[Federal Register Volume 78, Number 218 (Tuesday, November 12, 2013)]
[Notices]
[Pages 67402-67418]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-27025]
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NUCLEAR REGULATORY COMMISSION
[NRC-2013-0249]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
Background
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
[[Page 67403]]
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 October 17 to October 30, 2013. The last
biweekly notice was published on October 29, 2013 (78 FR 64541).
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-2013-0249. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected].
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: 3WFN-06-A44MP, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001.
For additional direction on accessing information and submitting
comments, see ``Accessing Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
SUPPLEMENTARY INFORMATION:
I. Accessing Information and Submitting Comments
A. Accessing Information
Please refer to Docket ID NRC-2013-0249 when contacting the NRC
about the availability of information regarding this document. You may
access publicly-available information related to this action by the
following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0249.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly available documents online in the NRC
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in
ADAMS by performing a search on the document date and docket number.
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-2013-0249 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. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS, and the NRC does not 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 in their comment submissions
that they do not want to be publicly disclosed. Your request should
state that the NRC will not edit comment submissions to remove such
information before making the comment submissions available to the
public or entering the comment submissions into ADAMS.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in section 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.
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 ``Rules of
Practice for Domestic Licensing Proceedings'' 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. NRC regulations are accessible electronically from the NRC
Library on the NRC 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
[[Page 67404]]
should specifically explain the reasons why intervention should be
permitted with particular reference to the following general
requirements: (1) The name, address, and telephone number of the
requestor or petitioner; (2) the nature of the requestor's/petitioner's
right under the Act to be made a party to the proceeding; (3) the
nature and extent of the requestor's/petitioner's property, financial,
or other interest in the proceeding; and (4) the possible effect of any
decision or order which may be entered in the proceeding on the
requestor's/petitioner's interest. The petition must also identify the
specific contentions which the requestor/petitioner seeks to have
litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment.
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 E-Filing rule (72 FR 49139,
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the Internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to request (1) a digital information (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/apply-certificates.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 Electronic Information Exchange System, users
will be required to install a Web browser plug-in from the NRC 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 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 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 agency'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 Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email at
[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
[[Page 67405]]
exemption request, in accordance with 10 CFR 2.302(g), with their
initial paper filing requesting authorization to continue to submit
documents in paper format. Such filings must be submitted by: (1) First
class mail addressed to the Office of the Secretary of the Commission,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
Attention: Rulemaking and Adjudications Staff; or (2) courier, express
mail, or expedited delivery service to the Office of the Secretary,
Sixteenth Floor, One White Flint North, 11555 Rockville Pike,
Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications
Staff. Participants filing a document in this manner are responsible
for serving the document on all other participants. Filing is
considered complete by first-class mail as of the time of deposit in
the mail, or by courier, express mail, or expedited delivery service
upon depositing the document with the provider of the service. A
presiding officer, having granted an exemption request from using E-
Filing, may require a participant or party to use E-Filing if the
presiding officer subsequently determines that the reason for granting
the exemption from use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. With respect to copyrighted works, except
for limited excerpts that serve the purpose of the adjudicatory filings
and would constitute a Fair Use application, participants are requested
not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the following three
factors in 10 CFR 2.309(c)(1): (i) The information upon which the
filing is based was not previously available; (ii) the information upon
which the filing is based is materially different from information
previously available; and (iii) the filing has been submitted in a
timely fashion based on the availability of the subsequent information.
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the NRC's PDR, located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through ADAMS in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS, should contact the NRC PDR Reference staff at 1-800-397-4209,
301-415-4737, or by email to [email protected].
Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit No. 1, Wake County, North Carolina
Date of amendment request: November 29, 2012, as supplemented by
letter dated January 3, 2013.
Description of amendment request: This is being re-noticed in its
entirety due to an error in the amendment description of the notice
published in the Federal Register on February 19, 2013 (78 FR 11691).
The proposed amendment would revise the degraded voltage time delay
values in Technical Specification (TS) Table 3.3-4. In conjunction with
planned plant modifications and reanalysis of the final safety analysis
design basis large break loss of coolant accident (LOCA), the revisions
would resolve a nonconservative TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. 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 Technical Specifications (TS)
Table 3.3-4, Functional Unit 9.b. Loss of Offsite Power, 6.9 kV
(kilovolt) Emergency Bus Undervoltage--Secondary time delay values.
The Loss of Offsite Power, 6.9 kV (kilovolt) Emergency Bus
Undervoltage--Secondary instrumentation functions are not initiators
to any accident previously evaluated. As such, the probability of an
accident previously evaluated is not increased. The revised values
continue to provide reasonable assurance that the Loss of Offsite
Power, 6.9 kV (kilovolt) Emergency Bus Undervoltage--Secondary
function will continue to perform its intended safety functions. As
a result, the proposed change will not increase the consequences of
an accident previously evaluated.
Concurrent with this proposed change, the Harris Nuclear Plant
is revising its large break loss of coolant accident analysis. The
revised analysis will be evaluated in accordance with 10 CFR 50.59
to confirm that a change to the technical specifications
incorporated in the license is not required, and the change does not
meet any of the criteria in Paragraph (c)(2) of that regulation. The
revised analysis will employ the plant-specific methodology ANP-
3011(P), Harris Nuclear Plant, Unit 1, Realistic Large Break LOCA
Analysis, Revision 1, as approved by NRC Safety Evaluation dated May
30, 2012.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change revises the TS Table 3.3-4, Functional Unit
9.b. Loss of Offsite Power, 6.9 kV (kilovolt) Emergency Bus
Undervoltage--Secondary time delay values. No new operational
conditions beyond those currently allowed are introduced. This
change is consistent with the safety analyses assumptions and
current plant operating practices. This simply corrects the setpoint
consistent with the accident analyses and therefore cannot create
the possibility of a new or different kind of accident from any
previously evaluated accident.
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
the margin of safety?
Response: No.
The proposed change revises the TS Table 3.3-4, Functional Unit
9.b. Loss of Offsite Power, 6.9 kV (kilovolt) Emergency Bus
Undervoltage--Secondary time delay values. This proposed change
implements a reduced time delay to isolate safety buses from offsite
power if a Loss of Coolant Accident were to occur coincident with a
sustained degraded voltage condition. This provides improved margin
to ensure that emergency core cooling system pumps inject water into
the reactor vessel within the time assumed and evaluated in the
accident analysis.
