[Federal Register Volume 79, Number 189 (Tuesday, September 30, 2014)]
[Notices]
[Pages 58812-58831]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-23015]
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NUCLEAR REGULATORY COMMISSION
[NRC-2014-0207]
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 4, 2014 to September 17, 2014. The
last biweekly notice was published on September 16, 2014.
DATES: Comments must be filed by October 30, 2014. A request for a
hearing must be filed by December 1, 2014.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0207. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; 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: 3WFN-06-A44M, 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: Mable Henderson, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-3760, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2014-0207 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-2014-0207.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in the
SUPPLEMENTARY INFORMATION section.
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-2014-0207 in the subject line of your
comment submission, in order to ensure that the NRC is able to make
your comment submission available to the public in this docket.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment submissions into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of Title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
[[Page 58813]]
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 should 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. Should the
Commission take 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. Should the Commission make 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
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license or
combined license. Requests for a hearing and a petition for leave to
intervene shall be filed in accordance with the Commission's ``Agency
Rules of Practice and Procedure'' in 10 CFR Part 2. Interested
person(s) 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 request for a hearing or petition for leave to intervene is
filed by the above date, 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 request
and/or 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 for leave to intervene
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 requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/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 requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/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
requestor/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 requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. 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 amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
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.
If a hearing is requested, 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 or safety of the
public, in which case it will issue an appropriate order or rule under
10 CFR Part 2.
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, 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). 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 request or
petition for hearing (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://
[[Page 58814]]
www.nrc.gov/site-help/e-submittals/getting-started.html. System
requirements for accessing the E-Submittal server are detailed in the
NRC's ``Guidance for Electronic Submission,'' which is available on the
agency's public Web site at http://www.nrc.gov/site-help/e-submittals.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 Meta System Help Desk
will not be able to offer assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through the Electronic Information Exchange System,
users will be required to install a Web browser plug-in from the NRC's
Web site. Further information on the Web-based submission form,
including the installation of the Web browser plug-in, is available on
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with the NRC's guidance
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.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 request/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 Meta System
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 Meta System Help Desk is available between 8 a.m. and 8 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
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 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, a request to intervene 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.
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)-(iii).
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant (PNP), Van Buren County, Michigan
Date of amendment request: July 29, 2014. A publicly-available
version is in ADAMS under Accession No. ML14211A520.
Description of amendment request: The amendment would change the
Operating License at PNP. Specifically, the amendment requests
authorization to implement 10 CFR 50.61a, ``Alternate fracture
toughness requirements for protection against pressurized thermal shock
events,'' in lieu of 10 CFR 50.61, ``Fracture toughness requirements
for protection against pressurized thermal shock events.'' PNP
currently complies with 10 CFR 50.61. The 10 CFR 50.61 screening
criteria define a limiting level of embrittlement beyond which plant
operation cannot continue without further evaluation. As described in
NUREG-1806, ``Technical Basis for Revision of the Pressurized Thermal
Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61),'' August
2007, the screening criteria in the PTS rule is overly conservative and
the risk of through-wall cracking due to a PTS event is much lower than
previously estimated. A publically-available version of NUREG-1806 is
in ADAMS under Accession No. ML072830074.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 58815]]
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This amendment request would allow implementation of the 10 CFR
50.61a alternate pressurized thermal shock (PTS) rule in lieu of the
10 CFR 50.61 PTS rule, and would not involve a significant increase
in the probability or consequences of an accident. Application of 10
CFR 50.61a in lieu of 10 CFR 50.61 would not result in physical
alteration of a plant structure, system or component, or
installation of new or different types of equipment. Further,
application of 10 CFR 50.61a would not significantly affect the
probability of accidents previously evaluated in the Updated Final
Safety Analysis Report (UFSAR) or cause a change to any of the dose
analyses associated with the UFSAR accidents because accident
mitigation functions would remain unchanged. Use of 10 CFR 50.61a
would change how fracture toughness of the reactor vessel is
assessed and does not affect reactor vessel neutron radiation
fluence. As such, implementation of 10 CFR 50.61a in lieu of 10 CFR
50.61 would not increase the likelihood of a malfunction.
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 type of accident from any accident previously evaluated?
Response: No.
The amendment request would allow implementation of the 10 CFR
50.61a alternate PTS rule in lieu of 10 CFR 50.61. No new accident
scenarios, failure mechanisms, or limiting single failures are
introduced as a result of the proposed change. No physical plant
alterations are made as a result of the proposed change. The
proposed change does not challenge the performance or integrity of
any safety-related system. 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 amendment request would authorize implementation of 10 CFR
50.61a in lieu of 10 CFR 50.61. Regulation 10 CFR 50.61a would
maintain the same functional requirements for the facility as 10 CFR
50.61. It establishes screening criteria that limit levels of
embrittlement beyond which operation cannot continue without further
plant-specific evaluation or modifications. Sufficient safety
margins are maintained to ensure that any potential increases in
core damage frequency and large early release frequency resulting
from implementation of 10 CFR 50.61a are negligible. As such, there
would be no significant reduction in the margin of safety as a
result of use of the alternate PTS rule. The margin of safety
associated with the acceptance criteria of accidents previously
evaluated in the UFSAR is unchanged. The proposed change would have
no effect on the availability, operability, or performance of the
safety-related systems and components.
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: Mr. William Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White
Plains, NY 10601.
NRC Branch Chief: David L. Pelton.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Units 1 and 2, Will County, Illinois
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Units 1 and 2, Ogle County, Illinois
Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit 1, DeWitt County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-010, 50-237 and 50-249,
Dresden Nuclear Power Station, Units 1, 2 and 3, Grundy County,
Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: August 11, 2014. A publicly-available
version is in ADAMS under Accession No. ML14224A245.
Description of amendment request: The proposed changes would revise
the description for the Emergency Response Organization (ERO)
requalification training frequency for Exelon personnel defined in
Exelon's governing Emergency Plans for the named stations from annually
to ``once per calendar year not to exceed 18 months between training
sessions.''
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:
Exelon has evaluated the proposed changes to the affected sites'
Emergency Plans and determined that the changes do not involve a
Significant Hazards Consideration. In support of this determination,
an evaluation of each of the three (3) standards, set forth in 10
CFR 50.92, ``Issuance of amendment,'' is provided below.
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not increase the probability or
consequences of an accident. The proposed changes do not involve the
modification of any plant equipment or affect plant operation. The
proposed changes will have no impact on any safety-related
Structures, Systems, or Components (SSC).
The proposed changes would revise the ERO requalification
frequency from an annual basis to once per calendar year not to
exceed 18 months between training sessions defined in the Emergency
Plan for the applicable Exelon facility. The proposed changes will
align the Exelon legacy plants under one standard regarding the
annual requalification training frequency for ERO personnel.
Therefore, the proposed changes to the Emergency Plan
requalification training frequency for the affected sites do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes have no impact on the design, function, or
operation of any plant SSC. The proposed changes do not affect plant
equipment or accident analyses. The proposed changes only affect the
administrative aspects of the annual ERO requalification training
frequency requirements.
Therefore, the proposed changes to the Emergency Plan
requalification training frequency for the affected sites do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analyses. There is no change being made to
safety analysis assumptions, safety limits, or limiting safety
system settings that would adversely affect plant safety as a result
of the proposed changes. Margins of safety are unaffected by the
proposed changes to the frequency in the ERO requalification
training requirements.
Therefore, the proposed changes to the Emergency Plan
requalification training frequency for the affected sites 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
[[Page 58816]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the requested amendments involve no
significant hazards consideration.
Attorney for licensee: Bradley Fewell, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station (PBAPS), Units 2 and
3, York and Lancaster Counties, Pennsylvania
Date of application for amendments: July 25, 2014. A publicly-
available version is in ADAMS under Accession No. ML14211A017.
Description of amendment request: The proposed amendment would
change the definition in the PBAPS, Units 2 and 3, Technical
Specifications (TS) for RECENTLY IRRADIATED FUEL. Specifically, the
amendment would revise requirements pertaining to secondary containment
hatches in order to facilitate activities performed during refueling
outages.
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. Will operation of the facility in accordance with the
proposed amendment involve a significant increase in the probability
or consequences of an accident previously evaluated?
Response: No.
The proposed changes to revise the PBAPS, Units 2 and 3, TS
definition for RECENTLY IRRADIATED FUEL do not introduce new
equipment or new equipment operating modes, nor do the proposed
changes alter existing system relationships. The proposed changes do
not affect plant operation, [any] design function, or any analysis
that verifies the capability of a Structure, System, or Component
(SSC) to perform a design function. There are no changes or
modifications to [any] plant SSC. The plant Engineered Safety
Features (ESFs) will continue to function as designed in all modes
of operation. There are no significant changes to procedures or
training being introduced by the proposed changes to the TS
definition.
