[Federal Register Volume 79, Number 140 (Tuesday, July 22, 2014)]
[Notices]
[Pages 42539-42557]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-17257]
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NUCLEAR REGULATORY COMMISSION
[NRC-2014-0169]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission
(NRC) is publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued, and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from June 26, 2014, to July 9, 2014.
DATES: Comments must be filed by August 21, 2014. A request for a
hearing must be filed by September 22, 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-2014-0169. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected].
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-2014-0169 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-2014-0169.
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
[[Page 42540]]
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-2014-0169 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 in comment submissions that you do not want to be publicly
disclosed in your comment submission. The NRC will post all comment
submissions at http://www.regulations.gov as well as enter the comment
submissions into ADAMS, and the NRC does not routinely edit comment
submissions to remove identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of Title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period 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.
[[Page 42541]]
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.
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
ten 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://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 NRC 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
[[Page 42542]]
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.
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: April 23, 2014, as supplemented by
letter dated June 19, 2014. Publicly available versions are in ADAMS
under Accession Nos. ML14113A445 and ML14170B201, respectively.
Description of amendment request: The proposed amendment would
revise the technical specification (TS) surveillance requirements (SRs)
associated with TS 3.8.4, ``DC Sources--Operating'' and TS 3.8.6,
``Battery Cell Parameters.''
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.
Performing the proposed changes in battery parameter surveillance
testing and verification is not a precursor of any accident previously
evaluated. Furthermore, these changes will help to ensure that the
voltage and capacity of the batteries is such that they will provide
the power assumed in calculations of design basis accident mitigation.
Therefore, DTE concludes that the proposed changes do not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve any modification of the plant
or how the plant is operated; they only involve surveillance testing
and verification activities.
Therefore, DTE concludes that these proposed changes do 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.
Margin of safety is related to the confidence in the ability of the
fission product barriers to perform their design functions during and
following an accident situation. These barriers include the fuel
cladding, the reactor coolant system, and the containment system. The
performance of the fuel cladding, reactor coolant, and containment
systems will not be impacted by the proposed changes.
The proposed Fermi 2 revisions of the SRs ensure the continued
availability and operability of the batteries. As such, sufficient
[direct current] capacity to support operation of mitigation equipment
remains within the design basis.
Therefore, DTE concludes that 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: Bruce R. Maters, DTE Energy, General
Counsel--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-1279.
NRC Branch Chief: Robert D. Carlson.
Duke Energy Progress Inc., Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: February 10, 2014, as supplemented by
letter dated April 4, 2014. Publicly-available versions are in ADAMS
under Accession Nos. ML14052A065 and ML14107A339, respectively.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.3.1 for the Reactor Protection System
Instrumentation Turbine Trip function on Low Auto Stop Oil (ASO)
Pressure to a Turbine Trip function on Low Electro-Hydraulic (EH) Fluid
Oil Pressure. The amendment would revise the Allowable Value and
Nominal Trip Setpoint and revise the TS by applying additional testing
requirements listed in Technical Specifications Task Force Traveler
493-A Revision 4, ``Clarify Application of Setpoint Methodology for
Limiting Safety System Setting Functions,'' for Low EH Fluid Oil
Pressure trip only.
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 reflects a design change to the turbine control
system that results in the use of an increased control oil pressure
system, necessitating a change to the value at which a low EH fluid oil
pressure initiates a reactor trip on turbine trip. The EH oil pressure
is an input to the reactor trip instrumentation in response to a
turbine trip event. The value at which the low Electro-Hydraulic fluid
oil initiates a reactor trip is not an accident initiator. A change in
the nominal control oil pressure does not introduce any mechanisms that
would increase the probability of an accident previously analyzed. The
reactor trip on turbine trip function is initiated by the same
protective signal as used for the ASO System trip signal. There is no
change in form or function of this signal and the probability or
consequences of previously analyzed accidents are not impacted.
The proposed change also adds test requirements to a TS instrument
function related to those variables that have a significant safety
function to ensure that instruments will function as required to
initiate protective systems or actuate mitigating systems at the point
assumed in the applicable setpoint calculation. Surveillance tests are
not an initiator to any accident previously evaluated. As a result, the
probability of any accident previously evaluated is not significantly
increased. The systems and components required by the TSs for which
surveillance tests are added are still required to be operable, meet
the acceptance criteria for the surveillance requirements, and be
capable of performing any mitigation function.
[[Page 42543]]
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 EH fluid oil pressure decreases in response to a turbine trip.
The value at which the low EH fluid oil initiates a reactor trip is not
an accident initiator. The proposed TS change reflects the higher
pressure that will be sensed after the pressure switches are relocated
from the ASO System to the AST [Auto Stop Trip] high pressure header.
Failure of the new switches would not result in a different outcome
than is considered in the current design basis. Further, the change
does not alter assumptions made in the safety analysis but ensures that
the instruments perform as assumed in the accident analysis.
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 change involves a parameter that initiates an anticipatory
reactor trip following a turbine trip. The safety analyses do not
credit this anticipatory trip for reactor core protection. The original
pressure switch configuration and the new pressure switch configuration
both generate the same reactor trip signal. The difference is that the
initiation of the trip will now be adjusted to a different system of
higher pressure. This system function of sensing and transmitting a
reactor trip signal on turbine trip remains the same. Also, the
proposed change adds test requirements that will assure that (1)
technical specifications instrumentation Allowable Values will be
limiting settings for assessing instrument channel operability and (2)
will be conservatively determined so that evaluation of instrument
performance history and the as left tolerance requirements of the
calibration procedures will not have an adverse effect on equipment
operability. The testing methods and acceptance criteria for systems,
structures, and components, specified in applicable codes and standards
(or alternatives approved for use by the NRC) will continue to be met
as described in the plant licensing basis including the updated Final
Safety Analysis Report. There is no impact to safety analysis
acceptance criteria as described in the plant licensing basis because
no change is made to the accident analysis assumptions.
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: Lara S. Nichols, Deputy General Counsel,
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A,
Charlotte, NC 28202.
NRC Acting Branch Chief: Lisa M. Regner.
Duke Energy Progress Inc., Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit No. 1, New Hill, North Carolina
Date of amendment request: April 24, 2014. A publicly-available
version is in ADAMS under Accession No. ML14114A743.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.4.5, ``Steam Generator Tube
Integrity,'' TS 6.8.4.I, ``Steam Generator Program,'' and TS 6.9.1.7,
``Steam Generator Tube Inspection Report'' to address implementation
associated with the inspections and reporting requirements as described
in Technical Specifications Task Force (TSTF) TSTF-510-A, Revision 2,
``Revision to Steam Generator Program Inspection Frequencies and Tube
Sample Selection.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Steam Generator (SG) Program to
modify the frequency of verification of SG tube integrity and SG tube
sample selection. A steam generator tube rupture (SGTR) event is one of
the design basis accidents that are analyzed as part of a plant's
licensing basis. The proposed SG tube inspection frequency and sample
selection criteria will continue to ensure that the SG tubes are
inspected such that the probability of a SGTR is not increased. The
consequences of a SGTR are bounded by the conservative assumptions in
the design basis accident analysis. The proposed change will not cause
the consequences of a SGTR to exceed those assumptions.