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
[[Page 67406]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: David T. Conley, Manager--Senior Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Jessie F. Quichocho.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida.
Date of amendment request: July 26, 2013, as supplemented by letter
dated October 16, 2013.
Description of amendment request: The proposed amendment would
align St. Lucie TSs with NUREG-1432, Revision 4, Combustion Engineering
Plants Standard Technical Specifications (STSs) describing the
Administrative Controls requirements for the Responsibility and
Organization, which includes Onsite and Offsite Organizations and the
Unit Staff. The proposed amendment will revise TSs 6.1, Responsibility
and 6.2, Organization to be consistent with STSs 5.1 Responsibility and
5.2 Organization, which directly reference the requirements in 10 CFR
50.54(m). The current Units 1 and 2 TSs 6.1 and 6.2 use custom language
to define the requirements of the regulation.
Basis for proposed no significant hazards consideration (NSHC)
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of NSHC, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes involve reformatting, renumbering, and
rewording. The revisions have no technical implications with respect
to the station organization, responsibilities, or unit staffing
requirements. The changes do not affect the minimum shift complement
in any mode of operation nor decrease the effectiveness of the shift
personnel. The proposed changes are minor or editorial in nature and
will not result in any significant increase in the probability of
consequences of an accident as previously evaluated, as the proposed
TS changes are consistent with the NUREG-1432, Combustion
Engineering Plant Standard Technical Specifications. Further, the
proposed changes do not introduce additional risk or greater
potential for consequences of an accident that has not previously
been evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are minor or editorial in nature. The
proposed changes do not involve a physical modification of the plant
or methods governing normal plant operation. No new or different
type of equipment will be installed. The proposed changes will not
introduce new failure modes/effects that could lead to an accident
not previously analyzed. The proposed changes will not impose any
new or change existing requirements that are not consistent with
NUREG-1432, Combustion Engineering Plant Standard Technical
Specifications.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes involve reformatting, renumbering, and
rewording. The revisions have no technical implications with respect
to the station organization, responsibilities, or unit staffing
requirements. The changes do not affect the minimum shift complement
in any mode of operation nor decrease the effectiveness of the shift
personnel. The proposed changes will not involve a significant
reduction in a margin of safety in that the changes are minor or
editorial in nature. No plant equipment or accident analyses will be
affected. Additionally, the proposed changes will not relax any
criteria used to establish safety limits, safety system settings, or
the bases for any limiting conditions for operation. Safety analysis
acceptance criteria are not affected. Plant operation will continue
within the design basis. The proposed changes do not adversely
affect systems that respond to safely shutdown the plant, and
maintain the plant in a safe shutdown condition. Consequently, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Acting Branch Chief: Douglas A. Broaddus.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant (CR-3), Citrus County, Florida
Date of amendment request: April 25, 2013, as supplemented on
September 4, 2013.
Description of amendment request: The proposed license amendment
request would revise certain requirements from Section 5,
``Administrative Controls,'' of the CR-3 Improved Technical
Specifications (ITSs). The revisions would revise and remove certain
requirements in Section 5.1 ``Responsibility,'' 5.2 ``Organization,''
5.6 ``Procedures, Programs and Manuals,'' 5.7 ``Reporting
Requirements,'' and 5.8 ``High Radiation Area,'' that are no longer
applicable to CR-3 in the permanently defueled condition. The September
4, 2013, supplement supersedes the April 25, 2013, application, and
replaces it in its entirety. In addition, the proposed no significant
hazards consideration determination in the basis section below corrects
a typographical numbering error for TS 5.2.1.b (the section was
incorrectly labeled ``5.1.2.b'' in Section 4.1 of Attachment B of the
September 4, 2013, application).
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 for each proposed change, which is presented below:
A. ITS Section 5.1.1:
This section defines the responsible position for overall unit
operation and for approval of each proposed test, experiment, or
modification to systems or equipment that affect stored nuclear fuel
and fuel handling. The responsible position title is changed from
the Plant General Manager to the Plant Manager.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The change reflects that the remaining credible accident is a
fuel handling accident or loss of spent fuel cooling. The change in
the position title of the responsible person is administrative and
cannot increase the probability or consequences of a fuel handling
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change reflects an organizational change to transition from
an operating plant to a permanently defueled plant. Such an
administrative change cannot create a new or different kind of
accident.
[[Page 67407]]
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The position title proposed here does not involve any physical
plant limits or parameters and therefore cannot affect any margin of
safety.
B. ITS Section 5.1.2:
This section identifies the responsibilities for the control
room command function associated with Modes of plant operation, and
is based on personnel positions and qualifications for an operating
plant. It identifies the need for a delegation of authority for
command in an operating plant when the principal assignee leaves the
control room.
This section is being changed to eliminate the MODE dependency
for this function and personnel qualifications associated with an
operating plant. The proposed change establishes the Shift
Supervisor as having command of the shift.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This is a change to the requirements for control room staffing.
In a permanently defueled plant, the fuel handling building accident
is the only credible accident previously evaluated. This action
cannot increase the probability or consequences of a fuel handling
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The changes proposed here for control room staffing cannot
create a new or different kind of accident since they do not change
the function of any plant structures, systems, or components.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The changes proposed here for control room staffing do not
directly involve any limits or parameters and therefore cannot
affect any margin of safety.
C. ITS Section 5.2.1.a:
The introduction to this section identifies that organizational
positions are established that are responsible for the safety of the
nuclear plant.
This is changed to require that positions be established that
are responsible for the safe storage and handling of nuclear fuel.
This change removes the implication that CR-3 can return to
operation.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change in the description of functional responsibility of
organizational positions places emphasis on the safe storage and
handling of nuclear fuel. This focus on their principal
responsibility cannot increase the probability or consequences of a
fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change in the description of functional responsibility of
organizational positions cannot create a new or different kind of
accident since they do not change the function of any plant
structures, systems, or components.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any physical limits or
parameters and therefore cannot affect any margin of safety.