Based upon the results of the [fuel handling accident (FHA)]
analysis, it has been demonstrated that, with the requested changes,
the dose consequences remain within the regulatory guidance provided
by the NRC as specified in 10 CFR 50.67 and associated Regulatory
Guide (RG) 1.183 [ADAMS Accession No. ML003716792]. The calculations
used to evaluate the consequences of the FHA accident in support of
the proposed changes do not by themselves affect the plant response,
but better represent the physical characteristics of the release, so
that appropriate mitigation techniques may be applied. Therefore,
the consequences of an accident previously evaluated are not
significantly increased.
There is no adverse impact on systems designed to mitigate the
consequences of accidents. The proposed changes do not adversely
affect system or component pressures, temperatures, or flowrates for
systems designed to prevent accidents or mitigate the consequences
of an accident. Since these conditions are not adversely affected,
the likelihood of failure of [an] SSC is not increased.
The proposed changes do not increase the likelihood of the
malfunction of any SSC or impact any analyzed accident.
Consequently, the probability or consequences of an accident
previously evaluated are not affected.
Based on the above, Exelon concludes that the proposed changes
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Will operation of the facility in accordance with 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 revise the PBAPS, Units 2 and 3, TS
definition for RECENTLY IRRADIATED FUEL do not alter the design
function or operation of any SSC. There are no changes or
modifications to [any] plant SSC. The plant ESFs will continue to
function as designed. There is no new system component being
installed, no new construction, and no performance of a new test or
maintenance function. The proposed TS changes do not create the
possibility of a new credible failure mechanism or malfunction. The
proposed changes do not introduce new accident initiators or
precursors of a new or different kind of accident. New equipment or
personnel failure modes that might initiate a new type of accident
are not created as a result of the proposed changes. [Secondary
containment] integrity is not adversely impacted and radiological
consequences from the analyzed FHA remain within specified
regulatory limits. The proposed changes do not adversely impact
system or component pressures, temperatures, or flowrates for
systems designed to prevent accidents or mitigate the consequences
of an accident. Since these conditions are not adversely impacted,
the likelihood of failure of [an] SSC is not increased.
Consequently, the proposed changes cannot create the possibility of
a new or different kind of accident from any accident previously
evaluated.
Based on the above, Exelon concludes that the proposed changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Will operation of the facility in accordance with the
proposed amendment involve a significant reduction in a margin of
safety?
Response: No.
The proposed changes to revise the PBAPS, Units 2 and 3, TS
definition for RECENTLY IRRADIATED FUEL do not alter the design
function or operation of any SSC. There are no changes or
modifications to [any] plant SSC. The plant ESFs will continue to
function as designed. The proposed changes do not increase system or
component pressures, temperatures, or flowrates for systems designed
to prevent accidents or mitigate the consequences of an accident.
Safety margins and analytical conservatisms have been evaluated
and have been found acceptable. The analyzed event has been
evaluated and margin has been retained to ensure that the analysis
adequately bounds the postulated FHA event. The dose consequences
resulting from analyzing the FHA design basis accident comply with
the requirements of 10 CFR 50.67 and the guidance of RG 1.183.
The proposed changes continue to ensure that the doses at the
Exclusion Area Boundary (EAB) and Low Population Zone (LPZ)
boundary, as well as the Main Control Room (MCR), remain within
corresponding regulatory limits.
Based on the above, Exelon concludes that 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: J. Bradley Fewell, Esquire, Vice President
and Deputy General Counsel, Exelon Generation Company, LLC, 200 Exelon
Way, Kennett Square, PA 19348.
Acting NRC Branch Chief: Robert G. Schaaf.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit 2, (BVPS-2) Beaver County,
Pennsylvania
Date of amendment request: June 2, 2014, as supplemented by letter
dated August 8, 2014. Publicly-available versions are in ADAMS under
Accession Nos. ML14153A388, and ML14223A540, respectively.
Description of amendment request: The amendment would change the
BVPS-2 technical specifications (TSs). Specifically, the proposed
license amendment would revise TS 4.3.2, ``Drainage,'' to correct the
minimum drain elevation for the spent fuel storage pool specified in
the TS. In accordance with 10 CFR Part 50, Appendix B, Section XVI,
``Corrective Action,'' the proposed amendment is required to resolve a
TS discrepancy regarding an existing plant design feature.
Basis for proposed no significant hazards consideration
determination:
[[Page 58817]]
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.
Previously evaluated accidents including a fuel handling
accident and spent fuel cask drop accident are not affected by the
proposed amendment. Reducing the minimum water level above fuel
stored in the spent fuel storage pool in the event of inadvertent
draining as proposed would not involve a significant increase in the
probability of a previously evaluated accident. Maloperation or
passive piping failure causing inadvertent draining of the spent
fuel storage pool is not postulated concurrent with the fuel
handling or spent fuel cask drop accident. The proposed amendment
would not result in any failure modes that could initiate an
analyzed accident, and does not increase the likelihood of a
malfunction of a system, structure or component; therefore, the
probability of analyzed accidents is not affected.
There are no changes to how the station will be operated,
limiting conditions for operation, or limiting safety system
settings. The proposed amendment does not affect the capability of a
system, structure or component to perform a design function. Since
design functions are not affected by the proposed amendment, the
consequences of previously evaluated accidents 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.
Reducing the minimum water level above fuel stored in the spent
fuel storage pool in the event of inadvertent draining as proposed
does not create any new failure mechanisms, malfunctions, or
accident initiators and does not change design functions or system
operation in a way that affects the ability of systems, structures,
and components to perform design functions.
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.
General Design Criterion 61, ``Fuel storage and handling and
radioactivity control,'' of 10 CFR 50, Appendix A, states in part
that fuel storage and handling systems shall be designed with
suitable shielding for radiation protection.
The proposed change involves a reduction in the minimum
elevation of piping and penetrations of the spent fuel storage pool
specified in the Technical Specifications. In the event maloperation
or passive piping failure causes inadvertent draining of the spent
fuel storage pool, the remaining water level in the pool ensures the
stored fuel remains covered, provides adequate shielding for
personnel, and affords adequate assurance of safety when judged
against the current regulatory standard of General Design Criterion
61.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear
Operating Company, FirstEnergy Corporation, 76 South Main Street,
Akron, OH 44308.
Acting NRC Branch Chief: Robert G. Schaaf.
FirstEnergy Nuclear Operating Company (FENOC), Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Perry, OH
Date of amendment request: June 23, 2014. A publicly-available
version is in ADAMS under Accession No. ML14174A633.
Description of amendment request: The proposed amendment updates
the technical specification (TS) pressure and temperature (P/T) figures
using an NRC approved methodology to adjust the P/T limit curves for
previously missing data, addresses the reactor coolant system (RCS)
vacuum condition that can occur under certain conditions, and aligns
the heatup/cooldown requirements of the TS with the limits in the
associated P/T figures. Additionally editorial changes are proposed
related to the P/T figures including clarifications and updates to the
associated titles, labeling, and notes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The P/T [pressure and temperature] limits define RCS [reactor
coolant system] operational limits to avoid encountering pressure,
temperature, and temperature rate of change conditions that reduce
safety margins with respect to nonductile brittle failure of the
reactor coolant pressure boundary (RCPB). The figures are not
accident initiators or accident mitigating features, but preclude
operation in an unanalyzed condition.
This proposed amendment does not change the design function of
the RCS or RCPB and does not change the way the plant is maintained
or operated when using the P/T limit curves. This proposed amendment
does not affect any plant systems that are accident initiators and
does not affect any accident mitigating feature.
The proposed amendment does not affect the operability
requirements for the RCS, as verification of operating within the P/
T limits will continue to be performed, as required. Compliance with
and continued verification of the P/T limits support the capability
of the RCS to perform its required design functions, consistent with
the plant safety analyses.
Changing the figures will not change any of the dose analyses
associated with the USAR [updated safety analysis report] Chapter 15
accidents because they do not affect the source term, containment
isolation or radiological release assumptions used in any accident
previously evaluated. Plant accident mitigation functions and
requirements remain unchanged.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The P/T limits define RCS operational parameters to protect the
RCPB and are not accident initiators or accident mitigating
features. The limits are conservatively calculated using an NRC
approved methodology. This proposed amendment does not change the
design function of the RCS or RCPB, and does not change the way the
plant is operated or maintained. This proposed amendment does not
affect any plant systems that are accident initiators, does not
affect any accident mitigating feature, and does not create a new or
different kind of accident.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The P/T limits define RCS operational parameters, which are
established to protect the reactor vessel. The analysis supporting
the curve changes utilize methods previously reviewed and approved
by the NRC.
Margin of safety is related to the ability of the fission
product barriers (fuel cladding, reactor coolant system, and primary
containment) to perform their design functions during and following
postulated accidents. This proposed amendment does not directly
involve or physically affect fuel cladding or the primary
containment.