Therefore, it is concluded that this change does not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the Steam Generator Program will not
introduce any adverse changes to the plant design basis or postulated
accidents resulting from potential tube degradation. The proposed
change does not affect the design of the SGs or their method of
operation. In addition, the proposed change does not impact any other
plant system or component.
Therefore, it is concluded that this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part of
the reactor coolant pressure boundary and, as such, are relied upon to
maintain the primary system's pressure and inventory. As part of the
reactor coolant pressure boundary, the SG tubes are unique in that they
are also relied upon as a heat transfer surface between the primary and
secondary systems such that residual heat can be removed from the
primary system. In addition, the SG tubes also isolate the radioactive
fission products in the primary coolant from the secondary system. In
summary, the safety function of a SG is maintained by ensuring the
integrity of its tubes. Steam generator tube integrity is a function of
the design, environment, and the physical condition of the tube. The
proposed change does not affect tube design or operating environment.
The proposed change will continue to require monitoring of the physical
condition of the SG tubes such that there will not be a reduction in
the margin of safety compared to the current requirements.
Therefore, it is concluded that the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 42544]]
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lara S. Nichols, Deputy General Counsel,
Duke Energy Corporation, 550 South Tyron Street, Mail Code DEC45A,
Charlotte, NC 28202.
NRC Acting Branch Chief: Lisa M. Regner.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: March 18, 2014. A publicly-available
version is in ADAMS under Accession No. ML14086A389.
Description of amendment request: The amendment would adopt
Technical Specification (TS) Task Force (TSTF) change traveler TSTF-
535, Revision 0, ``Revise Shutdown Margin [SDM] Definition to Address
Advanced Fuel Designs,'' at Columbia Generating Station. The notice of
availability of TSTF-535, Revision 0, was announced in the Federal
Register on February 26, 2013 (78 FR 13100).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the definition of SDM. SDM is not an
initiator to any accident previously evaluated. Accordingly, the
proposed change to the definition of SDM has no effect on the
probability of any accident previously evaluated. SDM is an assumption
in the analysis of some previously evaluated accidents and inadequate
SDM could lead to an increase in consequences for those accidents.
However, the proposed change revises the SDM definition to ensure that
the correct SDM is determined for all fuel types at all times during
the fuel cycle. As a result, the proposed change does not adversely
affect the consequences of any accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the definition of SDM. The change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operations. The change does not alter
assumptions made in the safety analysis regarding SDM.
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 definition of SDM. 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 ensures that the SDM assumed in
determining safety limits, limiting safety system settings or limiting
conditions for operation is correct for all BWR fuel types at all times
during the fuel cycle.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: March 24, 2014, as supplemented by
letter dated May 8, 2014. Publicly-available versions are in ADAMS
under Accession Nos. ML14098A400 and ML14141A538, respectively.
Description of amendment request: The amendment would revise
Columbia Generating Station Technical Specification (TS) Table 3.3.1.1-
1 to update Scram Discharge Volume (SDV) instrumentation nomenclature,
add a Surveillance Requirement (SR) which was previously omitted, and
add footnotes to an SR consistent with TS Task Force (TSTF) change
traveler TSTF-493, Revision 4, ``Clarify Application of Setpoint
Methodology for LSSS [Limiting Safety System Settings] Functions,''
Option A. The notice of availability of the models for plant-specific
adoption of TSTF-493, Revision 4, was announced in the Federal Register
on May 11, 2010 (75 FR 26294).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to Function 7 names are administrative in
nature and ensure that the description of SDV Water Level--High
instrumentation in TS matches the plant configuration. The addition of
a missing channel check SR and TSTF-493 footnotes for the new Function
7.b instruments makes the TS more comprehensive by ensuring the
appropriate surveillances and footnotes are applied to this
instrumentation.
The replacement instruments for Function 7.b meet the high
functional reliability standard of GDC 21 [General Design Criteria 21,
``Protection system reliability and testability,'' of 10 CFR Part 50,
Appendix A] and all pertinent requirements of 10 CFR 50.55a(h)(2). The
instrumentation modification was reviewed under 10 CFR 50.59(c)(1) and
determined to not meet any of the criteria in 10 CFR 50.59(c)(2).
The addition of a channel check to Function 7.a and addition of
TSTF-493 notes (d) and (e) to SR 3.3.1.1.10 for the Function 7.b
instrumentation do not change accident frequency or consequences. TS
requirements that govern operability or routine testing of plant
instruments are not assumed to be initiators of any analyzed event
because these instruments are intended to prevent, detect, or mitigate
accidents. Additionally, these proposed changes will not increase the
consequences of an accident previously evaluated because the proposed
changes do not adversely impact structures, systems, or components. The
proposed TS changes establish requirements that ensure components are
operable when necessary for the prevention or mitigation of accidents
or transients. Furthermore, there will be no change in the types or
significant increase in the amounts of any effluents released offsite.
[[Page 42545]]
In summary, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to administratively revise instrument
descriptions, incorporate a new SR, and add footnotes to an existing SR
do not change the parameters within which Columbia is operated.
The proposed changes do not adversely impact the manner in which
the SDV Water Level--High RPS [Reactor Protection System]
instrumentation will operate under normal and abnormal operating
conditions. The instrumentation design changes were reviewed under 10
CFR 50.59(c)(1) and determined to not meet any of the criteria of 10
CFR 50.59(c)(2). The proposed changes will not alter the functional
demands on credited equipment. No alteration in the procedures which
ensure that Columbia remains within analyzed limits are proposed and no
change is being made to procedures relied upon to respond to an off-
normal event.
Therefore, these proposed changes provide an equivalent level of
safety and will 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 to the function descriptions in TS Table
3.3.1.1-1 Functions 7.a and 7.b are considered administrative in
nature, and do not impact plant safety.
Margins of safety are established in the design of components, the
configuration of components to meet certain performance parameters, and
in the establishment of setpoints to initiate alarms and actions. The
proposed changes support a planned upgrade of the SDV instrumentation
that preserves the reliability of the RPS system. The proposed changes
do not adversely affect the probability of failure or availability of
the affected instrumentation. The instrumentation design changes were
evaluated under 10 CFR 50.59(c)(1) and determined not to meet any of
the criteria of 10 CFR 50.59(c)(2).
The addition of a Channel Check SR to TS Table 3.3.1.1-1 Function
7.a and the addition of TSTF-493 notes (d) and (e) to SR 3.3.1.1.10 for
the new scram discharge instrumentation in TS Table 3.3.1.1-1 Function
7.b are conservative changes that align the SRs for proper
determination of operability with that of similar instrumentation.
On this basis, is concluded that the proposed changes do not result
in a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: November 26, 2013. A publicly-available
version is in ADAMS under Accession No. ML13346A026.
Description of amendment request: The amendment would revise
Technical Specification 4.3.4, ``Heavy Loads'' limitation imposed on
maximum weight that could travel over the irradiated fuel in the spent
fuel pool.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (CFR) Section 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 Reactor Building crane is being upgraded to meet the applicable
single-failure-proof criteria of NUREG 0554 and NUREG 0612 for the
modification of the existing non single-failure-proof crane. While
loads in excess of 2,000 lbs [pounds] shall continue to be prohibited
from travel over irradiated fuel assemblies in the spent fuel pool by
the PNPS [Pilgrim Nuclear Power Station] Technical Specifications, a
Multi-Purpose Canister (MPC) lid will be permitted to travel over
irradiated fuel assemblies in a transfer cask, using a single-failure-
proof handling system as described in NUREG-0800 Section 9.1.5
Paragraph llI.4.C, to enable the conduct of dry cask storage loading
and unloading operations. Specifically, this will enable the MPC lid
and its associated lifting apparatus to travel over irradiated fuel
assemblies in a MPC. The probability of dropping this load onto an
irradiated fuel assembly in the canister is reduced as a result of the
reliability of the single-failure-proof handling system.