D. ITS Section 5.2.1.b:
This section identifies the organizational position responsible
for overall nuclear plant safety, for the safe operation of the
plant, and for control of activities necessary for the safe
operation and maintenance of the plant.
This section is being changed to recognize that the safety
concerns for a permanently defueled plant are for the safe storage
and handling of nuclear fuel. It changes responsibility for overall
safety for storage and handling of nuclear fuel to the
Decommissioning Director. It changes responsibility for control over
onsite activities necessary for safe handling and storage of nuclear
fuel to the Plant Manager.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change in the description of functional responsibility of
organizational positions places emphasis on the safe storage and
handling of nuclear fuel. This focus on their principal
responsibility cannot increase the probability or consequences of a
fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change in the description of functional responsibility of
organizational positions cannot create a new or different kind of
accident since they do not change the function of any plant
structures, systems, or components.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any physical limits or
parameters and therefore cannot affect any margin of safety.
E. ITS Section 5.2.1.c:
This paragraph addresses the requirement for organizational
independence of the operations, health physics, and quality
assurance personnel from operating pressures.
This is changed to replace ``operating staff'' with ``Certified
Fuel Handlers,'' and to replace ``their independence from operating
pressures'' to ``their ability to perform their assigned
functions.''
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change continues to ensure that personnel in specifically
identified positions retain independence from organizational
pressures and will not increase the probability or occurrence of a
fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
F. ITS Section 5.2.2.a:
This paragraph addresses that one auxiliary nuclear operator
must be assigned to the operating shift whenever fuel is in the
reactor.
Since this can never occur again at CR-3, the minimum
requirement is changed to a minimum crew compliment of one Shift
Supervisor and one Non-certified Operator.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change, in conjunction with new paragraph 5.2.2.f,
continues to ensure that personnel trained and qualified for the
safe handling and storage of nuclear fuel are onsite. This cannot
increase the probability or consequences of a fuel handling
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
G. ITS Section 5.2.2.b:
This paragraph addresses the conditions under which the minimum
shift compliment may be reduced. It contains a reference to 10 CFR
50.54(m) which establishes the minimum requirements for a licensed
operating staff for facility operation.
This reference is removed since CR-3 will not return to
operation in the future, and the requirement for licensed operating
personnel will no longer be required to protect public health and
safety.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change continues to ensure that the minimum shift
compliment of qualified
[[Page 67408]]
personnel will not be decreased for more than a limited period. It
removes the qualification requirements for personnel who are capable
of responding to operating plant transients and accidents. This does
not involve an increase in the probability or consequences of a fuel
handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
H. ITS Section 5.2.2.c:
This paragraph establishes the requirement for one licensed
Reactor Operator to be in the control room when fuel is in the
reactor and for one Senior Reactor Operator to be in the control
room during operating Modes 1-4.
The change establishes the requirements for either a Non-
certified operator or Certified Fuel handler to be in the control
room when fuel is stored in the pools.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change continues to ensure that personnel trained and
qualified for the handling and storage of nuclear fuel man the
control room. This cannot increase the probability or consequences
of a fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
I. ITS Section 5.2.2.d:
This paragraph established the requirement for a person
qualified in Radiation Protection procedures to be onsite when fuel
is in the reactor.
This paragraph is revised to require a person qualified in
Radiation Protection procedures to be onsite during fuel handling
operations and during movement of heavy loads over the fuel storage
racks.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This is an administrative change that cannot affect the
probability of a fuel handling accident. The consequences of a fuel
handling accident are governed by the characteristics of the fuel
element and are not affected by the presence or absence of radiation
protection trained personnel.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
J. ITS Section 5.2.2.e (New):
A new paragraph is added to establish the requirement for having
oversight of fuel handling operations to be performed by a Certified
Fuel Handler.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Certified Fuel Handlers are specifically trained and qualified
to safely handle irradiated fuel. Applying these qualifications to
fuel movement ensures that the probability or consequences of a fuel
handling accident are not increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
K. ITS Section 5.2.2.f (New):
A new paragraph is added to establish that the Shift Supervisor
must be a Certified Fuel Handler.
In the permanently defueled plant, the Certified Fuel Handler is
the senior position on the operating crew. It is not necessary for
the Shift Supervisor to hold a Senior Reactor Operator license if
the plant cannot operate to generate power.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Certified Fuel Handlers are specifically trained and qualified
to safely handle irradiated fuel. Applying these qualifications to
the supervision of fuel movement ensures that the probability or
consequences of a fuel handling accident are not increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
L. ITS Section 5.3.1:
This paragraph is changed to remove the requirements for the
Shift Technical Advisor since that position is only required for a
plant authorized for power operations.
The paragraph retains the previous requirements for the
personnel filling unit staff positions meet or exceed the minimum
qualifications of ANSI [American National Standard Institute] N18.1,
1971, and the Radiation Protection Manager meet or exceed the
qualifications of Regulatory Guide 1.8, September 1975.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Shift Technical Advisor position was established to assist
the control room operating personnel to diagnose the cause and
advise on the response to operating transients and accidents. The
absence of a staff member with those qualifications does not change
the probability or consequences of a fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any physical equipment
limits or parameters and therefore cannot affect any margin of
safety.
M. ITS Section 5.3.2:
This new paragraph is added to identify that responsibility for
the training and retraining of Certified Fuel Handlers is assigned
to the Plant Manager.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This section recognizes the importance of establishing and
maintaining Certified Fuel Handler qualifications and assigns a
manager responsibility for this program. Training and retraining
Certified Fuel Handlers specifically trained to safely handle
nuclear fuel will not increase the probability or consequences of a
fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of
[[Page 67409]]
accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any physical limits or
parameters and therefore cannot affect any margin of safety.
N. ITS Section 5.6.1.1.a:
This section states the requirement for procedures to be
established, implemented and maintained covering various plant
activities.