The amendment request proposes to update the P/T limit figures
using an NRC approved methodology. The curves maintain the margin of
safety for RCPB materials that are exposed to neutron radiation.
[[Page 58818]]
The proposed amendment does not involve a physical change to the
plant, does not change methods of plant operation within prescribed
limits, and does not change methods of maintenance on equipment
important to safety. Therefore, the proposed amendment does not
involve a significant reduction in a margin of safety.
Based on the responses to the three questions above, FENOC
[FirstEnergy Nuclear Operating Company] concludes that the proposed
amendment does not involve a significant hazards consideration under
the standards set forth in 10 CFR 50.92(c), and, accordingly, a
finding of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop. A-GO-15, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Travis L. Tate.
Florida Power and Light Company, et al. (FPL), Docket Nos. 50-335 and
50-389, St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida
Date of amendment request: June 9, 2014. A publicly-available
version is in ADAMS under Accession No. ML14175A121.
Description of amendment request: The amendments would revise
Technical Specification (TS) 6.2, Organization, specifically TS
6.2.2.e. to allow the station technical assistant (STA) position to be
manned by a single STA, a shift supervisor who meets the qualifications
for the STA, or an individual with a senior reactor operator's license
who meets the qualifications for the STA on each unit in MODES 1, 2, 3,
or 4. This criterion was omitted from FPL's license amendment request
dated July 26, 2013 (ADAMS Accession No. ML13219A840), that addressed
shift staffing requirements. As a result, it was omitted from the
corresponding license amendments dated February 7, 2014 (ADAMS
Accession No. ML14016A248). This criterion was previously approved by
the NRC and incorporated into the St. Lucie Units 1 and 2 TSs by
Amendment Nos. 173 and 113, respectively.
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 define the method for manning the shift
technical advisor (STA) position and do not reduce the unit staffing
requirements. In addition, the changes correct a typographical
error. The changes do not affect the minimum shift compliment in any
mode of operation nor decrease the effectiveness of shift personnel.
The STA position will continue to be manned by qualified personnel.
The proposed changes are administrative and editorial in nature and
will not result in any significant increase in the probability of
consequences of an accident as previously evaluated. Further, the
proposed changes do not introduce additional risk or greater
potential for consequences of an accident that has not previously
been evaluated. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes define the method for manning the shift
technical advisor position and do not reduce the unit staffing
requirements. In addition, the changes correct a typographical
error. The proposed changes are administrative and editorial in
nature. No new or different type of equipment will be installed. The
proposed changes will not introduce new failure modes/effects that
could lead to an accident for which consequences exceed that of
accidents previously analyzed. 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 define the method for manning the STA
position and do not reduce the unit staffing requirements. In
addition, the changes correct a typographical error. The changes do
not affect the minimum shift compliment in any mode of operation nor
decrease the effectiveness of shift personnel. The STA position will
continue to be manned by qualified personnel. The proposed changes
will not involve a significant reduction in a margin of safety in
that the changes are administrative and editorial in nature. No
plant equipment or accident analyses will be affected. Additionally,
the proposed changes will not relax any criteria used to establish
safety limits, safety system settings, or the bases for any limiting
conditions for operation. Safety analysis acceptance criteria are
not affected. Plant operation will continue within the design basis.
The proposed changes do not adversely affect systems that
respond to safely shutdown the plant, and maintain the plant in a
safe shutdown condition. Consequently, 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light, 700 Universe Blvd., MS LAW/JB, Juno
Beach, Florida 33408-0420.
Acting NRC Branch Chief: Lisa M. Regner.
Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant,
Unit 2, St. Lucie County, Florida
Date of amendment request: January 30, 2014. A publicly-available
version is in ADAMS under Accession No. ML14049A284.
Description of amendment request: The amendment would revise the
Technical Specification (TS) surveillance requirements (SRs) for
snubbers to conform to revisions to the Snubber Testing Program
allowing a year extension to the existing interval for the snubber
program transition. This revision would meet the requirements of the
Operation and Maintenance (OM) Code and Subsection ISTD, ``Preservice
and Inservice Examination and Testing of Dynamic Restraints (Snubbers)
in Light Water Reactor Nuclear Power Plants,'' of the American Society
of Mechanical Engineers OM Code, 2004 Edition with 2005 and 2006
Addenda.
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 would revise SR 4.7.9 to conform the TS to the
revised surveillance program for snubbers. Snubber examination, testing
and service life monitoring will continue to meet the requirements of
10 CFR 50.55a(g). Snubber examination, testing and service life
monitoring is not an initiator of any accident previously evaluated.
Therefore, the probability of an accident previously evaluated is not
significantly increased. Snubbers will continue to be demonstrated
OPERABLE by performance of a
[[Page 58819]]
program for examination, testing and service life monitoring in
compliance with 10 CFR 50.55a or authorized alternatives. The proposed
change to TS ACTION 3.7.9 for inoperable snubbers is administrative in
nature and is required for consistency with the proposed change to SR
4.7.9. The proposed change does not adversely affect plant operations,
design functions or analyses that verify the capability of systems,
structures, and components to perform their design functions therefore,
the consequences of accidents previously evaluated are not
significantly increased. Therefore, it is concluded that this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve any physical alteration of
plant equipment. The proposed changes do not alter the method by which
any safety-related system performs its function. As such, no new or
different types of equipment will be installed, and the basic operation
of installed equipment is unchanged. The methods governing plant
operation and testing remain consistent with current safety analysis
assumptions. Therefore, it is concluded that this change does not
create the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes ensure snubber examination, testing and
service life monitoring will continue to meet the requirements of 10
CFR 50.55a(g). Snubbers will continue to be demonstrated OPERABLE by
performance of a program for examination, testing and service life
monitoring in compliance with 10 CFR 50.55a or authorized alternatives.
The proposed change to TS ACTION 3.7.9 for inoperable snubbers is
administrative in nature and is required for consistency with the
proposed change to SR 4.7.9. Therefore, it is concluded that the
proposed change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light, 700 Universe Blvd., MS LAW/JB, Juno
Beach, Florida 33408-0420.
Acting NRC Branch Chief: Lisa M. Regner.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station (CNS), Nemaha County, Nebraska
Date of amendment request: July 14, 2014. A publicly-available
version is in ADAMS under Accession No. ML14202A205.
Description of amendment request: The proposed amendment would
delete Technical Specification 5.5.3, ``Post Accident Sampling,''
thereby eliminating the program requirements to have and maintain the
post-accident sampling system. The changes are consistent with NRC-
approved Industry/Technical Specification Task Force (TSTF) Standard
Technical Specification Change Traveler, TSTF-413, ``Elimination of
Requirements for a Post Accident Sampling System (PASS).'' The
availability of this technical specification improvement was announced
in the Federal Register on March 20, 2002, as part of the consolidated
line item improvement process. CNS will continue to have the ability to
obtain samples, utilizing PASS, following an accident.
Basis for proposed no significant hazards consideration
determination: The licensee stated in its application that it reviewed
the proposed no significant hazards consideration determination
published on December 27, 2001 (66 FR 66949), as part of the
consolidated line item improvement process. The licensee stated that it
concluded that the proposed determination presented in the notice is
applicable to CNS and the determination is incorporated by reference to
satisfy the requirements of 10 CFR 50.91(a). As required by 10 CFR
50.91(a), an analysis of the issue of no significant hazards
consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 [Three Mile Island, Unit 2] accident. The specific
intent of the PASS was to provide a system that has the capability
to obtain and analyze samples of plant fluids containing potentially
high levels of radioactivity, without exceeding plant personnel
radiation exposure limits. Analytical results of these samples would
be used largely for verification purposes in aiding the plant staff
in assessing the extent of core damage and subsequent offsite
radiological dose projections. The system was not intended to and
does not serve a function for preventing accidents and its
elimination would not affect the probability of accidents previously
evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
[[Page 58820]]
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the analysis and, based on this review,
it appears that the three standards of 10 CFR 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
Acting NRC Branch Chief: Eric R. Oesterle.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: July 17, 2014. A publicly-available
version is in ADAMS under Accession No. ML14203A045.
Description of amendment request: The proposed amendment would move
the Linear Heat Generation Rate (LHGR) and Single Loop Operation LHGR
Limit from the Technical Requirements Manual (TRM) to the Technical
Specifications (TS).
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.
LHGR limits have been defined to provide sufficient margin
between the steady-state operating condition and any fuel damage
condition to accommodate uncertainties and to assure that no fuel
damage results even during the worst anticipated transient condition
at any time. The proposed change to move the LHGR limits from the
TRM to TS, including the change to TS 3.4.1, Recirculation Loops
Operating, and TS 3.7.7, Main Turbine Bypass System, does not modify
the limits, change assumptions for the accident analysis, or change
operation of the station.