The proposed change does not affect the consequences of any
accidents previously evaluated in the PNPS UFSAR [Updated Final Safety
Analysis Report]. The change involves the travel of heavy loads over
irradiated fuel assemblies in a transfer cask using a single-failure-
proof handling system. Under these circumstances, no new load drop
accidents are postulated and no changes to the probabilities or
consequences of accidents previously evaluated are involved.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Section 10.3 of the PNPS UFSAR evaluates fuel storage and handling
operations. Section 14 of the PNPS UFSAR discusses the analysis of
design basis fuel handling accidents involving drop of an irradiated
assembly resulting in multiple fuel rod failures and consequent release
of radioactivity. The change involves the travel of heavy loads over
irradiated fuel assemblies in a transfer cask using a single-failure-
proof handling system. Under these circumstances, no new or different
load drop accidents are postulated to occur and there are no changes in
any of the load drop accidents previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The revised Technical Specification changes do not involve a
reduction in any margin of safety. Technical Specification 4.3.4
currently prohibits travel of heavy loads in excess of 2,000 lbs over
irradiated fuel assemblies in the spent fuel pool. The proposed change
will continue to restrict travel of heavy loads in excess of 2,000 lbs
over irradiated fuel assemblies in the spent fuel pool, with the
exception of the MPC lid over irradiated fuel assemblies in the
canister to enable dry cask storage operations. This exception is only
permitted when the heavy load is handled using a single-failure-proof
handling system. Due to the reliability of this upgraded handling
system that complies with the guidance of NUREG-0800 Section 9.1.5 for
a single-failure-proof handling system, a load drop accident is not
considered a credible event. Under these circumstances, no
[[Page 42546]]
new load drop accidents are postulated and no reductions in margins of
safety are involved.
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: Ms. Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Benjamin G. Beasley
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: March 24, 2014. A publicly available
version is in ADAMS under Accession No. ML14085A257.
Description of amendment request: The proposed amendment would
revise the site emergency plan for the permanently defueled condition
to reflect changes in the on-shift staffing and Emergency Response
Organization staffing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the SEP [Site Emergency Plan] do not impact
the function of plant structures, systems, or components (SSCs). The
proposed changes do not affect accident initiators or precursors, nor
does it alter design assumptions. The proposed changes do not prevent
the ability of the on-shift staff and ERO [Emergency Response
Organization] to perform their intended functions to mitigate the
consequences of any accident or event that will be credible in the
permanently defueled condition. The proposed changes only remove
positions that will no longer be credited in the SEP in the permanently
defueled condition.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes reduce the number of on-shift and ERO
positions commensurate with the hazards associated with a permanently
shutdown and defueled facility. The proposed changes do not involve
installation of new equipment or modification of existing equipment, so
that no new equipment failure modes are introduced. Also, the proposed
changes do not result in a change to the way that the equipment or
facility is operated so that no new accident initiators are created.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the level
of radiation dose to the public. The proposed changes are associated
with the SEP staffing and do not impact operation of the plant or its
response to transients or accidents. The change does not affect the
Technical Specifications. The proposed changes do not involve a change
in the method of plant operation, and no accident analyses will be
affected by the proposed changes. Safety analysis acceptance criteria
are not affected by the proposed changes. The revised SEP will continue
to provide the necessary response staff with the proposed changes.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Jeanne Cho, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Douglas A. Broaddus.
Exelon Generation Company (EGC), LLC, Docket Nos. STN 50-456 and STN
50-457, Braidwood Station, Units 1 and 2, Will County, Illinois, Docket
Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle
County, Illinois
Date of amendment request: April 17, 2014. A publicly-available
version is in ADAMS under Accession No. ML14111A257.
Description of amendment request: The proposed amendment would
revise required action notes in the Braidwood and Byron TS 3.3.1 and TS
3.3.2 to reflect the specific functions in TS 3.3.1 and TS 3.3.2 that
have bypass test capability installed and the specific functions that
do not have bypass test capability installed. The current wording is no
longer applicable because the installation and implementation of the
bypass test instrumentation modifications for certain functions have
been completed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change is administrative in nature as it revises
previously approved specific TS [Technical Specifications] Required
Actions Notes that are no longer applicable following plant
modification installation and implementation to reflect the applicable
RTS [Reactor Trip System] and ESFAS [Engineered Safety Feature
Actuation System] Functions with installed bypass test capability.
The proposed change does not impact any accident initiators,
analyzed events, or assumed mitigation of accident or transient events
modeled in the safety analyses. The proposed change does not alter the
design assumptions, conditions, or configuration of the facility, nor
does it affect the structural and functional integrity of the RTS and
ESFAS. The proposed change does not alter or prevent the ability of any
structures, systems, and components from performing their intended
design function to mitigate the consequences of an initiating event
within the applicable acceptance criteria.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
[[Page 42547]]
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to revise previously approved specific TS
Required Actions Notes that are no longer applicable to specific RTS
and ESFAS Functions with installed bypass test capability is
administrative in nature. The proposed change does not result in a
change to any design function or the manner in which the RTS and ESFAS
operates to provide plant protection. The RTS and ESFAS will continue
to have the same setpoints after the proposed change is implemented. In
addition, this change does not install or modify any plant equipment.
Therefore, no new failure modes are being created nor does the change
result in the creation of any changes to the existing accident
scenarios or do they create any new or different accident scenarios.
The types of accidents defined in the UFSAR [updated final safety
analysis report] continue to represent the credible spectrum of events
to be analyzed which determine safe plant operation.
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.
No safety analyses are changed or modified as a result of the
proposed change to revise previously approved specific TS Required
Actions Notes that are no longer applicable to RTS and ESFAS Functions
with installed bypass test capability. The proposed change does not
alter the manner in which the safety limits, limiting safety system
settings, or limiting conditions for operation are determined. Margins
associated with the current applicable safety analyses acceptance
criteria are unaffected. The current safety analyses remain bounding
since their conclusions are not affected by this change and the plant
will continue to operate in a manner consistent with the safety
analyses. The safety systems credited in the safety analyses will
continue to be available to perform their mitigation functions.
Therefore, the proposed change does not result in a significant
reduction in the margin of safety.
Based on the above evaluation, EGC concludes that the proposed
amendments do not involve a significant hazards consideration under the
standards set forth in 10 CFR 50.92, paragraph (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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
Date of amendment request: October 31, 2013. A publicly-available
version is in the ADAMS System under Accession No. ML13308A387.
Description of amendments request: The amendments would modify the
Technical Specification requirements regarding steam generator tube
inspections and reporting as described in Technical Specification Task
Force 510-A, Revision 2, ``Revision to Steam Generator Program
Inspection Frequencies and Tube Sample Selection.''
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
No.