The scope is reduced to procedures applicable to the safe
handling and storage of nuclear fuel.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The procedures necessary for the safe handling of nuclear fuel
are included in the group of procedures applicable to the safe
storage of nuclear fuel. With these procedures in effect for fuel
handling, the probability or consequences of a fuel handling
accident will not be increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The applicable procedures for the safe storage of nuclear fuel
will direct the correct use of fuel handling equipment. These
procedures are currently in place and have been used effectively for
the safe handling of fuel. These procedures will not direct the use
of plant structures, systems, or components in a different manner,
therefore, they cannot create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
O. ITS Section 5.6.2.3:
In this section, the authority for approval of changes to the
Offsite Dose Calculation Manual (ODCM) is changed from the Plant
General Manager to the Plant Manager consistent with the position
title change in 5.1.1.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This is a change to the requirements for the position
responsible for approving ODCM changes. In a permanently defueled
plant, the fuel handling accident is the only credible accident
previously evaluated. This action cannot increase the probability or
consequences of a fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The change proposed here, identifying a different position
responsible for ODCM change approval, cannot create a new or
different kind of accident since this does not change the function
of any plant structures, systems, or components.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The changes proposed here for ODCM approval do not directly
involve any limits or parameters for operating systems and therefore
cannot affect any margin of safety.
P. ITS Section 5.6.2.4: Primary Coolant Sources Outside
Containment:
This program was established to minimize leakage from portions
of systems outside containment that could contain highly radioactive
fluids during a serious transient or accident.
The program is being eliminated.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The fuel handling accident is the only credible accident for a
permanently defueled plant. This change eliminates an inspection
program that is no longer necessary to limit the consequences of
operating transients and accidents. This change cannot increase the
probability or consequences of the fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
Q. ITS Section 5.6.2.5: Component Cyclic or Transient Limit:
This program provided controls to track cyclic and transient
occurrences to ensure that components were maintained within their
design limits.
This program is being eliminated.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Eliminating an administrative event tracking program cannot
increase the probability of a fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Eliminating an administrative event tracking program cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
R. ITS Section 5.6.2.8: Inservice Inspection Program:
This program required periodic inspections, examinations, and
tests of plant pressure boundary components to ensure their
continued integrity for power operation.
This program is being eliminated.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Inservice Inspection Program does not apply to nuclear fuel
or fuel handling equipment. Therefore eliminating this program
cannot increase the probability or occurrence of a fuel handling
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
For an operating plant the Inservice Inspection Program provided
confidence that plant systems that were either a potential source of
an accident or transient or served to mitigate events continued to
meet their physical requirements. For a permanently shutdown plant,
no transient, or accident can occur, so ending this inspection
program cannot affect any margin of safety.
S. ITS Section 5.6.2.9: Inservice Testing Program:
This program required periodic testing of ASME Code Class 1, 2,
and 3, components including applicable supports in accordance with
the ASME Operations and Maintenance (OM) Code.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The Inservice Testing Program does not apply to nuclear fuel or
fuel handling equipment. Therefore eliminating this program cannot
increase the probability or occurrence of a fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
For an operating plant, the Inservice Testing Program provided
confidence that plant components that were required for safe
[[Page 67410]]
shutdown would perform as expected. For a permanently shutdown
plant, the transients or accidents that would require safe shutdown
equipment cannot occur, so ending this testing program cannot affect
any margin of safety.
T. ITS Section 5.6.2.10: Steam Generator (OTSG) Program:
The Steam Generator Program established and implemented
practices to ensure that OTSG tube integrity was maintained.
This program is being eliminated.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The condition of the steam generator tubes inside the
containment has no effect on fuel handling in the auxiliary building
within the spent fuel pools. Therefore, eliminating the program
cannot increase the probability or occurrence of a fuel handling
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The CR-3 steam generators will remain out of service until
removed from the plant. In this state, the condition of the steam
generator tubes is immaterial and cannot create a new or different
kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
U. ITS Section 5.6.2.11: Secondary Water Chemistry Program:
This program provided controls for monitoring secondary water
chemistry to inhibit steam generator tube degradation and low
pressure turbine disc stress corrosion cracking.
This program is being eliminated.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The secondary piping systems do not interconnect with the fuel
cooling or fuel handling systems. Therefore, eliminating the
Secondary Water Chemistry Program cannot increase the probability or
occurrence of a fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The components this program was intended to protect will no
longer function for power production. Therefore, eliminating this
program cannot affect any margin of safety.
V. ITS Section 5.6.2.13: Explosive Gas and Storage Tank
Radioactivity Monitoring Program:
This program provided controls for potentially explosive gas
mixtures contained in the Radioactive Waste Disposal (WD) System,
and the quantity of radioactivity contained in gas storage tanks or
fed into the offgas treatment system.
This program is being eliminated.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This program is required for an operating plant where hydrogen
and radioactive gases are created and must be controlled. Controlled
release of any gases currently in the tanks, in accordance with
existing procedures, will ensure there will be no hazard to public
health and safety. Therefore, elimination of this program cannot
increase the probability or consequences of a fuel handling
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This program is required for an operating plant where hydrogen
and radioactive gases are created and must be controlled. Controlled
release of any gases currently in the tanks, in accordance with
existing procedures, will ensure there will be no hazard to public
health and safety. Therefore, elimination of this program cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margins of safety.
W. ITS Section 5.6.2.18: Core Operating Limits Report (COLR):
This program established that core operating limits be
established prior to each reload cycle.
This program is being eliminated.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This program for controlling the design and operation of the
reactor core has no bearing on fuel storage after fuel has been
moved into the spent fuel pools. Therefore, eliminating this program
cannot increase the probability or occurrence of a fuel handling
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Since CR-3 can never load a core into the reactor again,
eliminating this control program cannot create a new or different
kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Since CR-3 can never load a core into the reactor again,
eliminating this control program cannot affect any margin of safety.
X. ITS 5.6.2.19: Reactor Coolant System (RCS) Pressure And
Temperature Limits Report (PTLR):
This program ensured that RCS pressure and temperature limits,
including heatup and cooldown rates, criticality, and hydrostatic
and leak test limits, be established and documented in the PTLR.
This program is being eliminated.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This program contains no actions or limits that affect the
storage or handling of nuclear fuel. Therefore, eliminating this
program cannot increase the probability or occurrence of a fuel
handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This report is no longer needed since the reactor coolant system
is not subject to pressurization and the reactor contains no fuel.
