Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 44167-44179 [2013-17370]
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NOV and a 1-year ban. Therefore, the
staff believes that, depending on the
significance of an individual’s actions,
the use of other sanctions in individual
enforcement actions warrants further
review. For example, two possible
alternatives whose impacts would fall
between those of an NOV and a 1-year
ban could be issuing a civil penalty or
a ban of 6 months.
Therefore, the staff intends to evaluate
advantages and disadvantages of
expanding the use of civil penalties in
cases involving deliberate misconduct
by individuals and of issuing bans for
less than 1 year. In considering these
options, the staff is soliciting public
comment on both the concept and
possible specifics related to a potential
revision to the Enforcement Policy and
other program documents describing
these alternatives. Specifically, the staff
is seeking stakeholder input including
but, not limited to, the following:
• Given that an individual who has
engaged in deliberate misconduct is
offered the opportunity to participate in
the NRC’s Alternative Dispute
Resolution (ADR) process, in which
modifications to an individual sanction
can include a ban for less than 1 year
or a civil penalty, is there a benefit to
modifying the Enforcement Policy?
• When individual action is deemed
necessary, how should the NRC
determine whether that action should be
an NOV, a civil penalty, or a ban?
• What is the risk of an employer
simply ‘‘reimbursing’’ an individual for
a civil penalty if production is put
ahead of safety? Should the NRC be
concerned with such a potential and, if
so, how would it be mitigated?
• Regarding the amount of a civil
penalty issued to individuals, how can
the NRC assure that the Enforcement
Policy would be applied in a fair and
consistent manner? Specifically, how
should the amount of a civil penalty be
determined? Should a set individual
civil penalty amount be used, or should
the individual civil penalty amount be
calculated based on specific factors:
Æ If a set individual civil penalty
amount should be used, what would be
the appropriate amount? Would it be
fair to propose the same civil penalty
amount on individuals regardless of
salaries?
Æ If a variable individual civil penalty
amount should be used, what factors
(e.g. salary level of individual, safety
significance of violation, benefit or
hardship to the individual, etc.) should
be considered, and how should they be
included in the calculation?
• With respect to the use of either
civil penalties or bans for less than 1
year, would there be any unintended
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consequences the NRC should consider?
If so, provide examples.
Based on the written comments
received from stakeholders, the staff
may conduct a public meeting to
provide for further discussions. The
NRC will use any public input received
as part of its evaluation to determine the
merits and potential implications of
expanding the use of civil penalties in
cases involving deliberate misconduct
by individuals and of issuing bans for
less than 1 year, including the feasibility
of developing criteria to ensure their fair
and consistent application. Following
its evaluation, the staff may propose
changes to the Enforcement Policy to
the Commission for its consideration.
Dated at Rockville, Maryland, this 16th day
of July 2013.
For the Nuclear Regulatory Commission.
Roy P. Zimmerman,
Director, Office of Enforcement.
[FR Doc. 2013–17641 Filed 7–22–13; 8:45 am]
A. Accessing Information
NUCLEAR REGULATORY
COMMISSION
[NRC–2013–0158]
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 (the Commission or NRC)
is publishing this regular biweekly
notice. The Act requires that the
Commission publish notice of any
amendments issued or proposed to be
issued and grants the Commission the
authority to issue and make
immediately effective any amendment
to an operating license or combined
license, as applicable, upon a
determination by the Commission that
such amendment involves no significant
hazards consideration, notwithstanding
the pendency before the Commission of
a request for a hearing from any person.
This biweekly notice includes all
notices of amendments issued, or
proposed to be issued from June 27,
2013 to July 10, 2013. The last biweekly
notice was published on July 9, 2013 (78
FR 41118).
ADDRESSES: You may submit comment
by any of the following methods (unless
this document describes a different
method for submitting comments on a
specific subject):
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• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2103–0158. Address
questions about NRC dockets to Carol
Gallagher; telephone: 301–492–3668;
email: Carol.Gallagher@nrc.gov. For
technical questions, contact the
individual(s) listed in the FOR FURTHER
INFORMATION CONTACT section of this
document.
• Mail comments to: Cindy Bladey,
Chief, Rules, Announcements, and
Directives Branch (RADB), Office of
Administration, Mail Stop: 3WFN–06A–
44MP, 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
BILLING CODE 7590–01–P
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Please refer to Docket ID NRC–2013–
0158 when contacting the NRC about
the availability of information regarding
this document. You may access
information related to this document,
which the NRC possesses and is
publicly available, by the following
methods:
• Federal Rulemaking Web site: Go to
https://www.regulations.gov and search
for Docket ID NRC–2013–0158.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may access publicly
available documents online in the NRC
Library at https://www.nrc.gov/readingrm/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 pdr.resource@nrc.gov.
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–
0158 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.
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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 https://
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
Title 10 of the Code of Federal
Regulations (10CFR), Section 50.92, 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 60day 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
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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 https://www.nrc.gov/reading-rm/
doc-collections/cfr/. If a request for a
hearing or petition for leave to intervene
is filed by the above date, the
Commission or a presiding officer
designated by the Commission or by the
Chief Administrative Judge of the
Atomic Safety and Licensing Board
Panel, will rule on the request and/or
petition; and the Secretary or the Chief
Administrative Judge of the Atomic
Safety and Licensing Board will issue a
notice of a hearing or an appropriate
order.
As required by 10 CFR 2.309, a
petition for leave to intervene shall set
forth with particularity the interest of
the petitioner in the proceeding, and
how that interest may be affected by the
results of the proceeding. The petition
should specifically explain the reasons
why intervention should be permitted
with particular reference to the
following general requirements: (1) The
name, address, and telephone number of
the requestor or petitioner; (2) the
nature of the requestor’s/petitioner’s
right under the Act to be made a party
to the proceeding; (3) the nature and
extent of the requestor’s/petitioner’s
property, financial, or other interest in
the proceeding; and (4) the possible
effect of any decision or order which
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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
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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 EFiling 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
hearing.docket@nrc.gov, 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 NRCissued 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 https://
www.nrc.gov/site-help/e-submittals/
apply-certificates.html. System
requirements for accessing the ESubmittal server are detailed in the
NRC’s ‘‘Guidance for Electronic
Submission,’’ which is available on the
agency’s public Web site at https://
www.nrc.gov/site-help/esubmittals.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,
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is available on the NRC’s public Web
site at https://www.nrc.gov/site-help/esubmittals.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
https://www.nrc.gov/site-help/esubmittals.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 https://
www.nrc.gov/site-help/esubmittals.html, by email at
MSHD.Resource@nrc.gov, or by a tollfree 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,
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44169
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 firstclass mail as of the time of deposit in
the mail, or by courier, express mail, or
expedited delivery service upon
depositing the document with the
provider of the service. A presiding
officer, having granted an exemption
request from using E-Filing, may require
a participant or party to use E-Filing if
the presiding officer subsequently
determines that the reason for granting
the exemption from use of E-Filing no
longer exists.
Documents submitted in adjudicatory
proceedings will appear in the NRC’s
electronic hearing docket which is
available to the public at https://
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
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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 https://
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
pdr.resource@nrc.gov.
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Entergy Nuclear Operations, Inc.,
Docket Nos. 50–247 and 50–286, Indian
Point Nuclear Generating Unit Nos. 2
and 3, Westchester County, New York
Date of amendment request: May 23,
2013.
Description of amendment request:
The proposed change would modify
Technical Specifications (TS) to riskinform requirements regarding selected
Required Action End States.
Specifically, the proposed change
would permit an end state of Mode 4
rather than an end state of Mode 5
contained in the current TS. The
proposed changes are consistent with
NRC-approved Technical Specification
Task Force (TSTF) Technical Change
Traveler 432–A Revision 1, ‘‘Change in
Technical Specifications End States
WCAP–16294.’’ This traveler revised the
Improved Standard Technical
Specifications.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change modifies the end
state (e.g., mode or other specified condition)
which the Required Actions specify must be
entered if compliance with the Limiting
Conditions for Operation (LCO) is not
restored. The requested Technical
Specifications (TS) permit an end state of
Mode 4 rather than an end state of Mode 5
contained in the current TS. In some cases,
other Conditions and Required Actions are
revised to implement the proposed change.