The proposed change revises the Steam Generator (SG) Program to
modify the frequency of verification of SG tube integrity and SG tube
sample selection. A steam generator tube rupture event (SGTR) is one of
the design basis accidents that are analyzed as part of a plant's
licensing basis. The proposed SG tube inspection frequency and sample
selection criteria will continue to ensure that the SG tubes are
inspected such that the probability of a SGTR is not increased. The
consequences of a SGTR are bounded by the conservative assumptions in
the design basis accident analysis. The proposed change will not cause
the consequences of a SGTR to exceed these assumptions.
Therefore, operation of the facility in accordance with the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different type of accident
from any accident previously evaluated; or
No.
The proposed changes to the SG Program will not introduce any
adverse changes to the plant design basis or postulated accidents
resulting from potential tube degradation. The proposed change does not
affect the design of the SGs or their method of operation. In addition,
the proposed change does not impact any other plant system or
component.
Therefore, the proposed amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
No.
The SG tubes in pressurized water reactors are an integral part of
the reactor coolant pressure boundary and, as such, are relied upon to
maintain the primary system's pressure and inventory. As part of the
reactor coolant pressure boundary, the SG tubes are unique in that they
are also relied upon as a heat transfer surface between the primary and
secondary systems such that residual heat can be removed from the
primary system. In addition, the SG tubes also isolate the radioactive
fission products in the primary coolant from the secondary system. In
summary, the safety function of a SG is maintained by ensuring the
integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change will continue to require monitoring of the physical
condition of the SG tubes such that there will not be a reduction in
the margin of safety compared to the current requirements.
Therefore, the proposed amendment would 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
amendments request involves no significant hazards consideration.
Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200
Exelon Way, Kennett Square, PA 19348.
NRC Branch Chief: Benjamin G. Beasley
[[Page 42548]]
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
Date of amendment request: November 13, 2013. A publicly-available
version is in the ADAMS System under Accession No. ML13318A892.
Description of amendments request: The amendments would modify the
Technical Specification requirements to adopt the changes described in
Technical Specification Task Force 426-A, Revision 5, ``Revise or Add
Actions to Preclude Entry into LCO 3.0.3--RITSTF Initiatives 6b and
6c.''
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
No.
The proposed change provides a short Completion Time to restore an
inoperable system for conditions under which the existing Technical
Specifications require a plant shutdown to begin within one hour in
accordance with Limiting Condition for Operation 3.0.3. Entering into
Technical Specification Actions is not an initiator of any accident
previously evaluated. As a result, the probability of an accident
previously evaluated is not significantly increased. The consequences
of any accident previously evaluated that may occur during the proposed
Completion Times are no different from the consequences of the same
accident during the existing one hour allowance. As a result, the
consequences of any accident previously evaluated are not significantly
increased.
Therefore, operation of the facility in accordance with the
proposed amendment does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different type of accident
from any accident previously evaluated; or
No.
No new or different accidents result from utilizing the proposed
change. The changes do 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 changes do not impose any new or different requirements. The
changes do not alter assumptions made in the safety analysis.
Therefore, the proposed amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
No.
The proposed change increases the time the plant may operate
without the ability to perform an assumed safety function. The analyses
in WCAP-16125-NP-A, ``Justification for Risk-Informed Modifications to
Selected Technical Specifications for Conditions Leading to Exigent
Plant Shutdown,'' Revision 2, August 2010, demonstrated that there is
an acceptably small increase in risk due to a limited period of
continued operation in these conditions and that this risk is balanced
by avoiding the risks associated with a plant shutdown. As a result,
the change to the margin of safety provided by requiring a plant
shutdown within one hour is not significant.
Therefore, the proposed amendment would 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
amendments request involves no significant hazards consideration.
Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200
Exelon Way, Kennett Square, PA 19348
NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
Date of amendment request: January 13, 2014. A publicly-available
version is in the ADAMS under Accession No. ML14015A138.
Description of amendments request: The amendments would add a
Technical Specification (TS) for the atmospheric dump valves (ADVs).
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 addition of a new TS to address the operability of the
ADVs does not alter the assumed initiators to any analyzed event. The
probability of an accident previously evaluated will not be increased
by this proposed change. This proposed change will not affect
radiological dose consequence analyses. The radiological dose
consequence analyses assume a certain release of radioactive material
through the ADVs following a steam generator tube rupture (SGTR), which
is not affected by the addition of the ADVs to the TS. The addition of
a Surveillance Requirement for the ADVs will continue to ensure that
the ADVs can perform their specified function. The consequences of an
accident previously evaluated will not be increased by this proposed
change.
Therefore, operation of the facility in accordance with the
proposed TS for the ADVs will 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 addition of a new TS to address the operability of the
ADVs has been evaluated to determine the effect of adding the new TS to
the operation of the plant. This change does not involve any alteration
in the plant configuration (no new or different type of equipment will
be installed) or make changes in the methods governing normal plant
operation. The change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Therefore, operation of the facility in accordance with the
proposed addition of a new TS to address the operability of the ADVs
would 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 margin of safety is related to the ability of the ADV to
release enough steam to cool the Reactor Coolant System down and be
isolated when required to limit the radioactive release from a SGTR.
The inclusion of the ADVs in the TS will provide limited time for
continued operation without both ADVs available. This ensures that the
margin of safety is maintained by ensuring that the ADV can meet the
assumptions for its operation specified in the SGTR analysis. Since the
radiological consequences of a SGTR are not affected
[[Page 42549]]
by the addition of the proposed TS, the margin of safety is not changed
significantly.
Therefore, the proposed addition of a new TS to address the
operability of the ADVs 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
amendments request involves no significant hazards consideration.
Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200
Exelon Way, Kennett Square, PA 19348.
NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
Date of amendment request: February 13, 2014. A publicly-available
version is in the ADAMS under Accession No. ML14050A374.
Description of amendments request: The amendments would modify the
as-found lift tolerances in the surveillance requirement for the
pressurizer safety valves (PSVs).
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?
No.
The proposed change, modifying the as-found and as-left lift
settings in the Surveillance Requirement of the PSVs, does not change
the design function or operation of the PSVs and it does not change the
way the PSVs are maintained, tested, or inspected. The PSVs are not
accident initiators; they operate in response to the pressurization of
the Reactor Coolant System (RCS). They limit the pressure of the RCS to
less than the allowable American Society of Mechanical Engineers Boiler
and Pressure Vessel, Section III Code during an accident or transient.
Analyses were performed of peak pressure events, which are evaluated
against the RCS limit. Action of the PSVs is required to mitigate the
consequences of these events. The change in the setpoint tolerance and
a change in one valve's nominal setpoint were explicitly considered in
the analysis of these events. The RCS pressure remained below the
required limits with these changes considered. Therefore, this change
does not impact the ability of the PSVs to perform their safety
function during evaluated accidents.
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?
No.
The proposed change, modifying the as-found and as-left lift
settings in the Surveillance Requirement of the PSVs, does not change
the PSVs design function to maintain RCS pressure below the RCS
pressure Safety Limit of 2750 psia [pounds per square inch absolute]
during design basis accidents nor does it affect the PSVs ability to
perform this design function. The proposed change does not require any
modification to the plant (other than the setpoint change) or change
equipment operation or testing. It also does not create any credible
new failure mechanisms, malfunctions, or accident initiators that would
cause an accident not previously considered.
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?
No.
