADAMS under
Accession No. ML16223A481; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-51: The amendment
revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 8, 2014 (79 FR
38589). The supplemental letters dated May 19, June 16, July 21, August
12, September 22, November 4, and November 17, 2015; and January 15,
March 25, April 7, May 19, and August 29, 2016, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 7, 2016.
No significant hazards consideration comments received: No.
[[Page 73445]]
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: October 15, 2015, as supplemented by a
letter dated May 6, 2016.
Description of amendment request: The amendments revised the St.
Lucie Plant, Unit Nos. 1 and 2, Technical Specifications (TSs) and
licensing bases to reflect the use of the commercially available
computer code ``Generation of Thermal-Hydraulic Information for
Containments (GOTHIC Version 7.2b(QA))'' to model the containment
response following the inadvertent actuation of the containment spray
system during normal plant operation (referred to as the vacuum
analysis). The amendments also updated the licensing bases to credit
the design basis ability of the containment vessel to withstand a
higher external pressure differential of 1.04 pounds per square inch
(psi) for Unit No. 1 and 1.05 psi for Unit No. 2, and updated TS
3.6.1.4 for each unit to revise the allowable containment operating
pressure range.
Date of issuance: October 5, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 236 (Unit No. 1) and 186 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML16166A424;
documents related to these amendments are listed in the Safety
Evaluation (SE) enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-67 and NPF-16:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: February 16, 2016 (81
FR 7839). The supplemental letter dated May 6, 2016, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in an SE dated October 5, 2016.
No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County,
Texas
Date of amendment request: November 12, 2015, as supplemented by
letters dated December 9, 2015, and March 14, March 29, April 7, April
20, August 16, September 16, September 21, and September 29, 2016.
Brief description of amendments: By order dated May 6, 2016, as
published in the Federal Register on May 23, 2016 (81 FR 32350), the
NRC approved the transfer of Facility Operating License (FOL) Nos. NPF-
87 and NPF-89 for Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2,
and the general license for the independent spent fuel storage
installation facility from the current holder, Luminant Generation
Company LLC, to Comanche Peak Power Company LLC, as owner, and TEX
Operations Company LLC, as operator. The conforming amendments revised
the FOLs to reflect the direct transfer of ownership and the indirect
transfer of control of the licenses.
Date of issuance: October 3, 2016.
Effective date: As of the date of issuance and shall be implemented
within 7 days from the date of issuance.
Amendment Nos.: 167 (Unit No. 1) and 167 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML16266A005;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Facility Operating License Nos. NPF-87 and NPF-89: Amendments
revised the FOLs.
Date of initial notice in Federal Register: February 8, 2016 (81 FR
6545). The supplemental letters dated March 14, March 29, April 7,
April 20, August 16, September 16, September 21, and September 29,
2016, provided additional information that clarified the application
and did not expand the scope of the application as originally noticed.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 6, 2016.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date amendment request: October 12, 2015.
Brief description of amendments: The amendments revised Salem
Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specifications
(TSs) by adding MODE 4 to the applicability of TS 3.6.2.3,
``Containment Cooling System.'' The amendments also revised TS 3.7.1.1,
``Safety Valves,'' to correct discrepancies between the applicable
modes and the action statements.
Date of issuance: September 29, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 315 (Unit No. 1) and 296 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML16229A519;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-70 and DPR-75: The
amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: January 5, 2016 (81 FR
264).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 29, 2016.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company and South Carolina Public Service
Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer Nuclear
Station, Units 2 and 3, Fairfield County, South Carolina
Date of amendment request: March 14, 2016, as supplemented by
letters dated May 12, 2016, and July 12, 2016.
Description of amendment: The amendments incorporated changes that
are consistent with those generically approved in WCAP-17524-P-A,
Revision 1, ``AP1000 Core Reference Report,'' dated February 19, 2015.
The amendments also approved changes to the Updated Final Safety
Analysis Report (UFSAR) in the form of departures from the incorporated
plant-specific Design Control Document Tier 2 licensing basis
information, involves changes to the UFSAR information that has been
designated as Tier 2* information, and involves changes to the plant-
specific Technical Specifications.
Date of issuance: September 20, 2016.
Effective date: As of the date of issuance and shall be implemented
within 50 days of issuance.