Required Actions are not an initiator of any
accident previously evaluated. Therefore, the
proposed change does not affect the
probability of any accident previously
evaluated. The affected systems continue to
be required to be operable by the TS and the
Completion Times specified in the TS to
restore equipment to operable status or take
other remedial Actions remain unchanged.
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WCAP–16294–NP–A, Rev. 1, ‘‘Risk-Informed
Evaluation of Changes to Tech Spec Required
Action End states for Westinghouse NSSS
PWRs,’’ demonstrates that the proposed
change does not significantly increase the
consequences of any accident previously
evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change modifies the end
state (e.g., mode or other specified condition)
which the Required Actions specify must be
entered if compliance with the LCO is not
restored. In some cases, other Conditions and
Required Actions are revised to implement
the proposed change. The change does not
involve a physical alteration of the plant (i.e.,
no new or different type of equipment will
be installed) or a change in the methods
governing normal plant operation. In
addition, the change does not impose any
new requirements. The change does not alter
assumptions made in the safety analysis.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The proposed change modifies the end
state (e.g., mode or other specified condition)
which the Required Actions specify must be
entered if compliance with the LCO is not
restored. In some cases, other Conditions and
Required Actions are revised to implement
the proposed change. Remaining within the
Applicability of the LCO is acceptable
because WCAP–16294–NP–A demonstrates
that the plant risk in MODE 4 is similar to
or lower than MODE 5. As a result, no margin
of safety is significantly affected.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. William C.
Dennis, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440
Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Robert Beall,
Acting.
Entergy Operations, Inc., Docket No. 50–
313, Arkansas Nuclear One, Unit No. 1,
Pope County, Arkansas
Date of amendment request: March
26, 2013.
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Description of amendment request:
The amendment request would
incorporate the NRC-approved
Technical Specifications Task Force
(TSTF) change traveler TSTF–431,
Revision 3, ‘‘Change in Technical
Specifications End States (BAW–2441),’’
and modify the Technical Specification
(TS) requirements for end states
associated with the implementation of
the approved B&W Owners Group
(B&WOG) Topical Report BAW–2441–
A, Revision 2, ‘‘Risk-Informed
Justification for LCO End-State
Changes,’’ January 2004, as well as
Required Actions revised by a specific
Note in TSTF–431, Revision 3. The TS
Actions End States modifications would
permit, for some systems, entry into a
hot shutdown (Mode 4) end state rather
than a cold shutdown (Mode 5) end
state that is the current TS requirement.
The NRC issued a ‘‘Notice of
Availability of the Models for PlantSpecific Adoption of Technical
Specifications Task Force (TSTF)
Traveler TSTF–431, Revision 3, ‘Change
in Technical Specifications End States
(BAW–2441),’ ’’ in the Federal Register
on December 6, 2010 (75 FR 75705–
75706), which included the no
significant hazards consideration, safety
evaluation, and required commitments
for the proposed changes as part of the
consolidated line item improvement
process (CLIIP).
In its application dated March 26,
2013, the licensee has concluded that
the technical basis presented in the
TSTF proposal and the safety evaluation
are applicable to Arkansas Nuclear One,
Unit 1, and the proposed amendment is
consistent with the Standard Technical
Specifications (STS) changes described
in TSTF–431, Revision 3, but with
certain variations and/or deviations
from TSTF–431, Revision 3.
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 allows a change to
certain required end states when the
Technical Specification (TS) Completion
Times (CTs) for remaining in power
operation are exceeded. Most of the
requested TS changes are to permit an end
state of hot shutdown (Mode 4) rather than
an end state of cold shutdown (Mode 5)
contained in the current TS. The request was
limited to: 1) those end states where entry
into the shutdown mode is for a short
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interval, 2) entry is initiated by inoperability
of a single train of equipment or a restriction
on a plant operational parameter, unless
otherwise stated in the applicable TS, and 3)
the primary purpose is to correct the
initiating condition and return to power
operation as soon as is practical. Risk
insights from both the qualitative and
quantitative risk assessments were used in
specific TS assessments. Such assessments
are documented in Sections 4 and 5 of BAW–
2441–A, Revision 2, ‘‘Risk Informed
Justification for LCO end-state Changes,’’ for
B&W Plants. The assessments provide an
integrated discussion of deterministic and
probabilistic issues, focusing on specific TSs,
which are used to support the proposed TS
end state and associated restrictions. The
staff finds that the risk insights support the
conclusions of the specific TS assessments.
Therefore, the probability of an accident
previously evaluated is not significantly
increased, if at all. The consequences of an
accident after adopting proposed TSTF–431,
Revision 3, are no different than the
consequences of an accident prior to its
adoption. The addition of a requirement to
assess and manage the risk introduced by this
change will further minimize possible
concerns.
Therefore, this change does not involve a
significant increase in the probability or
consequences of an accident previously
evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a
physical alteration of the plant (no new or
different type of equipment will be installed).
If risk is assessed and managed, allowing a
change to certain required end states when
the TS Completion Times for remaining in
power operation are exceeded; i.e., entry into
hot shutdown rather than cold shutdown to
repair equipment, will not introduce new
failure modes or effects and will not, in the
absence of other unrelated failures, lead to an
accident whose consequences exceed the
consequences of accidents previously
evaluated. The addition of a requirement to
assess and manage the risk introduced by this
change and the commitment by the licensee
to adhere to the guidance in TSTF–IG–07–01,
Implementation Guidance for TSTF–431,
Revision 1, ‘‘Changes in Technical
Specifications end states, BAW–2441–A,’’
will further minimize possible concerns.
Therefore, this change does not create the
possibility of a new or different kind of
accident from an accident previously
evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The proposed change allows, for some
systems, entry into hot shutdown rather than
cold shutdown to repair equipment, if risk is
assessed and managed. The B&WOG’s risk
assessment approach is comprehensive and
follows staff guidance as documented in
[NRC Regulatory Guide (RG) 1.174, Revision
1, ‘‘An Approach For Using Probabilistic Risk
Assessment In Risk-Informed Decisions On
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Plant-Specific Changes To The Licensing
Basis,’’ November 2002, and RG 1.177, ‘‘An
Approach For Plant-Specific, Risk-Informed
Decision Making: Technical Specifications,’’
August 1998]. In addition, the analyses show
that the criteria of the three-tiered approach
for allowing TS changes are met. The risk
impact of the proposed TS changes was
assessed following the three-tiered approach
recommended in RG 1.177. A risk assessment
was performed to justify the proposed TS
changes. The net change to the margin of
safety is insignificant.
Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Joseph A.
Aluise, Associate General Council—
Nuclear, Entergy Services, Inc., 639
Loyola Avenue, New Orleans, Louisiana
70113.
NRC Branch Chief: Michael T.
Markley.
Entergy Operations, Inc., Docket No. 50–
368, Arkansas Nuclear One, Unit No. 2
(ANO–2), Pope County, Arkansas
Date of amendment request:
December 17, 2012.
Description of amendment request:
The licensee has requested NRC review
and approval for adoption of a new fire
protection licensing basis which
complies with the requirements in 10
CFR 50.48(a), 10 CFR 50.48(c), and the
guidance in NRC Regulatory Guide (RG)
1.205, Revision 1, ‘‘Risk-Informed
Performance-Based Fire Protection for
Existing Light-Water Nuclear Power
Plants,’’ December 2009. The license
amendment request follows Nuclear
Energy Institute (NEI) 04–02, Revision 2,
‘‘Guidance for Implementing a RiskInformed, Performance-Based Fire
Protection Program under 10 CFR
50.48(c),’’ April 2008. This submittal
describes the methodology used to
demonstrate compliance with, and
transition to, National Fire Protection
Association (NFPA) 805, and includes
regulatory evaluations, probabilistic risk
assessment, change evaluations,
proposed modifications for noncompliances, and supporting
attachments.