The proposed change, modifying the as-found and as-left lift
settings in the Surveillance Requirement of the PSVs, does not involve
a significant reduction in the margin of safety in maintaining RCS
pressure below Safety Limits of 2750 psia during design basis
accidents. The analyses conducted in support of this proposed change
evaluated the ability of the PSVs to maintain an adequate safety margin
assuming the change in setpoint tolerances and a change in one valve's
nominal setpoint. The analysis determined that the response of the PSVs
would maintain an adequate safety margin to the reactor coolant Safety
Limit of 2750 psia.
Therefore the proposed change does not involve a significant
reduction in the margin of safety of maintaining RCS pressure the below
RCS pressure Safety Limit.
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
amendments request involves no significant hazards consideration.
Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200
Exelon Way, Kennett Square, PA 19348.
NRC Branch Chief: Benjamin G. Beasley.
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland
Date of amendment request: May 1, 2014. A publicly-available
version is in the ADAMS under Accession No. ML14125A015.
Description of amendments request: The amendments would modify the
Technical Specifications (TSs) by relocating specific surveillance
frequencies to a licensee-controlled program with the implementation of
Nuclear Energy Institute 04-10, ``Risk Informed Method for Control of
Surveillance Frequencies.''
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
Response: No.
The proposed change relocates the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program. Surveillance frequencies are
not an initiator to any accident previously evaluated. As a result, the
probability of any accident previously evaluated is not significantly
increased. The systems and components required by the Technical
Specifications for which the surveillance frequencies are relocated are
still required to be operable, meet the acceptance criteria for the
surveillance requirements, and be capable of performing any mitigation
function assumed in the accident analysis. As a result, 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. Create the possibility of a new or different type of accident
from any accident previously evaluated; or
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical
[[Page 42550]]
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 changes do not impose any new or different
requirements. The changes do not alter assumptions made in the safety
analysis. The proposed changes are consistent with the safety analysis
assumptions and current plant operating practice.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Response: No.
The design, operation, testing methods, and acceptance criteria for
systems, structures and components specified in applicable codes and
standards (or alternatives approved for use by the NRC) will continue
to be met as described in the plant licensing basis (including the
updated final safety analysis report and the bases to the TS), since
these are not affected by changes to the surveillance frequencies.
Similarly, there is no impact to safety analysis acceptance criteria as
described in the plant licensing basis. To evaluate a change in the
relocated surveillance frequency, Calvert Cliffs will perform a
probabilistic risk evaluation using the guidance contained in NRC
approved NEI 04-10, Revision 1 in accordance with the TS Surveillance
Frequency Control Program. Nuclear Energy Institute 04-10, Revision 1
methodology provides reasonable acceptance guidelines and methods for
evaluating the risk increase of proposed changes to surveillance
frequencies consistent with Regulatory Guide 1.177.
Therefore, this 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's request involves no significant hazards consideration.
Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200
Exelon Way, Kennett Square, PA 19348.
NRC Branch Chief: Benjamin G. Beasley.
Florida Power and Light Company (FPL), et al., Docket Nos. 50-335 and
50-389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: February 20, 2014. Available in ADAMS
under Accession No. ML14070A087.
Description of amendment request: The amendment would revise the
Technical Specifications (TSs) by relocating specific surveillance
frequency requirements to a licensee-controlled program with
implementation of Nuclear Energy Institute (NEI) 04-10, ``Risk Informed
Technical Specification Initiative 5b, Risk Informed Method for Control
of Surveillance Frequencies'' (ADAMS Accession No. ML071360456). The
licensee stated that the NEI 04-10 methodology provides reasonable
acceptance guidelines and methods for evaluating the risk increase of
proposed changes to surveillance frequencies, consistent with
Regulatory Guide 1.177, ``An Approach for Plant-Specific Risk-Informed
Decision-Making: Technical Specifications'' (ADAMS Accession No.
ML003740176). The licensee stated that the changes are consistent with
NRC-approved Technical Specification Task Force (TSTF) Standard
Technical Specifications change TSTF-425, ``Relocate Surveillance
Frequencies to Licensee Control--RITSTF [Risk Informed Technical
Specifications Task Force] Initiative 5b,'' Revision 3 (ADAMS Accession
No. ML090850642). The Federal Register notice published on July 6, 2009
(74 FR 31996), announced the availability of TSTF-425, Revision 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration,
which is presented as follows:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change relocates the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program. Surveillance frequencies are
not an initiator to any accident previously evaluated. As a result, the
probability of any accident previously evaluated is not significantly
increased. The systems and components required by the Technical
Specifications for which the surveillance frequencies are relocated are
still required to be operable, meet the acceptance criteria for the
surveillance requirements, and be [sic] capable of performing any
mitigation function assumed in the accident analysis. As a result, 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 any accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do 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 changes do not impose any new or different requirements. The
changes do not alter assumptions made in the safety analysis
assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria for
systems, structures, and components (SSCs), specified in applicable
codes and standards (or alternatives approved for use by the NRC) will
continue to be met as described in the plant licensing basis (including
the final safety analysis report and bases to TS), since these are not
affected by changes to the surveillance frequencies. Similarly, there
is no impact to safety analysis acceptance criteria as described in the
plant licensing basis. To evaluate a change in the relocated
surveillance frequency, FPL will perform a probabilistic risk
evaluation using the guidance contained in NRC-approved NEI 04-10,
Revision 1 in accordance with the TS Surveillance Frequency Control
Program. NEI 04-10, Revision 1, methodology provides reasonable
acceptance guidelines and methods for evaluating the risk increase of
proposed changes to surveillance frequencies consistent with Regulatory
Guide (RG) 1.177.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700
[[Page 42551]]
Universe Blvd. MS LAW/JB, Juno Beach, Florida 33408-0420.
NRC Acting Branch Chief: Lisa M. Regner.
Florida Power and Light Company (FPL), Docket Nos. 50-250 and 50-251,
Turkey Point Nuclear Generating Unit Nos. 3 and 4, Miami-Dade County,
Florida
Date of amendment request: April 9, 2014. Available in ADAMS under
Accession No. ML14105A042.
Description of amendment request: The amendment would revise the
Technical Specifications (TSs) by relocating specific surveillance
frequency requirements to a licensee-controlled program with
implementation of Nuclear Energy Institute (NEI) 04-10, ``Risk Informed
Technical Specification Initiative 5b, Risk Informed Method for Control
of Surveillance Frequencies'' (ADAMS Accession No. ML071360456). The
licensee stated that the NEI 04-10 methodology provides reasonable
acceptance guidelines and methods for evaluating the risk increase of
proposed changes to surveillance frequencies, consistent with
Regulatory Guide 1.177, ``An Approach for Plant-Specific Risk-Informed
Decision-Making: Technical Specifications'' (ADAMS Accession No.
ML003740176). The licensee stated that the changes are consistent with
NRC-approved Technical Specification Task Force (TSTF) Standard
Technical Specifications change TSTF-425, ``Relocate Surveillance
Frequencies to Licensee Control--RITSTF [Risk Informed Technical
Specifications Task Force] Initiative 5b,'' Revision 3 (ADAMS Accession
No. ML090850642). The Federal Register notice published on July 6, 2009
(74 FR 31996), announced the availability of TSTF-425, Revision 3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided
its analysis of the issue of no significant hazards consideration,
which is presented as follows:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change relocates the specified frequencies for
periodic surveillance requirements to licensee control under a new
Surveillance Frequency Control Program. Surveillance frequencies are
not an initiator to any accident previously evaluated. As a result, the
probability of any accident previously evaluated is not significantly
increased. The systems and components required by the Technical
Specifications for which the surveillance frequencies are relocated are
still required to be operable, meet the acceptance criteria for the
surveillance requirements, and be [sic] capable of performing any
mitigation function assumed in the accident analysis. As a result, 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 any accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes relocate the surveillance frequencies for
Surveillance Requirements that have a set periodicity from the TS to a
licensee controlled Surveillance Frequency Control Program. This change
does not alter any existing surveillance frequencies. Within the
constraints of the Program, the licensee will be able to change the
periodicity of these surveillance requirements. Relocating the
surveillance frequencies does not impact the ability of structures,
systems or components (SSCs) from performing there [sic] design
functions, and thus, does not create the possibility of a new or
different kind of accident from any previously evaluated.