Amendment Nos.: 52 (Unit 2) and 52 (Unit 3). A publicly-available
version is in ADAMS under Package Accession No. ML16144A591; documents
related to these amendments are listed in the Safety Evaluation
enclosed with the amendments.
Facility Combined Licenses Nos. NPF-93 and NPF-94: Amendments
revised the Facility Combined Licenses and Technical Specifications.
Date of initial notice in Federal Register: May 10, 2016 (81 FR
28900). The supplemental letters dated May 12, 2016, and July 12, 2016,
provided additional information that clarified the
[[Page 73446]]
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 20, 2016.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc.; Georgia Power Company;
Oglethorpe Power Corporation; Municipal Electric Authority of Georgia;
and City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I.
Hatch Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia
Date of amendment request: October 10, 2014, as supplemented by
letters dated May 4, 2015; October 15, 2015; and August 26, 2016.
Brief description of amendments: The amendments revised Technical
Specifications (TSs) by adopting 18 previously NRC-approved Technical
Specifications Task Force (TSTF) Travelers, two TSTF T-Travelers, and
one feature of the Improved Standard Technical Specifications not
associated with a Traveler.
Date of issuance: September 29, 2016.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: 279 (Unit 1) and 223 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16231A041; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Renewed Facility Operating Licenses and TSs.
Date of initial notice in Federal Register: March 31, 2015 (80 FR
17095). The supplemental letters dated May 4, 2015; October 15, 2015;
and August 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 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 September 29, 2016.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc.; Georgia Power Company;
Oglethorpe Power Corporation; Municipal Electric Authority of Georgia;
City of Dalton, Georgia; Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia
Date of amendment request: October 15, 2015, as supplemented by
letters dated March 16, May 9, and May 16, 2016.
Brief description of amendments: The amendments revised the
Technical Specifications Surveillance Requirement 3.6.4.1.3 to increase
the allowable time from 2 minutes to 10 minutes for the standby gas
treatment system to draw down the secondary containment to negative
pressure.
Date of issuance: September 30, 2016.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 280 (Unit No. 1) and 224 (Unit No. 2). A publicly-
available version is in ADAMS under Accession No. ML16235A287;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Renewed Facility Operating Licenses and
Technical Specifications.
Date of initial notice in Federal Register: November 24, 2015 (80
FR 73240). The supplemental letters dated March 16, May 9, and May 16,
2016, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 30, 2016.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: May 16, 2016.
Brief description of amendments: The amendments consisted of change
to the Completion Date of Cyber Security Plan (CSP) Implementation
Milestone 8--full implementation of the CSP from October 31, 2016 to
December 31, 2017.
Date of issuance: October 3, 2016.
Effective date: As of its date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 337 (Unit 1) and 330 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16228A096; documents related
to these amendments are listed in the Safety Evaluation (SE) enclosed
with the amendments.
Facility Operating License Nos. DPR-77 and DPR-79. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 8, 2016 (81 FR
44665).
The Commission's related evaluation of the amendments is contained
in an SE dated October 3, 2016.
No significant hazards consideration comments received: No.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: March 11, 2016, as supplemented by
letters dated May 31, 2016, and July 22, 2016.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) by adding a new Condition A to TS 3.7.8,
``Essential Raw Cooling Water (ERCW) System,'' to extend the allowed
completion time to restore ERCW System train to OPERABLE status from 72
hours to 7 days for planned maintenance when the opposite unit is
defueled or in Mode 6 following defueled under certain restrictions.
Date of issuance: September 29, 2016.
Effective date: As of its date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 336 (Unit 1) and 329 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML16225A276; documents related
to these amendments are listed in the Safety Evaluation (SE) enclosed
with the amendments.
Facility Operating License Nos. DPR-77 and DPR-79. Amendments
revised the TSs.
Date of initial notice in Federal Register: April 12, 2016 (81 FR
21603). The supplemental letters dated May 31, 2016, and July 22, 2016,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in an SE dated September 29, 2016.
[[Page 73447]]
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 13th day of October 2016.
For the Nuclear Regulatory Commission.
George A. Wilson, Jr.,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2016-25641 Filed 10-24-16; 8:45 am]
BILLING CODE 7590-01-P