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:
PO 00000
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Criterion 1
The Proposed Change Does Not Involve a
Significant Increase in the Probability or
Consequences of an Accident Previously
Evaluated. Operation of Arkansas Nuclear
One, Unit 2 (ANO–2) in accordance with the
proposed amendment does not result in a
significant increase in the probability or
consequences of accidents previously
evaluated. The proposed amendment does
not affect accident initiators or precursors as
described in the ANO–2 Safety Analysis
Report (SAR), nor does it adversely alter
design assumptions, conditions, or
configurations of the facility, and it does not
adversely impact the ability of structures,
systems, or components (SSCs) to perform
their intended function to mitigate the
consequences of accidents described and
evaluated in the SAR. The proposed changes
do not physically alter safety-related systems
nor affect the way in which safety-related
systems perform their functions as required
by the accident analysis. The SSCs required
to safely shut down the reactor and to
maintain it in a safe shutdown condition will
remain capable of performing their design
functions.
The purpose of this amendment is to
permit ANO–2 to adopt a new risk-informed,
performance-based fire protection licensing
basis that complies with the requirements in
10 CFR 50.48(a) and 10 CFR 50.48(c), as well
as the guidance contained in Regulatory
Guide (RG) 1.205. The NRC considers that
NFPA 805 provides an acceptable
methodology and performance criteria for
licensees to identify fire protection
requirements that are an acceptable
alternative to the 10 CFR Part 50, Appendix
R, fire protection features (69 FR 33536; June
16, 2004).
The purpose of the fire protection program
is to provide assurance, through defense-indepth, that the NRC’s fire protection
objectives are satisfied. These objectives are:
(1) preventing fires from starting; (2) rapidly
detecting and controlling fires and promptly
extinguishing those fires that do occur,
thereby limiting fire damage; (3) providing an
adequate level of fire protection for SSCs
important to safety, so that a fire that is not
promptly extinguished will not prevent
essential plant safety functions from being
performed; and (4) ensuring that fires will
not significantly increase the risk of
radioactive releases to the environment. In
addition, fire protection systems must be
designed such that their failure or
inadvertent operation does not adversely
impact the ability of the SSCs important to
safety to perform their safety-related
functions.
NFPA 805, taken as a whole, provides an
acceptable alternative for satisfying General
Design Criterion 3 (GDC 3) of Appendix A to
10 CFR Part 50, meets the underlying intent
of the NRC’s existing fire protection
regulations and guidance, and achieves
defense-in-depth along with the goals,
performance objectives, and performance
criteria specified in NFPA 805, Chapter 1. In
addition, if there are any increases in core
damage frequency (CDF) or risk as a result of
the transition to NFPA 805, the increase will
be small, bounded by the delta risk
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requirements of NFPA 805, and consistent
with the intent of the Commission’s Safety
Goal Policy.
Engineering analyses, which may include
engineering evaluations, probabilistic risk
assessments, and fire modeling calculations,
have been performed to demonstrate that the
performance-based requirements of NFPA
805 have been met. The SAR documents the
analyses of design basis accidents (DBAs) at
ANO–2. All accident analysis acceptance
criteria will continue to be met with the
proposed amendment. The proposed changes
will not affect the source term, containment
isolation, or radiological release assumptions
used in evaluating the radiological
consequences of any accident previously
evaluated. The proposed changes will not
alter any assumptions or change any
mitigation actions for the radiological
consequence evaluations in the ANO–2 SAR.
In addition, the applicable radiological dose
acceptance criteria will continue to be met.
Based on the above, the implementation of
this amendment to transition the Fire
Protection Plan (FPP) at ANO–2 to one based
on NFPA 805, in accordance with 10 CFR
50.48(c), does not result in a significant
increase in the probability of any accident
previously evaluated. In addition, all
equipment required to mitigate an accident
remains capable of performing the assumed
function. Therefore, the consequences of any
accident previously evaluated are not
significantly increased with the
implementation of this amendment.
Criterion 2
The Proposed Change Does Not Create the
Possibility of a New or Different Kind of
Accident from Any Accident Previously
Evaluated
Operation of ANO–2 in accordance with
the proposed amendment does not create the
possibility of a new or different kind of
accident from any accident previously
evaluated. Previously analyzed accidents
with potential offsite dose consequences
were included in the evaluation of the
transition to NFPA 805. The proposed
amendment does not impact these accident
analyses. The proposed change does not alter
the requirements or functions for systems
required during accident conditions as
assumed in the licensing basis analyses and/
or DBA [design-basis accident] radiological
consequences evaluations.
Implementation of the new risk-informed,
performance-based fire protection licensing
basis, which complies with the requirements
in 10 CFR 50.48(a) and 10 CFR 50.48(c), as
well as the guidance contained in RG 1.205,
will not result in new or different kinds of
accidents. The NRC considers that NFPA 805
provides an acceptable methodology and
performance criteria for licensees to identify
fire protection systems and features that are
an acceptable alternative to the 10 CFR 50,
Appendix R fire protection features (69 FR
33536, June 16, 2004). No new modes of
operation are introduced by the proposed
amendment, nor will it create any failure
mode not bounded by previously evaluated
accidents. Further, the impacts of the
proposed change are not directly assumed in
any safety analysis to initiate an accident
sequence.
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The requirements in NFPA 805 address
only fire protection and the impacts of fire
effects on the plant have been evaluated. The
proposed fire protection program changes do
not involve new failure mechanisms or
malfunctions that could initiate a new or
different kind of accident beyond those
already analyzed in the SAR. Based on this,
as well as the discussion above, the
implementation of this amendment to
transition the FPP at ANO–2 to one based on
NFPA 805, in accordance with 10 CFR
50.48(c), does not create the possibility of a
new or different kind of accident from any
accident previously evaluated.
Criterion 3
The Proposed Change Does Not Involve a
Significant Reduction in a Margin of safety.
Operation of ANO–2 in accordance with
the proposed amendment does not involve a
significant reduction in a margin of safety.
The transition to a new risk-informed,
performance-based fire protection licensing
basis that complies with the requirements in
10 CFR 50.48(a) and 10 CFR 50.48(c) does not
alter the manner in which safety limits,
limiting safety system settings, or limiting
conditions for operation are determined. The
safety analysis acceptance criteria are not
affected by this change. The proposed
amendment does not adversely affect existing
plant safety margins or the reliability of
equipment assumed in the SAR to mitigate
accidents. The proposed change does not
adversely impact systems that respond to
safely shut down the plant and maintain the
plant in a safe shutdown condition. In
addition, the proposed amendment will not
result in plant operation in a configuration
outside the design basis for an unacceptable
period of time without implementation of
appropriate compensatory measures.
The risk evaluations for plant changes, in
part as they relate to the potential for
reducing a safety margin, were measured
quantitatively for acceptability using the
delta risk (i.e., DCDF and DLERF) criteria
from Section 5.3.5, ‘‘Acceptance Criteria,’’ of
NEI 04–02, as well as the guidance contained
in RG 1.205. Engineering analyses, which
may include engineering evaluations,
probabilistic safety assessments, and fire
modeling calculations, have been performed
to demonstrate that the performance-based
methods of NFPA 805 do not result in a
significant reduction in the margin of safety.
As such, the proposed changes are evaluated
to ensure that risk and safety margins are
kept within acceptable limits. Based on the
above, the implementation of this
amendment to transition the FPP at ANO–2
to one based on NFPA 805, in accordance
with 10 CFR 50.48(c), will not significantly
reduce 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: Joseph A.
Aluise, Associate General Council—
PO 00000
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Fmt 4703
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Nuclear, Entergy Services, Inc., 639
Loyola Avenue, New Orleans, Louisiana
70113.
NRC Branch Chief: Michael T.
Markley.
Entergy Operations, Inc., Docket No. 50–
368, Arkansas Nuclear One, Unit No. 2,
Pope County, Arkansas
Date of amendment request: March
26, 2013.
Description of amendment request:
The amendment would incorporate the
NRC-approved Technical Specifications
Task Force (TSTF) change traveler
TSTF–422, Revision 2, ‘‘Change in
Technical Specifications End States (CE
NPSD–1186).’’ The proposed
amendment would modify Technical
Specifications (TS) to risk-inform
requirements regarding selected
Required Action End States.