No new or different accidents result from utilizing the proposed
change. The changes do 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 changes do not impose any new or different requirements. The
changes do not alter assumptions made in the safety analysis
assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction in
the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria for
structures, systems, and components (SSCs) specified in applicable
codes and standards (or alternatives approved for use by the NRC) will
continue to be met as described in the plant licensing basis (including
the final safety analysis report and bases to TS), since these are not
affected by changes to the surveillance frequencies. Similarly, there
is no impact to safety analysis acceptance criteria as described in the
plant licensing basis. To evaluate a change in the relocated
surveillance frequency, FPL will perform a probabilistic risk
evaluation using the guidance contained in NRC-approved NEI 04-10,
Revision 1 in accordance with the TS Surveillance Frequency Control
Program. NEI 04-10, Revision 1, methodology provides reasonable
acceptance guidelines and methods for evaluating the risk increase of
proposed changes to surveillance frequencies consistent with Regulatory
Guide (RG) 1.177, An Approach for Plant-Specific Risk-Informed
Decision-Making: Technical Specifications.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 50.92(c) are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd. MS LAW/JB,
Juno Beach, Florida 33408-0420.
NRC Acting Branch Chief: Lisa M. Regner.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: June 3, 2014. A publicly available
version is in ADAMS under Accession No. ML14154A136.
Description of amendment request: The amendments would revise the
Technical Specification Limiting Condition for Operation 3.3.1 and
Surveillance Requirement 3.2.4.2 regarding the reactor trip system
instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not adversely affect accident initiators or
precursors nor alter the design assumptions, conditions, or
configuration of the facility or the
[[Page 42552]]
manner in which the plant is operated and maintained. The proposed
changes do not alter or prevent the ability of structures, systems, and
components (SSCs) from performing their intended function to mitigate
the consequences of an initiating event within the assumed acceptance
limits. The proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating the
radiological consequences of an accident previously evaluated. Further,
the proposed changes do not increase the types or amounts of
radioactive effluent that may be released offsite, nor significantly
increase individual or cumulative occupational/public radiation
exposures. The proposed changes are consistent with safety analysis
assumptions and resultant consequences.
Therefore, the proposed changes do not increase the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not result in a change in the manner in
which the Reactor Trip System (RTS) and engineered safety features
actuation system (ESFAS) provide plant protection. The RTS and ESFAS
will continue to have the same setpoints after the proposed changes are
implemented. There are no design changes associated with the license
amendment.
The changes do 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 changes do not impose any new or different requirements or
eliminate any existing requirements. The changes do not alter
assumptions made in the safety analysis. The proposed changes are
consistent with the safety analysis assumptions and current plant
operating practice.
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 changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do 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 these changes. Redundant RTS and ESFAS trains are
maintained, and diversity with regard to the signals that provide
reactor trip and engineered safety features actuation is also
maintained. All signals credited as primary or secondary, and all
operator actions credited in the accident analyses will remain the
same. The proposed changes will not result in plant operation in a
configuration outside the design basis.
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: Leigh D. Perry, SVP & General Counsel of
Operations and Nuclear, Southern Nuclear Operating Company, 40 Iverness
Center Parkway, Birmingham, AL 35201.
NRC Branch Chief: Robert J. Pascarelli.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant (HNP), Unit Nos. 1 and 2, Appling County, Georgia
Date of amendment request: March 17, 2014. A publicly-available
version is in ADAMS under Accession No. ML14076A141.
Description of amendment request: The proposed amendments would
modify Technical Specification (TS) definition of Shutdown Margin (SDM)
to require calculation of the SDM at a reactor moderator temperature of
68[emsp14][deg]F or a higher temperature that represents the most
reactive state throughout the operating cycle. This change is needed to
address new Boiling Water Reactor (BWR) fuel designs which may be more
reactive at shutdown temperatures above 68 [deg]F.
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 is presented below:
SNC has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment(s) by focusing on
the three standards set forth in 10 CFR 50.92, Issuance of amendment,
as discussed 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 definition of SDM. SDM is not an
initiator to any accident previously evaluated. Accordingly, the
proposed change to the definition of SDM has no effect on the
probability of any accident previously evaluated. SDM is an assumption
in the analysis of some previously evaluated accidents and inadequate
SDM could lead to an increase in consequences for those accidents.
However, the proposed change revises the SDM definition to ensure that
the correct SDM is determined for all fuel types at all times during
the fuel cycle. As a result, the proposed change does not adversely
affect the consequences of any accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the definition of SDM. The change does
not involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operations. The change does not alter
assumptions made in the safety analysis regarding SDM.
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 definition of SDM. 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 ensures that the SDM assumed in
determining safety limits, limiting safety system settings or limiting
conditions for operation is correct for all BWR fuel types at all times
during the fuel cycle.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, SNC concludes that the proposed change presents
no significant hazards consideration under the standards set forth in
10 CFR 50.92
[[Page 42553]]
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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Branch Chief: Robert Pascarelli.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 And 50-
281 Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of amendment request: April 11, 2014. A publicly-available
version is in ADAMS under Accession No. ML14112A073.
Description of amendment request: The proposed license amendment
requests the changes to the Technical Specification (TS) TS 4.2,
``Augmented Inspections,'' and TS 4.15, ``Augmented Inservice
Inspection Program for High Energy Lines Outside of Containment,'' by
relocating to the Surry Technical Requirements Manual (TRM). In
addition, TS 6.4.U, ``Augmented Inspections and Examinations,'' will be
added to the Administrative Controls Section 6.4, ``Unit Operating
Procedures and Programs.'' The proposed relocation of the TS 4.2 and TS
4.15 requirements to the TRM is appropriate since these requirements do
not satisfy the categories and criteria of 10 CFR 50.36(c) for
inclusion in the TS. Along with the relocation of the TS 4.2 and TS
4.15 requirements to the TRM, the Bases for TS 4.2 and TS 4.15 are also
being relocated to the TRM.
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 relocates Technical Specification (TS) 4.2,
``Augmented Inspections,'' TS 4.15, ``Augmented Inservice Inspection
Program for High Energy Lines Outside of Containment,'' and the
associated TS Bases to the Surry Technical Requirements Manual (TRM).
In addition, TS 6.4.U, ``Augmented Inspections and Examinations,'' will
be added to the Surry TS. The proposed relocation of the TS 4.2 and TS
4.15 requirements to the TRM is appropriate since these requirements do
not satisfy the categories and criteria of 10CFR50.36(c), which
specifies what items qualify for inclusion in the TS.