Therefore, eliminating this control program cannot create a new or
different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The limits established in this report do not apply to nuclear
fuel stored in the spent fuel pools. Therefore, eliminating this
program cannot affect any margin of safety.
Y. ITS Section 5.6.2.20: Containment Leakage Rate Testing
Program:
This program was established to implement the leakage rate
testing of the containment.
This program is being eliminated in accordance with Regulatory
Guide 1.1.84.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Since fuel can never be returned to the CR-3 containment, ending
containment leakage rate testing cannot increase the probability or
occurrence of a fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change does not introduce any changes to the function of
any plant structures, systems, or components therefore it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This change does not directly involve any limits or parameters
and therefore cannot affect any margin of safety.
Z. ITS Section 5.7.2: Special Reports:
This section is being revised to eliminate reporting
requirements associated with programs that are being eliminated.
[[Page 67411]]
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Eliminating reporting requirements for programs that are no
longer required in a permanently defueled plant cannot increase the
probability or occurrence of a fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Eliminating reporting requirements that are no longer required
cannot create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Eliminating reporting requirements that are no longer required
cannot affect any margin of safety.
AA. ITS Section 5.8.2: High Radiation Area Controls:
Changes one of the personnel responsible for locked high
radiation area key control from the Control Room Supervisor to the
Shift Supervisor.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This is a change to the requirements for the position title
responsible for key control. In a permanently defueled plant, the
fuel handling accident is the only credible accident previously
evaluated. This action cannot increase the probability or
consequences of a fuel handling accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The change proposed here, identifying a different position title
responsible for key control, cannot create a new or different kind
of accident since they do not change the function of any plant
structures, systems, or components.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The changes proposed here for key control do not directly
involve any limits or parameters and therefore cannot affect any
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: Kathryn B. Nolan, 550 South Tryon Street,
Charlotte, North Carolina, 28202.
NRC Branch Chief: Jessie F. Quichocho.
Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220, and 50-410,
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2, Oswego County, New
York
Date of amendment request: October 7, 2013.
Description of amendment request: The proposed amendment modifies
the Nine Mile point Units 1 and 2 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[emsp14][deg]F.
The Nuclear Regulatory Commission (NRC) staff issued a notice of
opportunity for comment in the Federal Register on November 19, 2012;
77 FR 69507, on possible amendments to revise the plant specific TS, to
modify the 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, including a model safety
evaluation and model NSHC [no significant hazards consideration]
determination, using the consolidated line-item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on February 26, 2013 (78 FR 13100). The licensee affirmed the
applicability of the model NSHC determination in its application dated
October 7, 2013, which is presented below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC adopted by the licensee 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 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 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 analysis adopted by the licensee
and, based on this review, it appears that the standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the request for amendment involves NSHC.
Attorney for licensee: Gautam Sen, Senior Counsel, Constellation
Energy Nuclear Group, LLC, 100 Constellation Way, Suite 200C,
Baltimore, MD 21202.
NRC Branch Chief: Robert Beall.
South Carolina Electric and Gas, Docket Nos.: 52-027 and 52-028, Virgil
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: September 25, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer
Nuclear Station (VCSNS) Units 2 and 3 by departing from VCSNS Units 2
and 3 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by
revising reference document APP-OCS-GEH-220, ``AP1000 Human Factors
Engineering Task Support Verification Plan,'' from Revision B to
Revision 1. APP-OCS-GEH-220 is incorporated by
[[Page 67412]]
reference in the UFSAR as a means to implement the activities
associated with the human factors engineering verification and
validation.
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 HFE Task Support Verification Plan is one of several
verification and validation (V&V) activities performed on human-
system interface (HSI) resources and the Operation and Control
Centers System (OCS), where applicable. The Task Support
Verification Plan is used to assess and verify displays and
activities related to normal and emergency operation. The changes
are to the Task Support Verification Plan to clarify the scope and
amend the details of the methodology. The Task Support Verification
Plan does not affect the plant itself. Changing the Plan does not
affect prevention and mitigation of abnormal events, e.g.,
accidents, anticipated operational occurrences, earthquakes, floods
and turbine missiles, or their safety or design analyses. The
Probabilistic Risk Assessment is not affected. No safety-related
structure, system, component (SSC) or function is adversely
affected. The change does not involve nor interface with any SSC
accident initiator or initiating sequence of events, and thus, the
probabilities of the accidents evaluated in the UFSAR are not
affected. Because the changes do not involve any safety-related SSC
or function used to mitigate an accident, the consequences of the
accidents evaluated in the UFSAR are not affected.
Therefore, there is no significant increase in the probability
or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The changes to the Task Support Verification Plan change
information related to validation and verification on Human System
Interface and Operational Control Centers. Therefore, the changes do
not affect the safety-related equipment itself, nor do they affect
equipment which, if it failed, could initiate an accident or a
failure of a fission product barrier. No analysis is adversely
affected. No system or design function or equipment qualification
will be adversely affected by the changes. This activity will not
allow for a new fission product release path, nor will it result in
a new fission product barrier failure mode, nor create a new
sequence of events that would result in significant fuel cladding
failures. In addition, the changes do not result in a new failure
mode, malfunction, or sequence of events that could affect safety or
safety-related equipment.
Therefore, this activity does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The changes to the Task Support Verification Plan affect the
validation and verification on the Human System Interface and the
Operational Control Centers. Therefore, the changes do not affect
the plant itself. These changes do not affect the design or
operation of safety-related equipment or equipment whose failure
could initiate an accident, nor does it adversely interface with
safety-related equipment or fission product barriers. No safety
analysis or design basis acceptance limit/criterion is challenged or
exceeded by the requested change.
Therefore, the changes do not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart.
South Carolina Electric & Gas, Docket Nos.: 52-027 and 52-028, Virgil
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: September 25, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer
Nuclear Station (VCSNS) Units 2 and 3 by departing from VCSNS Units 2
and 3 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by
revising reference document APP-OCS-GEH-120, ``AP1000 Human Factors
Design Engineering Verification Plan,'' from Revision B to Revision 1.