The NRC issued a ‘‘Notice of
Availability (NOA) of the Models For
Plant-Specific Adoption of Technical
Specifications Task Force (TSTF)
Traveler TSTF–422, Revision 2, ‘Change
In Technical Specifications End States
(CE NPSD–1186),’ For Combustion
Engineering (CE) Pressurized Water
Reactor (PWR) Plants Using the
Consolidated Line Item Improvement
Process (CLIIP),’’ in the Federal Register
on April 7, 2011 (76 FR 19510), which
included the no significant hazards
consideration, safety evaluation, and
required commitments for the proposed
changes as part of the consolidated line
item improvement process (CLIIP).
In its application dated March 26,
2013, the licensee has concluded that
the technical basis presented in the
TSTF proposal and the safety evaluation
are applicable to Arkansas Nuclear One,
Unit 2, and the proposed amendment is
consistent with the Standard Technical
Specifications (STS) changes described
in TSTF–422, Revision 2, but with
certain variations and/or deviations
from TSTF–422, Revision 2.
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 allows a change to
certain required end states when the
Technical Specification (TS) Completion
Times (CTs) for remaining in power
operation are exceeded. Most of the
requested TS changes are to permit an end
state of hot shutdown (Mode 4) rather than
an end state of cold shutdown (Mode 5)
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contained in the current TS. The request was
limited to: (1) those end states where entry
into the shutdown mode is for a short
interval; (2) entry is initiated by inoperability
of a single train of equipment or a restriction
on a plant operational parameter, unless
otherwise stated in the applicable TS; and (3)
the primary purpose is to correct the
initiating condition and return to power
operation as soon as is practical. Risk
insights from both the qualitative and
quantitative risk assessments were used in
specific TS assessments. Such assessments
are documented in Section 5.5 of CE NPSD–
1186, Rev 0, ‘‘Technical Justification for the
Risk-Informed Modification to Selected
Required Action End States for CEOG
[Combustion Engineering Owners Group]
Member PWRs.’’ The assessments provide an
integrated discussion of deterministic and
probabilistic issues, focusing on specific TSs,
which are used to support the proposed TS
end state and associated restrictions.
Therefore, the probability of an accident
previously evaluated is not significantly
increased, if at all. The consequences of an
accident after adopting proposed TSTF–422
are no different than the consequences of an
accident prior to adopting TSTF–422.
Therefore, the consequences of an accident
previously evaluated are not significantly
affected by this change. The addition of a
requirement to assess and manage the risk
introduced by this change will further
minimize possible concerns.
Therefore, this change does not involve a
significant increase in the probability or
consequences of an accident previously
evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change does not involve a
physical alteration of the plant (no new or
different type of equipment will be installed).
Allowing a change to certain required end
states when the TS CTs for remaining in
power operation are exceeded, i.e., entry into
hot shutdown rather than cold shutdown to
repair equipment, if risk is assessed and
managed, will not introduce new failure
modes or effects and will not, in the absence
of other unrelated failures, lead to an
accident whose consequences exceed the
consequences of accidents previously
evaluated. The addition of a requirement to
assess and manage the risk introduced by this
change and the commitment by the licensee
to adhere to the guidance in WCAP–16364–
NP, Revision 2, ‘‘Implementation Guidance
for Risk Informed Modification to Selected
Required Action End States at Combustion
Engineering NSSS Plants (TSTF–422),’’ will
further minimize possible concerns.
Therefore, this change does not create the
possibility of a new or different kind of
accident from an accident previously
evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The proposed change allows, for some
systems, entry into hot shutdown rather than
cold shutdown to repair equipment, if risk is
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assessed and managed. The CEOG’s risk
assessment approach is comprehensive and
follows NRC staff guidance as documented in
[NRC Regulatory Guide (RG) 1.174, ‘‘An
Approach for Using Probabilistic Risk
Assessment in Risk-Informed Decision
Making on Plant Specific Changes to the
Licensing Basis,’’ August 1998, and RG 1.177,
‘‘An Approach for Pant Specific RiskInformed Decision Making: Technical
Specifications,’’ August 1998.]. In addition,
the analyses show that the criteria of the
three-tiered approach for allowing TS
changes are met. The risk impact of the
proposed TS changes was assessed following
the three-tiered approach recommended in
RG 1.177. A risk assessment was performed
to justify the proposed TS changes. The net
change to the margin of safety is
insignificant.
Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Joseph A.
Aluise, Associate General Council—
Nuclear, Entergy Services, Inc., 639
Loyola Avenue, New Orleans, Louisiana
70113.
NRC Branch Chief: Michael T.
Markley.
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: May 21,
2013.
Description of amendment request:
The proposed amendment would
modify the Technical Specifications
(TSs) moderator temperature coefficient
(MTC) surveillance requirements
associated with the implementation of
Topical Report WCAP–16011–P–A,
‘‘Startup Test Activity Reduction
(STAR) Program,’’ which describes the
methods to be used for the
implementation of reduction in the
startup testing requirements. The
changes are consistent with the Nuclear
Regulatory Commission (NRC)-approved
Industry/Technical Specification Task
Force (TSTF) Standard Technical
Specifications change TSTF–486,
Revision 2 as included in NUREG–1432,
Revision 4.0, Standard Technical
Specifications—Combustion
Engineering (CE) Plants.
The NRC staff published a notice of
opportunity for comment in the Federal
Register on July 27, 2007 (72 FR 41360),
on possible amendments adopting
TSTF–486 using the NRC’s consolidated
line-item improvement process for
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44173
amending licensees’ TSs, which
included a model safety evaluation (SE)
and model no significant hazards
consideration (NSHC) determination.
The NRC staff subsequently issued a
notice of availability of the models for
referencing in license amendment
applications in the Federal Register on
September 6, 2007 (72 FR 51259), which
included the resolution of public
comments on the model SE and model
NSHC determination. The licensee
affirmed in its application dated May
21, 2013, that the proposed changes to
the TSs satisfy the intent of TSTF–486.
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 NSHC, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
The proposed changes for St. Lucie Units
1 and 2 revise the MTC Technical
Specification 4.1.1.4.1 and 4.1.1.4.2 for each
Unit, to implement the requirements of the
topical report WCAP–16011–P–A, STAR
Program.
The MTC is not an initiator to any accident
previously evaluated. Therefore, there is no
significant increase in the probability of any
accident previously evaluated. The MTC is
an input to the accident analyses used to
predict plant behavior in the event of an
accident. The MTC limits specified in the
Technical Specifications/COLR [core
operating limit report] remain unchanged.
WCAP–16011–P–A demonstrated, and the
NRC concurred, that the modified MTC
verification is adequate to ensure that MTC
stays within the limits. The consequences of
an accident after adopting TSTF–486 are no
different than the consequences of an
accident prior to adoption. Likewise, the
deviations from the implementation of
TSTF–486 requirements being adopted in
this license amendment do not have any
effect on the probability of occurrence or
consequences of accidents previously
evaluated.
Therefore, the proposed changes do not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
No new or different accidents will result
from implementation of the proposed
changes. The changes do not involve a
physical alteration of the plant (i.e., no new
or different type of equipment will be
installed) or a change in the methods
governing normal plant operation. In
addition, the changes do not impose any new
or different operating requirements or
eliminate any existing requirements. The
changes do not alter limits and assumptions
made in the safety analysis. The proposed
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changes are consistent with the safety
analysis assumptions and current plant
operating practice. Therefore, the proposed
change does not create the possibility of a
new or different kind of accident from any
accident previously evaluated.
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?
TSTF–486 provides the means and
requirements for CE-designed plants to
implement the previously approved WCAP–
16011–P–A for MTC verification at startup.
MTC is a parameter controlled in the
licensee’s TS/COLR, including surveillance
requirements. As stated previously, WCAP–
16011–P–A describes methods to reduce the
requirements for startup testing. The
proposed changes to the TS, supported by
TSTF–486, have been reviewed and found to
be consistent with WCAP–16011–P–A. The
changes in the license amendment which
deviate from TSTF–486 requirements are
justified to be acceptable and do not affect
the margin of safety. The MTC limits are
unaffected and an acceptable method will be
used to verify the MTC to be within its limit.