Specifically, the TS 4.2 augmented inspections of the low head
safety injection piping located in the valve pit, the reactor coolant
pump flywheel, the low pressure turbine rotor blades, sensitized
stainless steel, and TS 4.15 augmented inspections of the welds in the
main steam and main feedwater lines in the main steam valve house of
each unit will be relocated to the TRM. The augmented inspections,
which are performed in addition to required ASME Code Section Xl
inspections/examinations, will continue to be performed as required by
the TRM.
The plant systems and components to which the augmented inspections
apply will not be operated in a different manner. The proposed
relocation of the augmented inspections does not involve a physical
change to the plant or a change in the manner in which the plant is
operated or controlled.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve any physical alteration of
plant equipment. As such, no new or different types of equipment will
be installed, and the basic operation of installed plant systems and
components, to which the augmented inspections apply, 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 proposed change will not reduce a margin of safety because the
relocation of the augmented inspections to the TRM has no impact on any
safety analysis assumptions, as indicated by the fact that the
requirements do not meet the 10CFR50.36(c) criteria for inclusion in
the TS. In addition, the augmented inspections will be moved to the TRM
without change and will continue to be performed as required by the
TRM.
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 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 Street, RS-2, Richmond, VA
23219.
NRC Branch Chief: Robert Pascarelli.
ZionSolutions LLC (ZS), Docket Nos. 50-295 and 50-304, Zion Nuclear
Power Station (ZNPS), Units 1 and 2, Lake County, Illinois
Date of amendment request: May 27, 2014. A publicly-available
version is in ADAMS under Accession No. ML14148A295.
Description of amendment request: The license amendment request
proposes changes to ZNPS Defueled Station Emergency Plan (DSEP) in
accordance with 10 CFR 50.54(q). ZS proposes removal of the various
emergency actions related to the former spent fuel pool, the transfer
of responsibility for implementing the Emergency Plan to the
Independent Spent Fuel Storage Installation (ISFSI) Shift Supervisor, a
revised emergency plan organization, abandonment of the Control Room
consistent with the current state of decommissioning, transition to NEI
99-01 Revision 6 and reformatting consistent with current industry
practice.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. ZS has, in effect, a U.S. Nuclear Regulatory Commission-
approved (NRC) emergency plan. The remaining ZNPS accident (Radioactive
Waste Handling Accident) and the credible accidents involving the ISFSI
and the Modular, Advanced Generation, Nuclear All-purpose Storage
(MAGNASTOR) system have been analyzed and determined that none result
in doses to the public beyond the owner controlled area boundary that
would exceed the U.S. Environmental Protection Agency's (EPA)
Protective Action Guides (PAGs). These analyses have not changed. With
spent fuel relocated to the ISFSI, the Spent Fuel Pool previously
analyzed events (Loss of Spent Fuel Pool Cooling,
[[Page 42554]]
Loss of Spent Fuel Pool Inventory, and Fuel Handling Accident in the
Fuel Building) are no longer credible.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
(2) Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. ZS has, in effect, an NRC-approved emergency plan. The
remaining ZNPS accident (Radioactive Waste Handling Accident) and the
credible accidents involving the ISFSI and MAGNASTOR system have been
analyzed and determined that none result in doses to the public beyond
the owner controlled area boundary that would exceed the EPA's PAGs.
These analyses have not changed. With spent fuel relocated to the
ISFSI, the Spent Fuel Pool previously analyzed events (Loss of Spent
Fuel Pool Cooling, Loss of Spent Fuel Pool Inventory, and Fuel Handling
Accident in the Fuel Building) are no longer credible. Accidents
associated with the ISFSI are addressed in the MAGNASTOR Final Safety
Analysis Report.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
(3) Does the change involve a significant reduction in a margin of
safety?
No. 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. ZS has, in effect, an NRC-approved emergency
plan. The remaining ZNPS accident (Radioactive Waste Handling Accident)
and the credible accidents involving the ISFSI and MAGNASTOR system
have been analyzed and determined that none result in doses to the
public beyond the owner controlled area boundary that would exceed the
EPA's PAGs These analyses have not changed. With spent fuel relocated
to the ISFSI, the Spent Fuel Pool previously analyzed events (Loss of
Spent Fuel Pool Cooling, Loss of Spent Fuel Pool Inventory, and Fuel
Handling Accident in the Fuel Building) are no longer credible.
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: Russ Workman, Deputy General Counsel,
EnergySolutions, 423 West 300 South, Suite 200, Salt Lake City, UT
84101.
NRC Branch Chief: Bruce Watson.
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.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New
York
Date of amendment request: May 23, 2013, as supplemented by letter
dated October 11, 2013.
Brief description of amendment(s): The amendments revise the
Technical Specifications to risk-inform requirements regarding selected
Required Action End States. Specifically, the changes permit an end
state of Mode 4 rather than an end state of Mode 5 consistent with
Technical Specification Task Force (TSTF) Traveler TSTF 432-A, Revision
1, ``Change in Technical Specifications End States WCAP-16294.''
Date of issuance: July 7, 2014.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: Unit 2-275; Unit 3-252. A publicly-available version
is in ADAMS under Accession No. ML14122A303; documents related to these
amendments are listed in the Safety Evaluation enclosed with the
amendments.
Facility Operating License Nos. DPR-26 and DPR-64: The amendment
revised the Facility Operating License and the Technical
Specifications.
Date of initial notice in Federal Register: July 23, 2013 (78 FR
44170). The supplemental letter 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 July 7, 2014.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: June 11, 2013, as supplemented by letter
dated December 11, 2013.
Brief description of amendment: The amendment revised Technical
Specification 2.1.1.1, to add a provision for the determination of the
maximum local fuel pin centerline temperature using the NRC reviewed
and approved COPERNIC fuel performance computer code.
Date of issuance: July 9, 2014.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 249. A publicly-available version is in ADAMS under
Accession No. ML14169A475; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-51: Amendment revised
the
[[Page 42555]]
Technical Specifications and the renewed facility operating license.
Date of initial notice in Federal Register: April 1, 2014 (79 FR
18331). The supplemental letter dated December 11, 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 amendment is contained
in a Safety Evaluation dated July 9, 2014.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: November 8, 2013.
Brief description of amendment: The amendment revised the Technical
Specification (TS) definition of ``Shutdown Margin'' (SDM) to require
calculation of the SDM at a reactor moderator temperature of 68 degrees
Fahrenheit ([deg]F) or a higher temperature that represents the most
reactive state throughout the operating cycle. This change is needed to
address new Boiling Water Reactor (BWR) fuel designs which may be more
reactive at shutdown temperatures above 68 [deg]F.
This TS change is part of the Consolidated Line Item Improvement
Process (CLIIP) TS Task Force (TSTF) Traveler TSTF-535, Revision 0,
``Revise Shutdown Margin Definition to Address Advanced Fuel Designs.''
The licensee stated there are no variations or deviations from the NRC
staff's model safety evaluation.
Date of issuance: June 30, 2014.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 198. A publicly-available version is in ADAMS under
Accession No. ML14106A133; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-29: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 4, 2014 (79 FR
12244).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 30, 2014.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Nine Mile Point Nuclear Station, LLC,
Docket No. 50-220, Nine Mile Point Nuclear Station, Unit No. 1, Oswego
County, New York
Date of application for amendment: June 11, 2012, as supplemented
by letters dated February 27, March 27, April 30, and December 9, 2013;
and January 22, March 14, April 15, May 9, and May 23, 2014.