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 modify the limits, change
assumptions for the accident analysis, or change operation of the
station.
The proposed change does move LHGR limits that have been defined
to provide sufficient margin between the steady-state operating
condition and any fuel damage condition to accommodate uncertainties
and to assure that no fuel damage results even during the worst
anticipated transient condition at any time from the TRM to TS.
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 proposed change to move the LHGR limits from the TRM to TS,
including the change to TS 3.4.1 and TS 3.7.7, does not modify the
limits, change assumptions for the accident analysis, or change
operation of the station.
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: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
Acting NRC Branch Chief: Eric R. Oesterle.
NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold
Energy Center, Linn County, Iowa
Date of amendment request: June 23, 2014. A publicly-available
version is in ADAMS under Accession No. ML14175B387.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) requirements to address NRC Generic
Letter (GL) 2008-01, ``Managing Gas Accumulation in Emergency Core
Cooling, Decay Heat Removal, and Containment Spray Systems,'' as
described in Technical Specification Task Force (TSTF) Change Traveler
TSTF-523, Revision 2, ``Generic Letter 2008-01, Managing Gas
Accumulation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the Proposed Change Involve a Significant Increase in
the Probability or Consequences of an Accident Previously Evaluated?
Response: No.
The proposed change revises or adds SRs [Surveillance
Requirements] that require verification that the Emergency Core
Cooling Systems (ECCS), Residual Heat Removal (RHR) System, and the
Reactor Core Isolation Cooling (RCIC) System are not rendered
inoperable due to accumulated gas and to provide allowances which
permit performance of the revised verification. Gas accumulation in
the subject systems is not an initiator of any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The proposed SRs ensure
that the subject systems continue to be capable to perform their
assumed safety function and are not rendered inoperable due to gas
accumulation. Thus, the consequences of any accident previously
evaluated are not significantly increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the Proposed Change Create the Possibility of a New or
Different Kind of Accident from any Accident Previously Evaluated?
Response: No.
The proposed change revises or adds SRs that require
verification that the ECCS, RHR System, and RCIC System are not
rendered inoperable due to accumulated gas and to provide allowances
which permit performance of the revised verification. The proposed
change does not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) or a change in
the methods governing normal plant operation. In addition, the
proposed change does not impose any new or different requirements
that could initiate an accident. The proposed change does not alter
assumptions made in the safety analysis and is consistent with the
safety analysis assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the Proposed Change Involve a Significant Reduction in a
Margin of Safety?
Response: No.
The proposed change revises or adds SRs that require
verification that the ECCS, RHR System, and RCIC System are not
rendered inoperable due to accumulated gas and to
[[Page 58821]]
provide allowances which permit performance of the revised
verification. The proposed change adds new requirements to manage
gas accumulation in order to ensure that the subject systems are
capable of performing their assumed safety functions. The proposed
SRs are more comprehensive than the current SRs and will ensure that
the assumptions of the safety analysis are protected. The proposed
change does not adversely affect any current plant safety margins or
the reliability of the equipment assumed in the safety analysis.
Therefore, there are no changes being made to any safety analysis
assumptions, safety limits, or limiting safety system settings that
would adversely affect plant safety as a result of the proposed
change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. James Petro, P. O. Box 14000 Juno Beach,
FL 33408-0420.
NRC Branch Chief: David L. Pelton.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, (Seabrook) Rockingham County, New Hampshire
Date of amendment request: July 24, 2014. A publicly-available
version is in ADAMS under Accession No. ML14209A918.
Description of amendment request: The proposed amendment would
revise Seabrook Technical Specifications (TSs) by increasing the
voltage limit for a full load rejection test of the emergency diesel
generator specified in surveillance requirement 4.8.1.1.2.f.3 of TS
3.8.1.1, ``A.C. [alternating current] Sources--Operating.'' The
proposed amendment also revises the TS definition of the terms
``Operable--Operability.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, with NRC staff revisions provided in [brackets], which
is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to increase in the [emergency diesel
generator] EDG full load rejection overvoltage limit from 4784
[volts] V to 4992V is not an accident initiator. The overvoltage
transient is an expected response to a full load rejection. The
magnitude and duration of the proposed overvoltage limit have been
considered and determined to have no detrimental effects on the
connected equipment that is exposed to the voltage transient. The
proposed change does not affect the EDG design function or how the
EDG is operated. Since the EDG is not impacted, the EDG remains
capable of performing its intended design function of supplying
power to emergency safeguards equipment. The proposed change to the
definition of operable--operability is administrative in nature and
does not alter the meaning of the defined terms.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to revise the definition of the terms
operable--operability and to increase the EDG full load rejection
overvoltage limit from 4784V to 4992V are not accident initiators.
The overvoltage transient is an expected response to a full load
rejection. The magnitude and duration of the proposed overvoltage
limit have been considered and determined to have no detrimental
effects on the connected equipment that is exposed to the voltage
transient. The proposed changes do not introduce any new failure
modes.
The changes do not involve a physical alteration to the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods for operating the plant. The proposed changes
do not affect the EDG design function or how the EDG is operated.
Since the EDG is not impacted, the EDG remains capable of performing
its intended design function of supplying power to emergency
safeguards equipment. The change to the definition of operable--
operability makes grammatical corrections and adds clarity but makes
no change to the meaning of the terms.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to increase in the EDG full load rejection
overvoltage limit from 4784V to 4992V has been evaluated with
consideration of the effect on the EDG and connected equipment that
would be exposed to the higher voltage transient. Based on review of
equipment specifications, test data, and manufacturer's input, it
was concluded that there would be no detrimental effects to the EDG
or connected equipment that is exposed to the higher voltage
transient. The EDG remains capable of performing its intended design
function of supplying power to emergency safeguards equipment.
The proposed change to the definition of operable--operability
is administrative in nature and does not alter any criterion used to
establish operability of plant structure, systems, or components.
The proposed amendment does not involve changes to any safety
analyses assumptions, safety limits, or limiting safety system
settings. The changes do not adversely impact plant operating
margins or the reliability of equipment credited in the safety
analyses.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William Blair, Managing Attorney, Florida
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
Acting NRC Branch Chief: Robert G. Schaaf.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, (Seabrook) Rockingham County, New Hampshire
Date of amendment request: July 24, 2014. A publicly-available
version is in ADAMS under Accession No. ML14209A919.
Description of amendment request: The proposed amendment would
revise the Seabrook Technical Specifications (TS). The proposed change
modifies TS 3.3.3.1, ``Radiation Monitoring for Plant Operations,'' to
eliminate duplicate requirements, resolve an inconsistency, and correct
a deficiency.
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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The instruments involved with the proposed changes to the
technical specifications (TS) are not initiators of any accidents
previously evaluated, and the probability and consequences of
accidents previously evaluated are unaffected by the proposed
changes. There is no change to any equipment response or accident
scenario, and the changes impose no additional challenges to fission
product barrier integrity. The proposed changes do not alter the
design, function, operation, or configuration of any plant
structure, system, or component (SSC). As a result, the outcomes of
accidents previously evaluated are unaffected. The
[[Page 58822]]
proposed changes modify the TS to eliminate duplicate requirements,
resolve an inconsistency, and correct a deficiency.
Therefore, the proposed changes do not result in a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
The changes do not challenge the integrity or performance of any
safety-related systems. No plant equipment is installed or removed,
and the changes do not alter the design, physical configuration, or
method of operation of any plant SSC. No physical changes are made
to the plant, so no new causal mechanisms are introduced.
Therefore, the proposed changes to the TS 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 ability of any operable SSC to perform its designated safety
function is unaffected by the proposed changes. The proposed changes
do not alter any safety analyses assumptions, safety limits,
limiting safety system settings, or method of operating the plant.
The changes do not adversely impact plant operating margins or the
reliability of equipment credited in the safety analyses.
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 Blair, Managing Attorney, Florida
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
Acting NRC Branch Chief: Robert G. Schaaf.
NextEra Energy Seabrook LLC, Docket No. 50-443, Seabrook Station, Unit
1, Rockingham County, New Hampshire
Date of amendment request: July 24, 2014. A publicly-available
version is in ADAMS under Accession No. ML14216A404.
Description of amendment request: The proposed amendment would
incorporate revised reactor coolant system (RCS) pressure-temperature
limits in the Technical Specification (TS) applicable to 55 effective
full-power years. The change will also provide new overpressure
protection setpoints and lower the RCS temperature at which the TS is
applicable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the Technical Specifications (TS) do not
impact the physical function of plant structures, systems, or
components (SSCs) or the manner in which SSCs perform their design
function. Operation in accordance with the proposed TS will ensure
that all analyzed accidents will continue to be mitigated by the
SSCs as previously analyzed. The proposed changes do not alter or
prevent the ability of operable SSCs to perform their intended
function to mitigate the consequences of an initiating event within
assumed acceptance limits. The proposed changes neither adversely
affect accident initiators or precursors, nor alter design
assumptions.