Therefore, the proposed changes do not
involve a significant reduction in a margin of
safety.
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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 Branch Chief: Jessie F.
Quichocho.
Florida Power Corporation, et al.,
Docket No. 50–302, Crystal River Unit 3
Nuclear Generating Plant, Citrus
County, Florida
Date of amendment request: April 25,
2013.
Description of amendment request:
The proposed license amendment
request would revise certain
requirements from Section 5,
‘‘Administrative Controls,’’ of the
Crystal River Unit 3 (CR–3) Improved
Technical Specifications (ITSs). The
revisions would include the following
sections: 5.1 ‘‘Responsibility;’’ 5.2
‘‘Organization;’’ 5.6 ‘‘Procedures,
Programs and Manuals;’’ 5.7 ‘‘Reporting
Requirements;’’ and 5.8 ‘‘High Radiation
Area,’’ which are no longer applicable,
as CR–3 is in a permanently defueled
condition.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
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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?
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?
No. This changes 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.
3. Does the proposed change involve a
significant reduction in a margin of safety?
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?
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?
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.
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3. Does the proposed change involve a
significant reduction in a margin of safety?
No. The changes proposed here for control
room staffing do not directly involve any
limits or parameters and therefore cannot
affect ant 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?
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?
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?
No. This change does not directly involve
any physical limits or parameters and
therefore cannot affect any margin of safety.
D. ITS Section 5.1.2.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?
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?
No. This change in the description of
functional responsibility of organizational
positions cannot create a new or different
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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?
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?
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?
No. This change does not introduce any
changes to the function of any plant
structures, systems, or components there it
cannot create a new or different kind of
accident.
3. Does the proposed change involve a
significant reduction in a margin of safety?
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?
No. This change, in conjunction with new
paragraph 5.2.2.e, 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?
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?
No. This change does not directly involve
any limits or parameters and therefore cannot
affect any margin of safety.
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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?
No. This change continues to ensure that
the minimum shift compliment of qualified
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?
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?
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?
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?
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?
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
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procedures to be onsite when fuel is in the
reactor.
This paragraph is deleted, since CR–3 is no
longer authorized to have fuel in the reactor.
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
No. This administrative change 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?
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?
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.d (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?
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?
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?
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.e (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?
No. Certified Fuel Handlers are specifically
trained and qualified to safely handle
irradiated fuel. Applying these qualifications
to the supervision of fuel movement ensures
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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?
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?
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?
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?
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?
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?
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
accident from any accident previously
evaluated?
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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?
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?
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?
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?
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?
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?
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?
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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?
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?
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?
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?
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?
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?
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?
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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?
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?
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 5.6.2.10: 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?
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?
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?
No. This change does not directly involve
any limits or parameters and therefore cannot
affect any margin of safety.
T. 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?
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
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accident from any accident previously
evaluated?
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?
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.
U. 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?
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?
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?
No. This change does not directly involve
any limits or parameters and therefore cannot
affect any margins of safety.
V. 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?
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
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accident from any accident previously
evaluated?
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?
No. Since CR–3 can never load a core into
the reactor again, eliminating this control
program cannot affect any margin of safety.
W. 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?
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?
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?
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.
X. 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?
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?
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?
No. This change does not directly involve
any limits or parameters and therefore cannot
affect any margin of safety.
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Y. ITS Section 5.7.2: Special Reports
This section is being eliminated.
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
No. Eliminating reporting requirements for
programs that are no longer required or
conditions that cannot exist 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?
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?
No. Eliminating reporting requirements
that are no longer required cannot affect any
margin of safety.
Z. ITS Section 5.8.2: High Radiation Area
Controls
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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?
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?
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?
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.
VerDate Mar<15>2010
15:40 Jul 22, 2013
Jkt 229001
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 https://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
pdr.resource@nrc.gov.
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Northern States Power Company—
Minnesota, Docket Nos. 50–282 and 50–
306, Prairie Island Nuclear Generating
Plant (PINGP), Units 1 and 2, Goodhue
County, Minnesota
Date of application for amendments:
July 25, 2012.
Brief description of amendments: The
amendments revise Technical
Specifications (TSs) 3.4.19—‘‘Steam
Generator (SG) Tube Integrity,’’ 5.5.8—
‘‘Steam Generator (SG) Program,’’ and
5.6.7—‘‘Steam Generator Tube
Inspection Report’’ to apply the
appropriate program attributes to the
Unit 2 replacement steam generators
that are planned for installation in fall
2013. The amendments also revise the
PINGP Units 1 and 2 TSs to adopt the
program improvements in Technical
Specifications Task Force Traveler
(TSTF) 510, Revision 2, ‘‘Revision to
Steam Generator Program Inspection
Frequencies and Tube Sample
Selection.’’
Date of issuance: July 2, 2013.
Effective date: As of the date of
issuance and shall be implemented
within 60 days after reactor startup
following Unit 2 steam generator
replacements.
Amendment Nos.: 208 and 195.
Renewed Facility Operating License
Nos. DPR–42 and DPR–60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal
Register: September 14, 2012 (77 FR
56881).
The Commission’s related evaluation
of the amendments is contained in a
Safety Evaluation dated July 2, 2013.
No significant hazards consideration
comments received: No.
Southern Nuclear Operating Company.
Docket Nos. 52–025 and 52–026, Vogtle
Electric Generating Plant (VEGP) Units
3 and 4, Burke County, Georgia
Date of application for amendments:
March 20, 2013.
Brief description of amendment: The
amendment authorizes a departure from
the Vogtle Electric Generating Plant
Units 3 and 4 plant-specific Design
Control Document (DCD) material
incorporated into the Updated Final
Safety Analysis Report (UFSAR) by
revising the structural analysis
requirements to provide alternative
requirements for development of headed
reinforcement bars (T-heads) within the
nuclear island structures above the
basemat elevation.
Date of issuance: May 22, 2013.
Effective date: As of the date of
issuance and shall be implemented
within 30 days of issuance.
Amendment No.: Unit 3–9 and Unit
4–9.
E:\FR\FM\23JYN1.SGM
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Federal Register / Vol. 78, No. 141 / Tuesday, July 23, 2013 / Notices
Facility Combined Licenses No. NPF–
91 and NPF–92: Amendment revised the
Facility Combined Licenses.
Date of initial notice in Federal
Register: April 16, 2013 (78 FR 22573).
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated May 22, 2013.
No significant hazards consideration
comments received: No.
Dated at Rockville, Maryland, this 15th day
of July 2013.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor
Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. 2013–17370 Filed 7–22–13; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[NRC–2013–0001]
Sunshine Act Meetings
AGENCY HOLDING THE MEETINGS: Nuclear
Regulatory Commission.
DATES: Weeks of July 22, 29, August 5,
12, 19, 26, 2013.
PLACE: Commissioners’ Conference
Room, 11555 Rockville Pike, Rockville,
Maryland.
STATUS: Public and Closed.
Week of July 22, 2013
There are no meetings scheduled for
the week of July 22, 2013.
Ex. 1 & 9) (Contact: Karen
Henderson, 301–415–0202)
*
*
*
*
*
*The schedule for Commission
meetings is subject to change on short
notice. To verify the status of meetings,
call (recording)—301–415–1292.
Contact person for more information:
Rochelle Bavol, 301–415–1651.
*
*
*
*
*
The NRC Commission Meeting
Schedule can be found on the Internet
at: https://www.nrc.gov/public-involve/
public-meetings/schedule.html.
*
*
*
*
*
The NRC provides reasonable
accommodation to individuals with
disabilities where appropriate. If you
need a reasonable accommodation to
participate in these public meetings, or
need this meeting notice or the
transcript or other information from the
public meetings in another format (e.g.
braille, large print), please notify
Kimberly Meyer, NRC Disability
Program Manager, at 301–287–0727, or
by email at kimberly.meyerchambers@nrc.gov. Determinations on
requests for reasonable accommodation
will be made on a case-by-case basis.