Brief description of amendment: The amendment authorizes the
transition of the Nine Mile Point Nuclear Station, Unit 1, fire
protection program to a risk-informed, performance-based program based
on National Fire Protection Association (NFPA) 805, in accordance with
10 CFR 50.48(c). NFPA 805 allows the use of performance-based methods
such as fire modeling and risk-informed methods such as fire
probabilistic risk assessment to demonstrate compliance with the
nuclear safety performance criteria.
Date of issuance: June 30, 2014.
Effective date: As of its date of issuance and shall be implemented
by 180 days from the date of issuance.
Amendment No.: 215. A publicly available version is in ADAMS under
Accession No. ML14126A003; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-63: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: September 11, 2012 (77
FR 55874).
The supplements dated February 27, March 27, April 30, and December
9, 2013; and January 22, March 14, April 15, May 9, and May 23, 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 amendment is contained
in a Safety Evaluation dated June 30, 2014.
No significant hazards consideration comments received: No.
Amendment No.: 215. A publicly available version is in ADAMS under
Accession No. ML14126A003; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-63: Amendment revised
the License and Technical Specifications.
Date of initial notice in Federal Register: September 11, 2012 (77
FR 55874).
The supplements dated February 27, March 27, April 30, and December
9, 2013; and January 22, March 14, April 15, May 9, and May 23, 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 amendment is contained
in a Safety Evaluation dated June 30, 2014.
No significant hazards consideration comments received: No.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2, Somervell County,
Texas
Date of amendment request: March 28, 2013, as supplemented by
letters dated July 16, October 22, November 26, and December 17, 2013,
and January 16, April 17, and May 1, 2014.
Brief description of amendment: The amendments revised Technical
Specification (TS) 3.7.16, ``Fuel Storage Pool Boron Concentration,''
TS 3.7.17, ``Spent Fuel Assembly Storage,'' TS 4.3, ``Fuel Storage,''
and TS 5.5, ``Programs and Manuals,'' for storage of uprated fuel in
Region II of the spent fuel pool. Changes to TS 3.7.16 reflect a change
in the required fuel storage pool soluble boron concentration based on
the results of a new criticality analysis. Changes to TS 3.7.17 include
new spent fuel pool loading restrictions in terms of allowable storage
patterns, and minimum burnup requirements as a function of enrichment,
fuel type, and fuel reactivity category. The revised TS 4.3 section
includes updates to the minimum soluble boron concentration, Region I
fuel assembly spacing, specific new or partially spent fuel assembly
storage restrictions in Region II consistent with TS 3.7.17, and
general Region II storage restrictions consistent with TS 3.7.17. The
change to TS 5.5 adds TS program 5.5.22, ``Neutron Absorber Monitoring
Program.''
Date of issuance: July 1, 2014.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: Unit 1-162; Unit 2-162. A publicly-available
version is in ADAMS under Accession No. ML14160A035; documents related
to these amendments are listed in the
[[Page 42556]]
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: November 5, 2013 (78 FR
66391). The NRC staff's original proposed no significant hazards
consideration determination was based on letters dated March 28, and
July 16, 2013. The supplements dated October 22, November 26, and
December 17, 2013, and January 16, April 17, and May 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 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 July 1, 2014.
No significant hazards consideration comments received: No.
NextEra Energy Point Beach, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: January 15, 2013, as supplemented on
March 1, April 18, and September 12, 2013, and March 11, 2014.
Description of amendment: The license amendment revised Technical
Specifications 5.6.5, ``Reactor Coolant System (RCS) Pressure and
Temperature Limits Report (PTLR),'' to allow the use of two new
methodologies for determining RCS pressure and temperature limits at
the Point Beach Nuclear Plant, Units 1 and 2.
Date of issuance: June 30, 2014.
Effective date: As of the date of issuance and shall be implemented
with 180 days.
Amendment Nos.: 250 (Unit 1) and 254 (Unit 2). A publicly-available
version is in ADAMS under Accession No. ML14126A378; documents related
to this amendment are listed in the Safety Evaluation enclosed with the
amendment.
Renewed Facility Operating License Nos. DPR-24 and DPR-27: The
amendment revised the Renewed Facility Operating License and the
Technical Specifications.
Date of initial notice in Federal Register: June 11, 2013 (78 FR
35062). The supplemental letters dated March 1, April 18, and September
12, 2013, and March 11, 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 amendment is contained
in a Safety Evaluation dated June 30, 2014.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of amendment requests: August 12, 2013, as supplemented by
letters dated January 24, March 13, and March 25, 2014.
Brief description of amendments: The licensee requested to revise
the Technical Specifications to, in effect, extend the Type A primary
containment Integrated Leak Rate Test intervals to fifteen years and
the Type C local leak rate test intervals to 75 months, and incorporate
the regulatory positions stated in RG 1.163.
Date of issuance: July 3, 2014.
Effective date: As of its date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: Unit 1, 282; Unit 2, 282. A publicly-available
version is in ADAMS under Accession No. ML14148A235; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Renewed Facility Operating License Nos. DPR-32 and DPR-37: The
amendments revise the Renewed Facility Operating Licenses and the
Technical Specifications.
Date of initial notice in Federal Register: October 29, 2013 (78 FR
64548). The supplemental letters dated January 24, March 13, and March
25, 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 July 3, 2014.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: September 26, 2013.
Brief description of amendment: The amendment revised Technical
Specification Surveillance Requirement (SR) 3.7.10.1 and SR 3.7.13.1 to
reduce the required run time for periodic operation of the control room
pressurization system filter trains and emergency exhaust system filter
trains, with heaters on, from 10 hours to 15 minutes. The amendment is
consistent with plant-specific options provided in the NRC's model
safety evaluation in Technical Specifications Task Force (TSTF)
Traveler TSTF-522, Revision 0, ``Revise Ventilation System Surveillance
Requirements to Operate for 10 hours per Month,'' as part of the
consolidated line item improvement process.
Date of issuance: July 1, 2014.
Effective date: As of its date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 209. A publicly-available version is in ADAMS under
Accession No. ML14175A390; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 21, 2014 (79 FR
3418).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 1, 2014.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 17, 2013.
Brief description of amendment: The amendment revised Technical
Specification Surveillance Requirement (SR) 3.7.10.1 and SR 3.7.13.1 to
reduce the required run time for periodic operation of the control room
pressurization system filter trains and emergency exhaust system filter
trains, with heaters on, from 10 hours to 15 minutes. The amendment is
consistent with plant-specific options provided in the NRC's model
safety evaluation in Technical Specifications Task Force (TSTF)
Traveler TSTF-522, Revision 0, ``Revise Ventilation System Surveillance
Requirements to Operate for 10 hours per Month,'' as part of the
consolidated line item improvement process.
Date of issuance: July 1, 2014.
[[Page 42557]]
Effective date: As of its date of issuance and shall be implemented
within 90 days of issuance.
Amendment No.: 208. A publicly-available version is in ADAMS under
Accession No. ML14157A082; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-42. The amendment
revised the Operating License and Technical Specifications.
Date of initial notice in Federal Register: March 18, 2014 (79 FR
15151).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 1, 2014.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 14th day of July 2014.
For the Nuclear Regulatory Commission.
Louise Lund,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2014-17257 Filed 7-21-14; 8:45 am]
BILLING CODE 7590-01-P