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 involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed), create new failure modes for existing equipment, or
create any new limiting single failures. The changes to the
pressure--temperature limits, power operated relief valve setpoints,
and the over pressure protection system effective temperature will
continue to ensure that appropriate fracture toughness margins are
maintained to protect against reactor vessel failure, during both
normal and low temperature operation. The proposed changes are
consistent with the applicable NRC approved methodologies (i.e.,
WCAP-14040, Rev. 4 and ASME Code Case N-641). Plant operation will
not be altered, and all safety functions will continue to perform as
previously assumed in accident analyses.
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
the 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 will not
adversely affect the operation of plant equipment or the function of
any equipment assumed in the accident analysis. The proposed changes
were developed using NRC approved methodologies and will continue to
ensure an acceptable margin of safety is maintained. The safety
analysis acceptance criteria are not affected by this change. The
proposed changes will not result in plant operation in a
configuration outside the design basis. The proposed changes do not
adversely affect systems that respond to safely shutdown the plant
and to maintain the plant in a safe shutdown condition.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William Blair, Managing Attorney, Florida
Power & Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
Acting NRC Branch Chief: Robert G. Schaaf.
Northern States Power Company--Minnesota (NSPM), Docket No. 50-263,
Monticello Nuclear Generating Plant (MNGP), Wright County, Minnesota
Date of amendment request: June 17, 2014. A publicly-available
version is in ADAMS under Accession No. ML14168A486.
Description of amendment request: The NSPM proposes to revise MNGP
Technical Specification (TS) 3.5.1, ``ECCS [Emergency Core Cooling
System]--Operating,'' to correct the requirements for the Alternate
Nitrogen System pressure. TS Surveillance Requirement (SR) 3.5.1.3
requires verification of limits for automatic depressurization system
(ADS) pneumatic pressure for both ADS pneumatic supplies. The proposed
change would revise the TS SR 3.5.1.3.b pressure limit for determining
operability of the Alternate Nitrogen System from greater than or equal
to (>=) 410 pounds per square inch gauge (psig) to a corrected value of
>= 700 psig.
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 provided 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 SR for the purpose of
restoring a value to be consistent with the licensing basis. The
proposed TS change does not introduce new
[[Page 58823]]
equipment or new equipment operating modes, nor does the proposed
change alter existing system relationships. The proposed change does
not affect plant operation, design function or any analysis that
verifies the capability of a system, structure or component (SSC) to
perform a design function. Further, the proposed change does not
increase the likelihood of the malfunction of any SSC or impact any
analyzed accident. Consequently, the probability of an accident
previously evaluated is not affected and there is not 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.
There proposed change revises the TS SR for the purpose of
restoring a value to be consistent with the licensing basis. The
change does not involve a physical alteration to 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 for the
components supplied by the Alternate Nitrogen System. 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 in a
margin of safety?
Response: No.
The proposed change revises the TS SR for the purpose of
restoring a value to be consistent with the licensing basis. 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 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: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: David L. Pelton.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County,
California
Date of amendment request: July 28, 2014. A publicly-available
version is in ADAMS under Accession No. ML14209B074.
Description of amendment request: The proposed amendments would
modify the technical specifications (TS) to risk-inform requirements
regarding selected Required Action End States. The proposed changes to
the Required Action End States are described in Table 1 of the
Enclosure to the licensee's letter dated July 28, 2014. The changes are
consistent with Technical Specification Task Force (TSTF) Traveler
TSTF-432, Revision 1, ``Change in Technical Specifications End States
(WCAP-16294)'' (ADAMS Accession No. ML103430249).
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 modifies the end state (e.g., mode or other
specified condition) which the Required Actions specify must be
entered if compliance with the Limiting Conditions for Operation
(LCO) is not restored. The requested Technical Specifications (TS)
permit an end state of Mode 4 rather than an end state of Mode 5
contained in the current TS. In some cases, other Conditions and
Required Actions are revised to implement the proposed change.
Required Actions are not an initiator of any accident previously
evaluated. Therefore, the proposed change does not affect the
probability of any accident previously evaluated. The affected
systems continued to be required to be operable by the TS and the
Completion Times specified in the TS to restore equipment to
operable status or take other remedial Actions remain unchanged.
WCAP-16294-NP-A, Revision 1, ``Risk-Informed Evaluation of Changes
to [Technical Specification] Required Action Endstates for
Westinghouse NSSS [Nuclear Steam Supply System] PWRs [Pressurized
Water Reactors],'' demonstrates that the proposed change does not
significantly increase 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 accident from any accident previously evaluated?
Response: No.
The proposed change modifies the end state (e.g., mode or other
specified condition) which the Required Actions specify must be
entered if compliance with the LCO is not restored. In some cases,
other Conditions and Required Actions are revised to implement the
proposed change. The change does not involve a physical alteration
of the plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the change does not impose any new
requirements. The change does not alter assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different 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 modifies the end state (e.g., mode or other
specified condition) which the Required Actions specify must be
entered if compliance with the LCO is not restored. In some cases,
other Conditions and Required Actions are revised to implement the
proposed change. Remaining within the Applicability of the LCO is
acceptable because WCAP-16294-NP-A demonstrates that the plant risk
in MODE 4 is similar to, or lower than, MODE 5. As a result, no
margin of safety is significantly 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 requests involve no significant hazards consideration.
Attorney for licensee: Jennifer Post, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
Acting NRC Branch Chief: Eric R. Oesterle.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028,
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: June 12, 2014. A publicly-available
version is in ADAMS under Accession No. ML14164A098.
Description of amendment request: The proposed license amendment
request (LAR) proposes to revise Plant Specific Tier 2* material within
the Updated Final Safety Analysis Report (UFSAR) by making editorial
and consistency corrections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the
[[Page 58824]]
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 editorial and consistency update does not involve a
technical change, i.e., there is no design parameter or requirement,
calculation, analysis, function, or qualification change. No
structure, system, component (SSC), design, or function would be
adversely affected. No design or safety analysis would be adversely
affected. The proposed changes do not adversely affect any accident
initiating event or component failure, thus the probabilities of the
accidents previously evaluated are not adversely affected. No
function used to mitigate a radioactive material release and no
radioactive material release source term is involved, thus the
radiological releases in the accident analyses are not adversely
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 editorial and consistency update would not affect
the design or function of any SSC, but will instead provide
consistency between the SSC designs and functions and the
discussions currently presented in the UFSAR via Tier 2*
information. The proposed nontechnical changes would not introduce a
new failure mode, fault, or sequence of events that could result in
a radioactive material release.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed editorial and consistency update is nontechnical
and thus would not affect any design parameter, function, or
analysis. There would be no change to an existing design basis,
design function, regulatory criterion, or analyses. No safety
analysis or design basis acceptance limit/criterion is involved.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW, Washington, DC, 20004-2514.
NRC Branch Chief: Lawrence J. Burkhart.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028,
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: August 28, 2014. A publicly-available
version is in ADAMS under Accession No. ML14245A601.
Description of amendment request: The proposed license amendment
request (LAR) proposes to revise Tier 2* and Tier 2 information related
to the design details of connections in several locations between the
steel plate composite construction (SC) used for the shield building
and the standard reinforced concrete (RC) walls, floors, and roofs of
the auxiliary building and the lowers of the shield building.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The design function of the nuclear island structures is to
provide support, protection, and separation for the seismic Category
I mechanical and electrical equipment located in the nuclear island.
The nuclear island structures are structurally designed to meet
seismic Category I requirements as defined in Regulatory Guide 1.29.
The changes to the detail design of connections between the RC
and SC structures do not have an adverse impact on the response of
the nuclear island structures to safe shutdown earthquake ground
motions or loads due to anticipated transients or postulated
accident conditions. The changes to the detail design do not impact
the support, design, or operation of mechanical and fluid systems.
There is no change to plant systems or the response of systems to
postulated accident conditions. There is no change to the predicted
radioactive releases due to postulated accident conditions. The
plant response to previously evaluated accidents or external events
is not adversely affected, nor do the changes described create any
new accident precursors.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes are to the detail design of connections
between the RC and SC structures. The changes to the detail design
of connections do not change the criteria and requirements for the
design and analysis of the nuclear island structures. The changes to
the detail design of connections do not change the design function,
support, design, or operation of mechanical and fluid systems. The
changes to the detail design of connections do not change the
methods used to connect the RC to SC. The changes of the detail
design of connections do not result in a new failure mechanism for
the nuclear island structures or new accident precursors. As a
result, the design functions of the nuclear island structures are
not adversely affected by the proposed changes.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
No safety analysis or design basis acceptance limit/criterion is
challenged or exceeded by the proposed changes; and 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 Avenue NW, Washington, DC, 20004-2514.