*
*
*
*
*
This notice is distributed
electronically to subscribers. If you no
longer wish to receive it, or would like
to be added to the distribution, please
contact the Office of the Secretary,
Washington, DC 20555 (301–415–1969),
or send an email to
darlene.wright@nrc.gov.
Week of July 29, 2013—Tentative
There are no meetings scheduled for
the week of July 29, 2013.
Dated: July 18, 2013.
Rochelle C. Bavol,
Policy Coordinator, Office of the Secretary.
[FR Doc. 2013–17756 Filed 7–19–13; 4:15 pm]
BILLING CODE 7590–01–P
Week of August 5, 2013—Tentative
There are no meetings scheduled for
the week of August 5, 2013.
Week of August 12, 2013—Tentative
SECURITIES AND EXCHANGE
COMMISSION
There are no meetings scheduled for
the week of August 12, 2013.
Submission for OMB Review;
Comment Request
Week of August 19, 2013—Tentative
Upon Written Request Copies Available
From: Securities and Exchange
Commission, Office of Investor
Education and Advocacy,
Washington, DC 20549–0213.
There are no meetings scheduled for
the week of August 19, 2013.
Week of August 26, 2013—Tentative
ehiers on DSK2VPTVN1PROD with NOTICES
Tuesday, August 27, 2013
9:00 a.m. Briefing on NRC’s
Construction Activities (Public
Meeting); (Contact: Michelle Hayes,
301–415–8375).
This meeting will be webcast live at
the Web address—https://www.nrc.gov.
3:00 p.m. Briefing on NRC
International Activities (Closed—
VerDate Mar<15>2010
15:40 Jul 22, 2013
Jkt 229001
Extension:
Form S–8; OMB Control No. 3235–0066,
SEC File No. 270–66.
Notice is hereby given that, pursuant
to the Paperwork Reduction Act of 1995
(44 U.S.C. 3501 et seq.), the Securities
and Exchange Commission
(‘‘Commission’’) has submitted to the
Office of Management and Budget this
request for extension of the previously
PO 00000
Frm 00088
Fmt 4703
Sfmt 4703
44179
approved collection of information
discussed below.
Form S–8 (17 CFR 239.16b) under the
Securities Act of 1933 (15 U.S.C. 77a et
seq.) is the primary registration
statement used by eligible registrants to
register securities to be issuers in
connection with an employee benefit
plan. Form S–8 provides verification of
compliance with securities law
requirements and assures the public
availability and dissemination of such
information. The likely respondents will
be companies. The information must be
filed with the Commission on occasion.
Form S–8 is a public document. All
information provided is mandatory. We
estimate that Form S–8 takes
approximately 24 hours per response to
prepare and is filed by approximately
2,200 respondents. In addition, we
estimate that 50% of the preparation
time (12 hours) is completed in-house
by the filer for a total annual reporting
burden of 26,400 hours (12 hours per
response x 2,200 responses)
An agency may conduct or sponsor,
and a person is not required to respond
to, a collection of information unless it
displays a currently valid control
number.
The public may view the background
documentation for this information
collection at the following Web site,
www.reginfo.gov. Comments should be
directed to: (i) Desk Officer for the
Securities and Exchange Commission,
Office of Information and Regulatory
Affairs, Office of Management and
Budget, Room 10102, New Executive
Office Building, Washington, DC 20503,
or by sending an email to:
Shagufta_Ahmed@omb.eop.gov; and (ii)
Thomas Bayer, Director/Chief
Information Officer, Securities and
Exchange Commission, c/o Remi PavlikSimon, 100 F Street NE., Washington,
DC 20549, or send an email to:
PRA_Mailbox@sec.gov. Comments must
be submitted to OMB within 30 days of
this notice.
Dated: July 17, 2013.
Kevin M. O’Neill,
Deputy Secretary.
[FR Doc. 2013–17597 Filed 7–22–13; 8:45 am]
BILLING CODE 8011–01–P
SECURITIES AND EXCHANGE
COMMISSION
Submission for OMB Review;
Comment Request
Upon Written Request Copies Available
From: Securities and Exchange
Commission, Office of Investor
Education and Advocacy,
Washington, DC 20549–0213.
E:\FR\FM\23JYN1.SGM
23JYN1
Agencies
[Federal Register Volume 78, Number 141 (Tuesday, July 23, 2013)]
[Notices]
[Pages 44167-44179]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2013-17370]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[NRC-2013-0158]
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 (the
Commission or NRC) is publishing this regular biweekly notice. The Act
requires that the Commission publish notice of any amendments issued or
proposed to be issued and grants the Commission the authority to issue
and make immediately effective any amendment to an operating license or
combined license, as applicable, upon a determination by the Commission
that such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from June 27, 2013 to July 10, 2013. The last
biweekly notice was published on July 9, 2013 (78 FR 41118).
ADDRESSES: You may submit comment 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 https://www.regulations.gov and search for Docket ID NRC-2103-0158. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: Carol.Gallagher@nrc.gov. For technical questions, contact
the individual(s) listed in the FOR FURTHER INFORMATION CONTACT section
of this document.
Mail comments to: Cindy Bladey, Chief, Rules,
Announcements, and Directives Branch (RADB), Office of Administration,
Mail Stop: 3WFN-06A-44MP, 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-2013-0158 when contacting the NRC
about the availability of information regarding this document. You may
access information related to this document, which the NRC possesses
and is publicly available, by the following methods:
Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2013-0158.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly available documents online in the NRC
Library at https://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 pdr.resource@nrc.gov. 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-2013-0158 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.
[[Page 44168]]
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 https://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 Title 10 of the Code of Federal
Regulations (10CFR), Section 50.92, 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 https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or a presiding
officer designated by the Commission or by the Chief Administrative
Judge of the Atomic Safety and Licensing Board Panel, will rule on the
request and/or petition; and the Secretary or the Chief Administrative
Judge of the Atomic Safety and Licensing Board will issue a notice of a
hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment.
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
[[Page 44169]]
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 hearing.docket@nrc.gov, 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 https://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 https://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 https://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 https://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 https://www.nrc.gov/site-help/e-submittals.html, by email at
MSHD.Resource@nrc.gov, 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
https://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
[[Page 44170]]
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 https://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 pdr.resource@nrc.gov.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New
York
Date of amendment request: May 23, 2013.
Description of amendment request: The proposed change would modify
Technical Specifications (TS) to risk-inform requirements regarding
selected Required Action End States. Specifically, the proposed change
would permit an end state of Mode 4 rather than an end state of Mode 5
contained in the current TS. The proposed changes are consistent with
NRC-approved Technical Specification Task Force (TSTF) Technical Change
Traveler 432-A Revision 1, ``Change in Technical Specifications End
States WCAP-16294.'' This traveler revised the Improved Standard
Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change modifies the end state (e.g., mode or other
specified condition) which the Required Actions specify must be
entered if compliance with the Limiting Conditions for Operation
(LCO) is not restored. The requested Technical Specifications (TS)
permit an end state of Mode 4 rather than an end state of Mode 5
contained in the current TS. In some cases, other Conditions and
Required Actions are revised to implement the proposed change.
Required Actions are not an initiator of any accident previously
evaluated. Therefore, the proposed change does not affect the
probability of any accident previously evaluated. The affected
systems continue to be required to be operable by the TS and the
Completion Times specified in the TS to restore equipment to
operable status or take other remedial Actions remain unchanged.
WCAP-16294-NP-A, Rev. 1, ``Risk-Informed Evaluation of Changes to
Tech Spec Required Action End states for Westinghouse NSSS PWRs,''
demonstrates that the proposed change does not significantly
increase the consequences of any accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change modifies the end state (e.g., mode or other
specified condition) which the Required Actions specify must be
entered if compliance with the LCO is not restored. In some cases,
other Conditions and Required Actions are revised to implement the
proposed change. The change does not involve a physical alteration
of the plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the change does not impose any new
requirements. The change does not alter assumptions made in the
safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change modifies the end state (e.g., mode or other
specified condition) which the Required Actions specify must be
entered if compliance with the LCO is not restored. In some cases,
other Conditions and Required Actions are revised to implement the
proposed change. Remaining within the Applicability of the LCO is
acceptable because WCAP-16294-NP-A demonstrates that the plant risk
in MODE 4 is similar to or lower than MODE 5. As a result, no margin
of safety is significantly affected.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Robert Beall, Acting.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: March 26, 2013.