APP-OCS-GEH-120 is incorporated by reference in the updated UFSAR as a
means to implement the activities associated with the human factors
engineering verification and validation.
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.
Design verification provides a final check of the adequacy of
the Human System Interface (HSI) Resources and Operation and Control
Centers System (OCS) design. The changes do not affect the plant
itself, and so there is no change to the probability or consequences
of an accident previously evaluated. Changing the design
verification plan does not affect prevention and mitigation of
abnormal events, e.g., accidents, anticipated operational
occurrences, earthquakes, floods and turbine missiles, or their
safety or design analyses as the purpose of the plan is simply to
verify implementation of design criteria. The Probabilistic Risk
Assessment is not affected. No safety-related structure, system,
component (SSC) or function is adversely affected. The change does
not involve nor interface with any SSC accident initiator or
initiating sequence of events, and thus, the probabilities of the
accidents evaluated in the UFSAR are not affected. Because the
changes do not involve any safety-related SSC or function used to
mitigate an accident, the consequences of the accidents evaluated in
the UFSAR are not affected.
Therefore, there is no 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.
Design verification provides a final check of the adequacy of
the HSI Resources and Operation and Control Centers System design.
The changes do not affect the plant itself, and so there is no new
or different kind of accident from any accident previously
evaluated. Therefore, the changes do not affect safety-related
equipment, nor does it affect equipment which, if it failed, could
initiate an accident or a failure of a fission product barrier. No
analysis is adversely affected. No system or design function or
equipment qualification is adversely affected by the changes. This
activity will not allow for a new fission product release path, nor
will it result in a new fission product barrier failure mode, nor
create a new sequence of events that would result in significant
fuel cladding failures. In addition, the changes do not result in a
new failure mode, malfunction, or sequence of events that could
affect safety or safety-related equipment.
Therefore, this activity does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The changes to the design verification plan provide a final
check of the adequacy of the HSI Resources and Operation and Control
Centers System design. The changes do not affect the assessments or
the plant itself. The changes do not affect safety-related equipment
or equipment whose failure could initiate an accident, nor does it
adversely interface with safety-related equipment or
[[Page 67413]]
fission product barriers. No safety analysis or design basis
acceptance limit/criterion is challenged or exceeded by the
requested change.
Therefore, the changes do not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart.
South Carolina Electric and Gas, Docket Nos.: 52-027 and 52-028, Virgil
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: October 3, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer
Nuclear Station (VCSNS) Units 2 and 3 by departing from VCSNS Units 2
and 3 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by
revising material by revising reference document APP-OCS-GEH-520,
``AP1000 Plant Startup Human Factors Engineering Design Verification
Plan,'' from Revision B to Revision 2. APP-OCS-GEH-520 is incorporated
by reference in the UFSAR as a means to implement the activities
associated with the human factors engineering verification and
validation.
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 APP-OCS-GEH-520, document confirms aspects of the human
system interface (HSI) and Operation and Control Centers Systems
(OCS) design features that could not be evaluated in other Human
Factors Engineering (HFE) verification and validation (V&V)
activities. It also confirms that the as-built in the plant HSIs,
procedures, and training conform to the design that resulted from
the HFE program. Additionally, it confirms that all HFE-related
issues (including human error discrepancies (HEDs)) documented in
the SmartPlant Foundation (SPF) Human Factors (HF) Tracking System
are verified as adequately addressed or resolved. Finally, it
confirms the HFE adequacy for risk-important human actions in the
local plant, including the ability for the tasks to be completed
within the time window according to the Probabilistic Risk
Assessment (PRA). The changes to the plan are to clarify the scope
and amend the details of the methodology. The plan does not affect
the plant itself. Changing the plan does not affect prevention and
mitigation of abnormal events, e.g., accidents, anticipated
operational occurrences, earthquakes, floods and turbine missiles,
or their safety or design analyses. The PRA is not affected. No
safety-related Structure, System, or Component (SSC) or function is
adversely affected. The document revision change does not involve
nor interface with any SSC accident initiator or initiating sequence
of events, and thus, the probabilities of the accidents evaluated in
the Updated Final Safety Analysis Report (UFSAR) are not affected.
Because the changes to the plan do not involve any safety-related
SSC or function used to mitigate an accident, the consequences of
the accidents evaluated in the UFSAR are not affected.
Therefore, there is no 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.
APP-OCS-GEH-520, ``AP1000 Plant Startup Human Factors
Engineering Design Verification Plan'' is the plan to confirm
aspects of the HSI and OCS design features that could not be
evaluated in other HFE V&V activities. The plan also confirms that
the as-built in the plant HSIs, procedures, and training conform to
the design that resulted from the HFE program. Additionally, it
confirms that all HFE-related issues (including HEDs) documented in
the SPF HF Tracking System are verified as adequately addressed or
resolved. Finally, it confirms the HFE adequacy for risk-important
human actions in the local plant, including the ability for the
tasks to be completed within the time window according to the PRA.
These functions support evaluating the HSI and OCS. Therefore, the
changes do not affect the safety-related equipment itself, nor do
they affect equipment which, if it failed, could initiate an
accident or a failure of a fission product barrier. No analysis is
adversely affected. No system or design function or equipment
qualification will be adversely affected by the changes. This
activity will not allow for a new fission product release path, nor
will it result in a new fission product barrier failure mode, nor
create a new sequence of events that would result in significant
fuel cladding failures. In addition, the changes do not result in a
new failure mode, malfunction, or sequence of events that could
affect safety or safety-related equipment.
Therefore, this activity does not create the possibility of a
new or different kind of accident than any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
APP-OCS-GEH-520, ``AP1000 Plant Startup Human Factors
Engineering Design Verification Plan'' is the plan to confirm
aspects of the HSI and OCS design features that could not be
evaluated in other HFE V&V activities. The plan also confirms that
the as-built in the plant HSIs, procedures, and training conform to
the design that resulted from the HFE program. Additionally, it
confirms that all HFE-related issues (including HEDs) documented in
the SPF HF Tracking System are verified as adequately addressed or
resolved. Finally, it confirms the HFE adequacy for risk-important
human actions in the local plant, including the ability for the
tasks to be completed within the time windows in the PRA. These
functions support evaluating the HSI and OCS. The proposed changes
to the plan do not affect the design or operation of safety-related
equipment or equipment whose failure could initiate an accident, nor
does the plan adversely affect the interfaces with safety-related
equipment or fission product barriers. No safety analysis or design
basis acceptance limit/criterion is challenged or exceeded by the
requested changes.