NRC Branch Chief: Lawrence J. Burkhart.
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: July 29, 2014. A publicly-available
version is available in ADAMS under Accession No. ML14210A646.
Description of amendment request: The proposed license amendment
request would revise the Combined Licenses (COLs) with regard to Tier 1
material and promote consistency with the Updated Final Safety Analysis
Report Tier 2.
Nuclear Operating Company has also requested an exemption from the
provisions of 10 CFR part 52, appendix D, section III.B, ``Design
Certification Rule for the AP1000 Design, Scope and Contents,'' to
allow a departure from the elements of the certification information in
Tier 1 of the generic Design Control Document.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 58825]]
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
1. Does the requested amendment involve a significant increase
in the probability or consequences of an accident previously
evaluated?
Response: No
The proposed editorial and consistency COL Appendix C and
corresponding plant-specific Tier 1 update does not involve a
technical change, e.g., there is no design parameter or requirement,
calculation, analysis, function or qualification change. No
structure, system, or component (SSC) design or function would be
affected. No design or safety analysis would be affected. The
proposed changes do not affect any accident initiating event or
component failure, thus the probabilities of the accidents
previously evaluated are not affected. No function used to mitigate
a radioactive material release and no radioactive material release
source term is involved, thus the radiological releases in the
accident analyses are not affected.
Therefore, the requested amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the requested amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No
The proposed editorial and consistency COL Appendix C and
corresponding plant-specific Tier 1 update would not affect the
design or function of any SSC, but will instead provide consistency
between the SSC designs and functions currently presented in the
UFSAR, COL Appendix C, and the Tier 1 information. The proposed
changes would not introduce a new failure mode, fault or sequence of
events that could result in a radioactive material release.
Therefore, the proposed amendment does not create the possibility of
a new or different kind of accident.
Therefore, the requested amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No
The proposed editorial and consistency COL Appendix C and
corresponding plant-specific Tier 1 update would not affect the
design or function of any SSC, but will instead provide consistency
between the SSC designs and functions currently presented in the
UFSAR, COL Appendix C, and the Tier 1 information. The proposed
changes would not introduce a new failure mode, fault or sequence of
events that could result in a radioactive material release.
Therefore, the requested amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Blach & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence Burkhart.
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: July 30, 2014. A publicly-available
version is in ADAMS under Accession No. ML14211A666.
Description of amendment request: The proposed license amendment
request would revise the combined licenses (COLs) with regard to Tier
2* material within the Updated Final Safety Analysis Report (UFSAR) to
resolve inconsistencies with other Tier 2* information elsewhere in the
UFSAR.
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 requested amendment involve a significant increase
in the probability or consequences of an accident previously
evaluated?
Response: No.
The proposed editorial and consistency update does not involve a
technical change, i.e., there is no design parameter or requirement,
calculation, analysis, function, or qualification change. No
structure, system, or component, design, or function would be
adversely affected. No design or safety analysis would be adversely
affected. The proposed changes do not adversely affect any accident
initiating event or component failure, thus the probabilities of the
accidents previously evaluated are not adversely affected. No
function used to mitigate a radioactive material release and no
radioactive material release source term is involved, thus the
radiological releases in the accident analyses are not adversely
affected.
Therefore, the requested amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the requested amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed editorial and consistency update would not affect
the design or function of any structure, system, or component, but
will instead provide consistency between the structure, system, and
component designs and functions and the discussions currently
presented in the UFSAR via Tier 2* information. The proposed non-
technical changes would not introduce a new failure mode, fault, or
sequence of events that could result in a radioactive material
release.
Therefore, the requested amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed editorial and consistency update is non-technical
and thus would not affect any design parameter, function, or
analysis. There would be no change to an existing design basis,
design function, regulatory criterion, or analyses. No safety
analysis or design basis acceptance limit/criterion is involved.
Therefore, the requested amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. M. Stanford Blanton, Blach & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence Burkhart.
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: April 11, 2014. A publicly-available
version is in ADAMS under Accession No. ML14101A459. The amendment
request was supplemented by letter dated April 18, 2014. A publicly-
available version is in ADAMS under Accession No. ML14108A093. The
amendment request was further supplemented by two letters dated August
28, 2014. Publicly-available versions of the two letters are in ADAMS
under Accession Nos. ML14241A250 and ML14241A264.
Description of amendment request: The license amendment request was
originally noticed in the Federal Register on June 6, 2014 (79 FR
32771). This notice is being reissued in its entirety to include the
revised analysis of the issue of no significant hazards consideration
submitted by the licensee in its August 28, 2014, submission (ADAMS
Accession No. ML14241A250). The proposed license amendment request
would depart from the plant-specific Design Control Document Tier 1 and
Tier 2 material to describe modifications to increase the efficiency of
the return of condensate utilized by the passive core cooling system to
the
[[Page 58826]]
in-containment refueling water storage tank to support the capability
for long term cooling.
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 containment condensate flow path changes provide
sufficient condensate return flow to maintain In-containment
Refueling Water Storage Tank (IRWST) level above the top of the
Passive Residual Heat Removal Heat Exchanger (PRHR HX) tubes long
enough to prevent PRHR HX performance degradation from that
considered in the UFSAR [Updated Final Safety Analysis Report]
Chapter 15 safety analyses. The added components are seismically
qualified and constructed of only those materials appropriately
suited for exposure to the reactor coolant environment as described
in UFSAR Section 6.1. No aluminum is permitted to be used in the
construction of these components so that they do not contribute to
hydrogen production in containment.
The proposed changes clarify the design basis for the PRHR HX,
which removes decay heat from the Reactor Coolant System (RCS)
during a non-loss of coolant accident (non-LOCA). With operator
action to avoid unnecessary Automatic Depressurization System (ADS)
actuation based on RCS conditions, PRHR HX operation can be extended
longer than would be maintained automatically by the protection
system. Though analysis shows significantly greater capacity, the
extent of the capability of the PRHR HX would be changed from
operating indefinitely to operating for at least 72 hours. If PRHR
HX capability were exhausted after 72 hours, the ADS would be
actuated, which could result in significant containment floodup.
However, probabilistic analysis shows the probability of design
basis containment floodup after PRHR HX operation during a non-LOCA
event is significantly lower than the probability of a small break
LOCA, for which comparable containment floodup is anticipated.
Therefore, the probability of significant containment floodup is
not increased.
The proposed changes do not affect any components whose failure
could initiate a previously evaluated event, thus the probabilities
of the accidents previously evaluated are not affected. The affected
equipment does not adversely affect or interact with safety-related
equipment or another radioactive material barrier. The proposed
changes clarify the post-accident performance requirements for the
PRHR HX. However, the proposed changes do not prevent the engineered
safety features from performing their safety-related accident
mitigating functions. The radioactive material source terms and
release paths used in the safety analyses are unchanged, thus the
radiological releases in the UFSAR accident analyses are not
affected.
Therefore, the proposed amendment does not involve an 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 long-term safe shutdown analysis results show that the PRHR
HX continues to meet its acceptance criterion, i.e., to cool the
Reactor Coolant System (RCS) to below 420[ordm]F in 36 hours. The
added equipment does not adversely interface with any component
whose failure could initiate an accident, or any component that
contains radioactive material. The modified components do not
incorporate any active features relied upon to support normal
operation. The downspout and gutter return components are
seismically qualified to remain in place and functional during
seismic and dynamic events. The containment condensate flow path
changes do not create a new fault or sequence of events that could
result in a radioactive material release.
The proposed change quantifies the duration that the PRHR HX is
capable of maintaining adequate core cooling, and specifies that if
PRHR HX cooling capability is exhausted, the ADS would be actuated.
This involves the possibility of opening the ADS valves after the
IRWST water level has decreased below the spargers, which promote
steam condensation in the IRWST. During this condition, the loads on
the IRWST, spargers and any internal structures or components in the
IRWST would still be less than their limiting loads, and these SSCs
would not be adversely affected or cause a different mode of
operation. Therefore, no new type of accident could be created by
this condition.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not reduce the redundancy or diversity
of any safety-related function. The added components are classified
as safety-related, seismically qualified, and are designed to comply
with applicable design codes. The proposed containment condensate
flow path changes provide sufficient condensate return flow to
maintain adequate IRWST water level for those events using the PRHR
HX cooling function. The long-term Shutdown Temperature Evaluation
results in UFSAR Chapter 19E show the PRHR HX continues to meet its
acceptance criterion. The UFSAR Chapters 6 and 15 analyses results
are not affected, thus margins to their regulatory acceptance
criteria are unchanged. The former design basis, which stated the
PRHR HX could bring the plant to 420[deg]F within 36 hours, is
changed to state the heat exchanger can establish safe, stable
conditions in the reactor coolant system after a design basis event.