Description of amendment request: The amendment request would
incorporate the NRC-approved Technical Specifications Task Force (TSTF)
change traveler TSTF-431, Revision 3, ``Change in Technical
Specifications End States (BAW-2441),'' and modify the Technical
Specification (TS) requirements for end states associated with the
implementation of the approved B&W Owners Group (B&WOG) Topical Report
BAW-2441-A, Revision 2, ``Risk-Informed Justification for LCO End-State
Changes,'' January 2004, as well as Required Actions revised by a
specific Note in TSTF-431, Revision 3. The TS Actions End States
modifications would permit, for some systems, entry into a hot shutdown
(Mode 4) end state rather than a cold shutdown (Mode 5) end state that
is the current TS requirement.
The NRC issued a ``Notice of Availability of the Models for Plant-
Specific Adoption of Technical Specifications Task Force (TSTF)
Traveler TSTF-431, Revision 3, `Change in Technical Specifications End
States (BAW-2441),' '' in the Federal Register on December 6, 2010 (75
FR 75705-75706), which included the no significant hazards
consideration, safety evaluation, and required commitments for the
proposed changes as part of the consolidated line item improvement
process (CLIIP).
In its application dated March 26, 2013, the licensee has concluded
that the technical basis presented in the TSTF proposal and the safety
evaluation are applicable to Arkansas Nuclear One, Unit 1, and the
proposed amendment is consistent with the Standard Technical
Specifications (STS) changes described in TSTF-431, Revision 3, but
with certain variations and/or deviations from TSTF-431, Revision 3.
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 allows a change to certain required end
states when the Technical Specification (TS) Completion Times (CTs)
for remaining in power operation are exceeded. Most of the requested
TS changes are to permit an end state of hot shutdown (Mode 4)
rather than an end state of cold shutdown (Mode 5) contained in the
current TS. The request was limited to: 1) those end states where
entry into the shutdown mode is for a short
[[Page 44171]]
interval, 2) entry is initiated by inoperability of a single train
of equipment or a restriction on a plant operational parameter,
unless otherwise stated in the applicable TS, and 3) the primary
purpose is to correct the initiating condition and return to power
operation as soon as is practical. Risk insights from both the
qualitative and quantitative risk assessments were used in specific
TS assessments. Such assessments are documented in Sections 4 and 5
of BAW-2441-A, Revision 2, ``Risk Informed Justification for LCO
end-state Changes,'' for B&W Plants. The assessments provide an
integrated discussion of deterministic and probabilistic issues,
focusing on specific TSs, which are used to support the proposed TS
end state and associated restrictions. The staff finds that the risk
insights support the conclusions of the specific TS assessments.
Therefore, the probability of an accident previously evaluated is
not significantly increased, if at all. The consequences of an
accident after adopting proposed TSTF-431, Revision 3, are no
different than the consequences of an accident prior to its
adoption. The addition of a requirement to assess and manage the
risk introduced by this change will further minimize possible
concerns.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
If risk is assessed and managed, allowing a change to certain
required end states when the TS Completion Times for remaining in
power operation are exceeded; i.e., entry into hot shutdown rather
than cold shutdown to repair equipment, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change and the commitment by the licensee to adhere to the guidance
in TSTF-IG-07-01, Implementation Guidance for TSTF-431, Revision 1,
``Changes in Technical Specifications end states, BAW-2441-A,'' will
further minimize possible concerns.
Therefore, this change does not create the possibility of a new
or different kind of accident from an accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change allows, for some systems, entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed. The B&WOG's risk assessment approach is
comprehensive and follows staff guidance as documented in [NRC
Regulatory Guide (RG) 1.174, Revision 1, ``An Approach For Using
Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-
Specific Changes To The Licensing Basis,'' November 2002, and RG
1.177, ``An Approach For Plant-Specific, Risk-Informed Decision
Making: Technical Specifications,'' August 1998]. In addition, the
analyses show that the criteria of the three-tiered approach for
allowing TS changes are met. The risk impact of the proposed TS
changes was assessed following the three-tiered approach recommended
in RG 1.177. A risk assessment was performed to justify the proposed
TS changes. The net change to the margin of safety is insignificant.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2 (ANO-2), Pope County, Arkansas
Date of amendment request: December 17, 2012.
Description of amendment request: The licensee has requested NRC
review and approval for adoption of a new fire protection licensing
basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR
50.48(c), and the guidance in NRC Regulatory Guide (RG) 1.205, Revision
1, ``Risk-Informed Performance-Based Fire Protection for Existing
Light-Water Nuclear Power Plants,'' December 2009. The license
amendment request follows Nuclear Energy Institute (NEI) 04-02,
Revision 2, ``Guidance for Implementing a Risk-Informed, Performance-
Based Fire Protection Program under 10 CFR 50.48(c),'' April 2008. This
submittal describes the methodology used to demonstrate compliance
with, and transition to, National Fire Protection Association (NFPA)
805, and includes regulatory evaluations, probabilistic risk
assessment, change evaluations, proposed modifications for non-
compliances, and supporting attachments.
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:
Criterion 1
The Proposed Change Does Not Involve a Significant Increase in
the Probability or Consequences of an Accident Previously Evaluated.
Operation of Arkansas Nuclear One, Unit 2 (ANO-2) in accordance with
the proposed amendment does not result in a significant increase in
the probability or consequences of accidents previously evaluated.
The proposed amendment does not affect accident initiators or
precursors as described in the ANO-2 Safety Analysis Report (SAR),
nor does it adversely alter design assumptions, conditions, or
configurations of the facility, and it does not adversely impact the
ability of structures, systems, or components (SSCs) to perform
their intended function to mitigate the consequences of accidents
described and evaluated in the SAR. The proposed changes do not
physically alter safety-related systems nor affect the way in which
safety-related systems perform their functions as required by the
accident analysis. The SSCs required to safely shut down the reactor
and to maintain it in a safe shutdown condition will remain capable
of performing their design functions.
The purpose of this amendment is to permit ANO-2 to adopt a new
risk-informed, performance-based fire protection licensing basis
that complies with the requirements in 10 CFR 50.48(a) and 10 CFR
50.48(c), as well as the guidance contained in Regulatory Guide (RG)
1.205. The NRC considers that NFPA 805 provides an acceptable
methodology and performance criteria for licensees to identify fire
protection requirements that are an acceptable alternative to the 10
CFR Part 50, Appendix R, fire protection features (69 FR 33536; June
16, 2004).
The purpose of the fire protection program is to provide
assurance, through defense-in-depth, that the NRC's fire protection
objectives are satisfied. These objectives are: (1) preventing fires
from starting; (2) rapidly detecting and controlling fires and
promptly extinguishing those fires that do occur, thereby limiting
fire damage; (3) providing an adequate level of fire protection for
SSCs important to safety, so that a fire that is not promptly
extinguished will not prevent essential plant safety functions from
being performed; and (4) ensuring that fires will not significantly
increase the risk of radioactive releases to the environment. In
addition, fire protection systems must be designed such that their
failure or inadvertent operation does not adversely impact the
ability of the SSCs important to safety to perform their safety-
related functions.
NFPA 805, taken as a whole, provides an acceptable alternative
for satisfying General Design Criterion 3 (GDC 3) of Appendix A to
10 CFR Part 50, meets the underlying intent of the NRC's existing
fire protection regulations and guidance, and achieves defense-in-
depth along with the goals, performance objectives, and performance
criteria specified in NFPA 805, Chapter 1. In addition, if there are
any increases in core damage frequency (CDF) or risk as a result of
the transition to NFPA 805, the increase will be small, bounded by
the delta risk
[[Page 44172]]
requirements of NFPA 805, and consistent with the intent of the
Commission's Safety Goal Policy.