Therefore, the changes do not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart.
South Carolina Electric & Gas, Docket Nos.: 52-027 and 52-028, Virgil
C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield County,
South Carolina
Date of amendment request: October 3, 2013.
Description of amendment request: The proposed change would amend
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer
Nuclear Station (VCSNS) Units 2 and 3 by departing from VCSNS Units 2
and 3 Updated Final Safety Analysis Report (UFSAR) Tier 2* material by
revising reference document APP-OCS-GEH-420, ``AP1000 Human Factors
Engineering Discrepancy Resolution Process,'' from Revision B to
Revision 1. APP-OCS-GEH-420 is incorporated by reference in the UFSAR
as a means to implement the activities associated with the human
factors engineering
[[Page 67414]]
verification and validation (TAC No. RQ0403) (LAR 13-18).
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 HFE Discrepancy Resolution Process is used to capture and
resolve Human Engineering Discrepancies (HEDs) identified during the
Human Factors Engineering (HFE) verification and validation (V&V)
activities. These discrepancy resolution process activities are used
to support the final check of the adequacy of the HFE design of the
Human-System Interface (HSI) resources and the Operation and Control
Centers Systems (OCS) design. The discrepancy resolution process
activities are performed as part of the V&V activities against the
final configuration and control documentation, simulator or
installed target system. The changes are to the Discrepancy
Resolution Process to clarify the scope and amend the details of the
methodology. The Discrepancy Resolution Process does not affect the
plant itself. Changing the Discrepancy Resolution Process does not
affect prevention and mitigation of abnormal events, e.g.,
accidents, anticipated operational occurrences, earthquakes, floods
and turbine missiles, or their safety or design analyses. No safety-
related structure, system, component (SSC) or function is adversely
affected. The document revision does not involve nor interface with
any SSC accident initiator or initiating sequence of events, and
thus the probabilities of the accidents evaluated in the Updated
Final Safety Analysis Report (UFSAR) are not affected. Because the
changes do not involve any safety-related SSC or function used to
mitigate an accident, the consequences of the accidents evaluated in
the UFSAR are not affected.
Therefore, there is no significant increase in the probability
or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The changes to the Discrepancy Resolution Process information
are related to discrepancy resolution of HEDs during the HFE V&V
activities on the HSI and the OCS. Therefore, the changes do not
affect the safety-related equipment itself, nor do they affect
equipment which, if it failed, could initiate an accident or a
failure of a fission product barrier. No analysis is adversely
affected. No system or design function or equipment qualification
will be adversely affected by the changes. This activity will not
allow for a new fission product release path, nor will it result in
a new fission product barrier failure mode, nor create a new
sequence of events that would result in significant fuel cladding
failures. In addition, the changes do not result in a new failure
mode, malfunction, or sequence of events that could affect safety or
safety-related equipment.
Therefore, this activity does not create the possibility of a
new or different kind of accident than any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The changes to the Discrepancy Resolution Process affect
discrepancy resolution of HEDs during the HFE V&V activities on the
HSI and the OCS. Therefore, the changes do not affect the
assessments or the plant itself. These changes do not affect the
design or operation of safety-related equipment or equipment whose
failure could initiate an accident, nor does it adversely interface
with safety-related equipment or fission product barriers. No safety
analysis or design basis acceptance limit/criterion is challenged or
exceeded by the requested change.
Therefore, the changes do not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis &
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
NRC Branch Chief: Lawrence Burkhart.
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 are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Room O1-F21, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. Publicly available documents created or received at the
NRC are accessible electronically through the Agencywide Documents
Access and Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by
email to [email protected].
Carolina Power and Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: April 20, 2013.
Brief description of amendment: The amendments revised the
corporate name of the licensee in each facility's operating license
from Carolina Power & Light Company to Duke Energy Progress, Inc.
Date of issuance: October 21, 2013.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 263, 291, 142, and 236.
Renewed Facility Operating License Nos. DPR-71, DPR-62, NPF-63, and
DPR-23: Amendments revised the Licenses and Appendix cover pages.
[[Page 67415]]
Dates of initial notice in Federal Register: May 28, 2013 (78 FR
31982) and correction to initial notice on June 21, 2013 (78 FR 37595).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 21, 2013.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: March 20, 2013.
Brief description of amendment: The amendment changes the name of
the Licensee in the Facility Operating License.
Date of issuance: October 18, 2013.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment No.: 243.
Facility Operating License No. DPR-72: Amendment revises the
Facility Operating License.
Date of initial notice in Federal Register: April 30, 2013 (78 FR
25314).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 18, 2013.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: July 1, 2011, as supplemented
by letters dated September 2, 2011, April 27, 2012, June 29, 2012,
August 9, 2012, October 15, 2012, November 9, 2012, January 14, 2013,
February 1, 2013, May 1, 2013, June 21, 2013, and September 16, 2013.
Brief description of amendments: The amendments revised the
facility operating licenses and transitions the Donald C. Cook Nuclear
Plant fire protection program to a new risk-informed, performance-based
alternative in accordance with 10 CFR 50.48(c), which incorporates by
reference the National Fire Protection Association (NFPA) Standard 805
(NFPA 805), ``Performance-Based Standard for Fire Protection for Light
Water Reactor Electric Generating Plants--2001.''
Date of issuance: October 24, 2013.
Effective date: As of the date of issuance and shall be implemented
by October 24, 2014.
Amendment Nos.: Unit 1--322; Unit 2--305.
Facility Operating License No. DPR-58 and DPR-74: Amendments
revised the Renewed Facility Operating Licenses.
Date of initial notice in Federal Register: October 4, 2011 (76 FR
61396). The supplemental letters dated September 2, 2011, April 27,
2012, June 29, 2012, August 9, 2012, October 15, 2012, November 9,
2012, January 14, 2013, February 1, 2013, May 1, 2013, June 21, 2013,
and September 16, 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 October 24, 2013.