Such safe stable conditions may not coincide with an RCS temperature
of 420[deg]F. However, the PRHR HX is able to bring the RCS to a
sufficiently low temperature such that RCS conditions would be
comparable to those achieved at 420[deg]F--peak cladding
temperatures, departure from nucleate boiling, and pressurizer level
would be maintained within acceptable limits of the evaluation
criteria. No safety analysis or design basis acceptance limit/
criterion is challenged or exceeded by the proposed changes, thus no
margin of safety is significantly reduced.
Therefore, the proposed amendment does not reduce 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: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence J. Burkhart.
Southern Nuclear Operating Company, Inc. Docket Nos.: 52-025 and 52-
026, Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke
County, Georgia
Date of amendment request: July 14, 2014. A publicly-available
version is in ADAMS under Accession No. ML14195A296.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-91 and NPF-92 for the VEGP, Units 3 and 4 to
modify the fire area fire barriers of the turbine building switchgear
rooms on Elevations 141'-3'' and 158'-7'' of the turbine building to
accommodate the revised layout of the low and medium voltage switchgear
and associated equipment. The proposed changes also provide an
editorial change to a fire area number. The requested amendment
requires changes to Updated Final Safety Analysis Report (UFSAR)
information, which include changes to plant-specific Tier 2*
information and changes to Tier 2 information that involve changes to
this plant-specific Tier 2* information.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
[[Page 58827]]
The proposed reconfiguration of the turbine building switchgear
rooms, the control system cabinet room, the new electrical equipment
room, and the associated heating, ventilation and air conditioning
(HVAC) room and the proposed editorial change would not adversely
affect any safety-related equipment or function. The modified
configuration will maintain the fire protection function (i.e.,
barrier) as evaluated in Updated Final Safety Analysis Report
(UFSAR) Appendix 9A, thus, the probability of a spread of a fire
from these areas is not significantly increased. The safe shutdown
fire analysis is not affected, and the fire protection analysis
results are not adversely affected. The proposed changes affect
nonsafety-related electrical switchgear and do not involve any
accident, initiating event, or component failure; thus, the
probabilities of the accidents previously evaluated are not
affected. The proposed changes do not interface with or affect any
system containing radioactivity or affect any radiological material
release source terms; thus, the radiological releases in the
accident analyses are not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to the fire zones in the turbine building
related to the turbine building switchgear rooms, the control system
cabinet room, the new electrical equipment room, the associated HVAC
room, and stairway will maintain the fire barrier fire protection
function as evaluated in the UFSAR Appendix 9A. The changes to the
fire areas and fire zones do not affect the function of any safety-
related structure, system, or component, and thus, do not introduce
a new failure mode. The affected turbine building areas and
equipment do not interface with any safety-related equipment or any
equipment associated with radioactive material and, thus, do not
create a new fault or sequence of events that could result in a new
or different kind of accident.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed reconfiguration of the fire zones associated with
the turbine building switchgear rooms, the electrical equipment
room, and the associated HVAC room and the proposed editorial change
will maintain the fire barrier fire protection function as evaluated
in the UFSAR Appendix 9A. The fire barriers and equipment in the
turbine building do not interface with any safety-related equipment
or affect any safety-related function. The changes to the area
barriers associated with the turbine building switchgear and
associated HVAC continue to comply with the existing design codes
and regulatory criteria, and do not affect any safety analysis.
Therefore, the proposed amendment 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: Mr. M. Stanford Blanton, Balch & Bingham
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Lawrence J. Burkhart.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: July 24, 2014. A publicly-available
version is in ADAMS under Accession No. ML14210A051.
Description of amendment request: The amendment would revise the
reactor coolant pump (RCP) flywheel inspection surveillance
requirements to extend the allowable inspection interval to 20 years,
consistent with Technical Specification Task Force (TSTF) Standard
Technical Specification Change Traveler, TSTF-421, ``Revision to RCP
Flywheel Inspection Program (WCAP-15666).'' The Nuclear Regulatory
Commission (NRC) staff believes that this amendment made use of both
TSTF-421 and TSTF-237, Revision 1, ``Relaxation of Reactor Coolant Pump
Flywheel Examinations.''
The NRC staff published a notice of opportunity for comment in the
Federal Register on June 24, 2003 (68 FR 37590), on possible amendments
adopting TSTF-421, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line-item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on October 22,
2003 (68 FR 60422). The licensee affirmed the applicability of the
model NSHC determination in its application dated July 24, 2014.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration determination is presented
below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change to the RCP flywheel examination frequency
does not change the response of the plant to any accidents. The RCP
will remain highly reliable and the proposed change will not result
in a significant increase in the risk of plant operation. Given the
extremely low failure probabilities for the RCP motor flywheel
during normal and accident conditions, the extremely low probability
of a loss-of-coolant accident (LOCA) with loss of offsite power
(LOOP), and assuming a conditional core damage probability (CCDP) of
1.0 (complete failure of safety systems), the core damage frequency
(CDF) and change in risk would still not exceed the NRC's acceptance
guidelines contained in Regulatory Guide (RG) 1.174 (<1.0E-6 per
year). Moreover, considering the uncertainties involved in this
evaluation, the risk associated with the postulated failure of an
RCP motor flywheel is significantly low. Even if all four RCP motor
flywheels are considered in the bounding plant configuration case,
the risk is still acceptably low.
The proposed change does not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, or configuration of the facility, or the manner in which
the plant is operated and maintained; alter or prevent the ability
of structures, systems, components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits or affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed change does not increase
the type or amount of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiations exposure. The proposed change is
consistent with the safety analysis assumptions and resultant
consequences.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from Any Accident Previously
Evaluated
The proposed change in flywheel inspection frequency does not
involve any change in the design or operation of the RCP. Nor does
the change to examination frequency affect any existing accident
scenarios, or create any new or different accident scenarios.
Further, the change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or alter the methods governing normal plant operation. In
addition, the change does not impose any new or different
requirements or eliminate any existing requirements, and does not
alter any assumptions made in the safety analysis. The proposed
change is consistent with the
[[Page 58828]]
safety analysis assumptions and current plant operating practice.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
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 acceptance criteria
are not impacted by this change. The proposed change will not result
in plant operation in a configuration outside of the design basis.
The calculated impact on risk is insignificant and meets the
acceptance criteria contained in RG 1.174. There are no significant
mechanisms for inservice degradation of the RCP flywheel.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the analysis and, based on this review,
it appears that the three standards of 10 CFR 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Jessie F. Quichocho.
Virginia Electric and Power Company, Docket Nos.: 50-280 and 50-281,
Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of amendment request: June 3, 2014. A publicly-available
version is in ADAMS under Accession No. ML14160A607.
Description of amendment request: The amendments would revise the
Surry Power Station (Surry) Units 1 and 2, Technical Specifications
(TS). Specifically, TS Figures 3.1-1 and 3.1-2, Surry, Units 1 and 2,
Reactor Coolant System Heatup Limitations and Surry, Units 1 and 2,
Reactor Coolant System Cooldown Limitations, respectively, are being
revised for clarification and to be fully representative of the
allowable operating conditions during Reactor Coolant System startup
and cooldown evolutions.
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 clarification of TS Figures 3.1-1 and 3.1-2 does
not involve a physical change to the plant and does not change the
manner in which plant systems or components are operated or
controlled. The proposed change does not alter or prevent the
ability of structures, system, and components (SSCs) to perform
their intended function to mitigate the consequences of an
initiating event within the assumed acceptance limits. The P/T
limits curves on TS Figures 3.1-1 and 3.1-2 are not being modified
and remain valid.
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 clarification of TS Figures 3.1-1 and 3.1-2 does
not involve any physical alteration of plant equipment;
consequently, no new or different types of equipment will be
installed. The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility. The P/T limits curves
on TS Figures 3.1-1 and 3.1-2 are not being modified, and the basic
operation of installed plant systems and components is unchanged.
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 existing RCS P/T limits curves on TS Figures 3.1-1 and 3.1-2
are not being modified. The proposed clarification of TS Figures
3.1-1 and 3.1-2 does not alter any plant equipment, does not change
the manner in which the plant is operated or controlled, and has no
impact on any safety analysis assumptions. The proposed change does
not alter the manner in which safety limits, limiting safety system
settings, or limiting conditions for operation are determined. The
proposed change does not result in plant operation in a
configuration outside the analyses or design basis and does not
adversely affect systems that respond to safely shut down the plant
and to maintain the plant in a safe shutdown condition.
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: Robert J. Pascarelli.
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.
Arizona Public Service Company, et al., Docket Nos.: STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1,
2, and 3, Maricopa County, Arizona
Date of application for amendment: September 27, 2013, as
supplemented by letter dated December 12, 2013.
Brief description of amendment: The amendments revised Technical
Specification (TS) 3.3.3, ``Control Element Assembly Calculators
(CEACs),'' to reinstate an inadvertently omitted 4-hour completion time
for
[[Page 58829]]
Required Action B.2.2. Additionally, the amendments revised a test
frequency note in Surveillance Requirement (SR) 3.3.6.2 under TS 3.3.6,
``Engineered Safety Features Actuation System (ESFAS) Logic and Manual
Trip,'' which should have been included in the license amendment
request for Technical Specifications Task Force (TSTF) change traveler
TSTF-425, Revision 3, ``Relocate Surveillance Frequencies to Licensee
Control--RITSTF [Risk-Informed TSTF] Initiative 5b.''
Date of issuance: September 9, 2014.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Unit 1--194; Unit 2--194; Unit 3--194. A publicly-
available version is in ADAMS under Accession No. ML14202A378;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Renewed Facility Operating License Nos. NPF-41, NPF-51, and NPF-74:
The amendment revised the Operating Licenses and Technical
Specifications.
Date of initial notice in Federal Register: February 4, 2014 (79 FR
6640). The supplemental letter dated December 12, 2013, 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 9, 2014.
No significant hazards consideration comments received: No.
Dairyland Power Cooperative, Docket Nos.: 50-409 and 72-046, La Crosse
Boiling Water Reactor (LACBWR), La Crosse County, Wisconsin
Date of application for amendment: August 6, 2013, supplemented by
letters dated January 16, 2014, and April 14, 2014.
Brief description of amendment: The amendment approves changes to
the Emergency Plan, including removal of the various emergency actions
related to the former spent fuel pool, the transfer of responsibility
for implementing the Emergency Plan to the Security Shift Supervisors
at the ISFSI, a revised emergency plan organization, removal of the
fire brigade, and abandonment of the LACBWR Control Room consistent
with the current state of decommissioning, in that all of the spent
nuclear fuel has now been transferred from the spent fuel pool to an
independent spent fuel storage installation.
Date of issuance: September 8, 2014.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 73.
Possession Only License No. DPR-45.
Date of initial notice in Federal Register: October 29, 2013 (78 FR
64543). The supplements dated January 16, 2014, and April 14, 2014,
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 on
October 29, 2013 (78 FR 64543).
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated September 8, 2014.
No significant hazards consideration comments received: No.
Duke Energy Carolinas, LLC, Docket Nos.: 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: October 28, 2013, as
supplemented by letter dated June 3, 2014.
Brief description of amendments: The amendments modify Technical
Specification (TS) 3.8.4. Specifically, the change allows a one-time
extension of the completion time for Required Action A.2.2 to support
replacement of the existing shared 125 VDC vital batteries.
Date of issuance: September 10, 2014.
Effective date: This license amendment is effective as of its date
of issuance and shall be implemented within 30 days of issuance.
Amendment Nos.: 274 and 254. A publicly-available version is in
ADAMS under Accession No. ML14231A634; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the licenses and technical specifications.
Date of initial notice in Federal Register: January 21, 2014 (79 FR
3415). The supplemental letter dated June 3, 2014, 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 10, 2014.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit 2, Westchester County, New York
Date of amendment request: January 16, 2014, as supplemented by
letters dated April 2, and April 15, 2014.
Brief description of amendment: The amendment revises Indian Point
Nuclear Generating Unit 2 Technical Specification (TS) 5.5.7, ``Steam
Generator (SG) Program,'' to exclude portions of the SG tubes below the
top of the SG tubesheet from periodic inspections and plugging by
implementing the alternate repair criteria ``H*.'' In addition, TS
3.4.13, ``RCS [reactor coolant system] Operational Leakage,'' is being
revised to reduce the allowable primary to secondary leakage through
any one SG from 150 to 85 gallons per day and TS 5.6.7, ``Steam
Generator Tube Inspection Program,'' is being revised to include
additional reporting requirements.
Date of issuance: September 5, 2014.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 277. A publicly-available version is in ADAMS under
Accession No. ML14198A161; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. DPR-26: The amendment revised the
Facility Operating License and the TSs.
Date of initial notice in Federal Register: March 18, 2014 (79 FR
15147). The supplemental letters provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 5, 2014.
No significant hazards consideration comments received: No.
[[Page 58830]]
Florida Power and Light Company, et al., Docket Nos.: 50-335 and 50-
389, St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida
Date of application for amendment: May 21, 2013, as supplemented by
letter dated October 4, 2013.
Brief description of amendment: The amendments revised the
Technical Specifications (TSs) moderator temperature coefficient
surveillance requirements associated with the implementation of Topical
Report WCAP-16011-P-A, ``Startup Test Activity Reduction (STAR)
Program,'' which describes the methods to be used for the
implementation of reduction in the startup testing requirements. The
changes are consistent with the NRC-approved Technical Specification
Task Force (TSTF) Standard Technical Specifications change TSTF-486,
Revision 2 as included in NUREG-1432, Revision 4.0, Standard Technical
Specifications--Combustion Engineering Plants.
The NRC staff published a notice of opportunity for comment in the
Federal Register on July 27, 2007 (72 FR 41360), on possible amendments
adopting TSTF-486 using the NRC's consolidated line-item improvement
process for amending licensees' TSs, which included a model safety
evaluation (SE) and model no significant hazards consideration (NSHC)
determination. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on September 6, 2007 (72 FR
51259), which included the resolution of public comments on the model
SE and model NSHC determination. The licensee affirmed in its
application dated May 21, 2013, that the proposed changes to the TSs
satisfy the intent of TSTF-486.
Date of issuance: September 16, 2014.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 219 and 168 (ADAMS Accession No. ML14218A180).
Documents related to these amendments are provided in an SE enclosed
with the amendments.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the License and TSs.
Date of initial notice in Federal Register: July 23, 2013 (78 FR
44173). The supplement dated October 4, 2013, 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 16, 2014.
No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos.: 50-445 and 50-446,
Comanche Peak Nuclear Power Plant (CPNPP), Units 1 and 2, Somervell
County, Texas
Date of amendment request: November 21, 2013, as supplemented by
letters dated February 4 and April 1, 2014.
Brief description of amendment: The amendments revised the date of
cyber security plan (CSP) full implementation schedule (Milestone 8)
and the existing license condition 2.H in the facility operating
licenses NPF-87 and NPF-89 for CPNPP, Units 1 and 2, respectively. The
CSP and the implementation schedule for CPNPP, Units 1 and 2, were
previously approved by the NRC staff by letter dated July 26, 2011
(ADAMS Accession No. ML111780745).
Date of issuance: September 8, 2014.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: Unit 1--163; Unit 2--163. A publicly-available
version is in ADAMS under Accession No. ML14183A342; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: April 8, 2014 (79 FR
19399). The supplements dated February 4 and April 1, 2014, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 8, 2014.
No significant hazards consideration comments received: No.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
1, Rockingham County, New Hampshire
Date of amendment request: January 30, 2012, as supplemented by
letters dated May 10, 2012, September 20, 2012, March 27, 2013,
December 20, 2013, and January 29, 2014.
Description of amendment request: The original application proposed
revisions to the technical specifications (TSs) for new and spent fuel
storage as a result of the new criticality analyses for the new fuel
vault (NFV) and spent fuel pool (SFP). By letter dated December 20,
2013 (ADAMS Accession No. ML13360A045), NextEra requested that the SFP
and NFV be separated into two separate license amendment requests. This
amendment revised the TSs related to spent fuel storage as a result of
new criticality analyses for the SFP. The license amendment request for
the NFV will be processed under TAC No. MF3283.
Date of issuance: September 3, 2014.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 142. A publicly-available version is in ADAMS under
Accession No. ML14184A795; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-86: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 14, 2012 (77 FR
48559). The supplemental letters dated May 10, 2012, September 20,
2012, March 27, 2013, December 20, 2013, and January 29, 2014, 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.
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 3, 2014.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: April 4, 2014, as supplemented by the
letter dated May 27, 2014.
Brief description of amendment: The amendment revises Tier 2*
information, incorporated into the VEGP Units 3 and 4 Updated Final
Safety Analysis Report (UFSAR). Specifically, the amendment revises the
details regarding the structural floor of the Auxiliary Building and
its constructability. Notes are added to drawings in subsection 3H.5 of
the UFSAR in order to clarify variations in detail design such as size
[[Page 58831]]
and spacing or reinforcement and spans of the noncritical sections of
floors.
Date of issuance: July 3, 2014.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 21. A publicly-available version is in ADAMS under
Accession No. ML14150A133; documents related to these amendments are
listed in the Safety Evaluation enclosed with the amendments.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised
the Facility Combined Licenses.
Date of initial notice in Federal Register: April 29, 2014 (79 FR
24025).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 3, 2014.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 22nd day of September 2014.
For the Nuclear Regulatory Commission.
George A. Wilson,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2014-23015 Filed 9-29-14; 8:45 am]
BILLING CODE 7590-01-P