No significant hazards consideration comments received: No.
NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit
No. 1, Rockingham County, New Hampshire
Date of amendment request: March 27, 2013, as supplemented June 25,
2013.
Brief description of amendment: The amendment revised the Seabrook
TS. The amendment modifies TS requirements regarding steam generator
tube inspections and reporting as described in TS Task Force (TSTF)-
510, Revision 2, ``Revision to Steam Generator Program Inspection
Frequencies and Tube Sample Selection,'' using the Consolidated Line
Item Improvement Process (CLIIP). The changes are consistent with
Industry/TSTF Standard Technical Specification Change Traveler, TSTF-
510. The availability of this TS improvement was announced in the
Federal Register on October 27, 2011 (76 FR 66763), as part of the
CLIIP.
Date of issuance: October 25, 2013.
Effective date: As of its date of issuance and shall be implemented
within 60 days.
Amendment No.: 138.
Facility Operating License No. NPF-86: The amendment revised the
Facility Operating License and TS.
Date of initial notice in Federal Register: April 30, 2013 (78 FR
25316). The supplement dated June 25, 2013, 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 October 25, 2013.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc. Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County,
Georgia
Date of amendment request: June 19, 2013, and revised by the letter
dated August 27, 2013.
Brief description of amendment: The proposed amendment would depart
from VEGP Units 3 and 4 plant-specific Design Control Document (DCD)
Tier 2* and associated Tier 2 material incorporated into the Updated
Final Safety Analysis Report (UFSAR) by revising requirements for
design spacing of shear studs and the design of structural elements in
order to address interferences and obstructions other than wall
openings.
Date of issuance: October 8, 2013.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: Unit 3-14, and Unit 4-14.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised
the Facility Combined Licenses.
Date of initial notice in Federal Register: August 6, 2013 (78 FR
47792).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 8, 2013.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Combined Licenses and Final Determination of No Significant Hazards
Consideration and Opportunity for a Hearing (Exigent Public
Announcement or Emergency Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish,
[[Page 67416]]
for public comment before issuance, its usual notice of consideration
of issuance of amendment, proposed no significant hazards consideration
determination, and opportunity for a hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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 51.12(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
application for amendment, (2) the amendment to Facility Operating
License or Combined License, as applicable, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment, as
indicated. All of these items are available for public inspection at
the NRC's Public Document Room (PDR), located at One White Flint North,
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland
20852. Publicly available documents created or received at the NRC are
accessible electronically through the Agencywide Documents Access and
Management System (ADAMS) in the NRC Library at http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
PDR Reference staff at 1-800-397-4209, 301-415-4737 or by email to
[email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. 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 ``Rules of Practice for Domestic Licensing
Proceedings'' 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, and electronically on the
Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the document,
contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737, or
by email to [email protected]. 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 petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. 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 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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is
[[Page 67417]]
requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
All documents filed in the 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 E-Filing rule (72 FR 49139,
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the Internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to request (1) a digital information (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 NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in 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 Electronic Information Exchange System, users
will be required to install a Web browser plug-in from the NRC 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 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 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 agency'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 Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email at
[email protected], or by a toll-free call at 1-866-672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants. Filing is considered
complete by first-class mail as of the time of deposit in the mail, or
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. With respect to copyrighted works, except
for limited excerpts that serve the purpose of the adjudicatory filings
and would constitute a Fair Use application, participants are requested
not to include copyrighted materials in their submission.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of amendment request: October 7, 2013, as supplemented by
letters dated October 8 and October 9, 2013.
Description of amendment request: This notice was previously
published in the Federal Register on October 29, 2013 (78 FR 64550).
This notice is being reissued in its entirety as it was
[[Page 67418]]
inadvertently placed in the incorrect section of the Biweekly report
published on October 29, 2013. The amendment revised Technical
Specification (TS) 3.6.9, ``Distributed Ignition System (DIS),'' to
allow Train B of the DIS to be considered operable with two inoperable
ignitors. The existing TS defines train operability as having no more
than one ignitor inoperable. The amendment also allows one of five
specific primary containment regions to have zero ignitors operable.
The existing TS requires that at least one ignitor be operable in each
region. The proposed TS revision is applicable until the fall 2014
refueling outage, or until the unit enters a mode that allows
replacement of the affected ignitors without exposing personnel to
significant radiation and safety hazards.
Date of issuance: October 9, 2013.
Effective date: As of the date of issuance, to be implemented
within 1 day.
Amendment No.: 321.
Renewed Facility Operating License No. DPR-58: Amendment revised
the Technical Specifications and License.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, state consultation, and final NSHC
determination are contained in a safety evaluation dated October 9,
2013.
Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
NRC Branch Chief: Robert D. Carlson.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: October 6, 2013, as supplemented by
letters dated October 15, 21, and 22, 2013 and two letters dated
October 23, 2013.
Description of amendment request: The amendment revised the Updated
Safety Analysis Report (USAR) for pipe break criteria for high energy
piping outside of containment. Specifically, the proposed amendment
would allow the use of NRC guidance provided in Branch Technical
Position Mechanical Engineering Branch 3-1, Revision 2, which allows
for the exemption of specific piping sections from postulated failures
if certain criteria are met.
Date of issuance: October 25, 2013.
Effective date: As of its issuance date and shall be implemented
upon approval.
Amendment No.: 273.
Renewed Facility Operating License No. DPR-40: The amendment
revised the facility operating license.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (Omaha-World Herald, located in Omaha,
Nebraska, from October 9 through October 15, 2013). The notice provided
an opportunity to submit comments on the Commission's proposed NSHC
determination. One comment was received and evaluated.
The supplemental letters dated October 15, 21, and 22, 2013, and
two letters dated October 23, 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 Omaha-World Herald from October 9 through 15, 2013.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
(including the comment received on the NSHC) are contained in a safety
evaluation dated October 25, 2013.
Attorney for licensee: David A. Repka, Esq., Winston & Strawn, 1700
K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Dated at Rockville, Maryland, this 1st day of November 2013.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2013-27025 Filed 11-8-13; 8:45 am]
BILLING CODE 7590-01-P