Engineering analyses, which may include engineering evaluations,
probabilistic risk assessments, and fire modeling calculations, have
been performed to demonstrate that the performance-based
requirements of NFPA 805 have been met. The SAR documents the
analyses of design basis accidents (DBAs) at ANO-2. All accident
analysis acceptance criteria will continue to be met with the
proposed amendment. The proposed changes will not affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of any accident
previously evaluated. The proposed changes will not alter any
assumptions or change any mitigation actions for the radiological
consequence evaluations in the ANO-2 SAR. In addition, the
applicable radiological dose acceptance criteria will continue to be
met.
Based on the above, the implementation of this amendment to
transition the Fire Protection Plan (FPP) at ANO-2 to one based on
NFPA 805, in accordance with 10 CFR 50.48(c), does not result in a
significant increase in the probability of any accident previously
evaluated. In addition, all equipment required to mitigate an
accident remains capable of performing the assumed function.
Therefore, the consequences of any accident previously evaluated are
not significantly increased with the implementation of this
amendment.
Criterion 2
The Proposed Change Does Not Create the Possibility of a New or
Different Kind of Accident from Any Accident Previously Evaluated
Operation of ANO-2 in accordance with the proposed amendment
does not create the possibility of a new or different kind of
accident from any accident previously evaluated. Previously analyzed
accidents with potential offsite dose consequences were included in
the evaluation of the transition to NFPA 805. The proposed amendment
does not impact these accident analyses. The proposed change does
not alter the requirements or functions for systems required during
accident conditions as assumed in the licensing basis analyses and/
or DBA [design-basis accident] radiological consequences
evaluations.
Implementation of the new risk-informed, performance-based fire
protection licensing basis, which complies with the requirements in
10 CFR 50.48(a) and 10 CFR 50.48(c), as well as the guidance
contained in RG 1.205, will not result in new or different kinds of
accidents. The NRC considers that NFPA 805 provides an acceptable
methodology and performance criteria for licensees to identify fire
protection systems and features that are an acceptable alternative
to the 10 CFR 50, Appendix R fire protection features (69 FR 33536,
June 16, 2004). No new modes of operation are introduced by the
proposed amendment, nor will it create any failure mode not bounded
by previously evaluated accidents. Further, the impacts of the
proposed change are not directly assumed in any safety analysis to
initiate an accident sequence.
The requirements in NFPA 805 address only fire protection and
the impacts of fire effects on the plant have been evaluated. The
proposed fire protection program changes do not involve new failure
mechanisms or malfunctions that could initiate a new or different
kind of accident beyond those already analyzed in the SAR. Based on
this, as well as the discussion above, the implementation of this
amendment to transition the FPP at ANO-2 to one based on NFPA 805,
in accordance with 10 CFR 50.48(c), does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Criterion 3
The Proposed Change Does Not Involve a Significant Reduction in
a Margin of safety.
Operation of ANO-2 in accordance with the proposed amendment
does not involve a significant reduction in a margin of safety. The
transition to a new risk-informed, performance-based fire protection
licensing basis that complies with the requirements in 10 CFR
50.48(a) and 10 CFR 50.48(c) does not alter the manner in which
safety limits, limiting safety system settings, or limiting
conditions for operation are determined. The safety analysis
acceptance criteria are not affected by this change. The proposed
amendment does not adversely affect existing plant safety margins or
the reliability of equipment assumed in the SAR to mitigate
accidents. The proposed change does not adversely impact systems
that respond to safely shut down the plant and maintain the plant in
a safe shutdown condition. In addition, the proposed amendment will
not result in plant operation in a configuration outside the design
basis for an unacceptable period of time without implementation of
appropriate compensatory measures.
The risk evaluations for plant changes, in part as they relate
to the potential for reducing a safety margin, were measured
quantitatively for acceptability using the delta risk (i.e.,
[Delta]CDF and [Delta]LERF) criteria from Section 5.3.5,
``Acceptance Criteria,'' of NEI 04-02, as well as the guidance
contained in RG 1.205. Engineering analyses, which may include
engineering evaluations, probabilistic safety assessments, and fire
modeling calculations, have been performed to demonstrate that the
performance-based methods of NFPA 805 do not result in a significant
reduction in the margin of safety. As such, the proposed changes are
evaluated to ensure that risk and safety margins are kept within
acceptable limits. Based on the above, the implementation of this
amendment to transition the FPP at ANO-2 to one based on NFPA 805,
in accordance with 10 CFR 50.48(c), will not significantly reduce 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: Joseph A. Aluise, Associate General
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: March 26, 2013.
Description of amendment request: The amendment would incorporate
the NRC-approved Technical Specifications Task Force (TSTF) change
traveler TSTF-422, Revision 2, ``Change in Technical Specifications End
States (CE NPSD-1186).'' The proposed amendment would modify Technical
Specifications (TS) to risk-inform requirements regarding selected
Required Action End States.
The NRC issued a ``Notice of Availability (NOA) of the Models For
Plant-Specific Adoption of Technical Specifications Task Force (TSTF)
Traveler TSTF-422, Revision 2, `Change In Technical Specifications End
States (CE NPSD-1186),' For Combustion Engineering (CE) Pressurized
Water Reactor (PWR) Plants Using the Consolidated Line Item Improvement
Process (CLIIP),'' in the Federal Register on April 7, 2011 (76 FR
19510), which included the no significant hazards consideration, safety
evaluation, and required commitments for the proposed changes as part
of the consolidated line item improvement process (CLIIP).
In its application dated March 26, 2013, the licensee has concluded
that the technical basis presented in the TSTF proposal and the safety
evaluation are applicable to Arkansas Nuclear One, Unit 2, and the
proposed amendment is consistent with the Standard Technical
Specifications (STS) changes described in TSTF-422, Revision 2, but
with certain variations and/or deviations from TSTF-422, Revision 2.
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 allows a change to certain required end
states when the Technical Specification (TS) Completion Times (CTs)
for remaining in power operation are exceeded. Most of the requested
TS changes are to permit an end state of hot shutdown (Mode 4)
rather than an end state of cold shutdown (Mode 5)
[[Page 44173]]
contained in the current TS. The request was limited to: (1) those
end states where entry into the shutdown mode is for a short
interval; (2) entry is initiated by inoperability of a single train
of equipment or a restriction on a plant operational parameter,
unless otherwise stated in the applicable TS; and (3) the primary
purpose is to correct the initiating condition and return to power
operation as soon as is practical. Risk insights from both the
qualitative and quantitative risk assessments were used in specific
TS assessments. Such assessments are documented in Section 5.5 of CE
NPSD-1186, Rev 0, ``Technical Justification for the Risk-Informed
Modification to Selected Required Action End States for CEOG
[Combustion Engineering Owners Group] Member PWRs.'' The assessments
provide an integrated discussion of deterministic and probabilistic
issues, focusing on specific TSs, which are used to support the
proposed TS end state and associated restrictions. Therefore, the
probability of an accident previously evaluated is not significantly
increased, if at all. The consequences of an accident after adopting
proposed TSTF-422 are no different than the consequences of an
accident prior to adopting TSTF-422. Therefore, the consequences of
an accident previously evaluated are not significantly affected by
this change. The addition of a requirement to assess and manage the
risk introduced by this change will further minimize possible
concerns.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing a change to certain required end states when the TS CTs for
remaining in power operation are exceeded, i.e., entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change and the
commitment by the licensee to adhere to the guidance in WCAP-16364-
NP, Revision 2, ``Implementation Guidance for Risk Informed
Modification to Selected Required Action End States at Combustion
Engineering NSSS Plants (TSTF-422),'' will further minimize possible
concerns.
Therefore, this change does not create the possibility of a new
or different kind of accident from an accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change allows, for some systems, entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed. The CEOG's risk assessment approach is
comprehensive and follows NRC staff guidance as documented in [NRC
Regulatory Guide (RG) 1.174, ``An Approach for Using Probabilistic
Risk Assessment in Risk-Informed Decision Making on Plant Specific
Changes to the Licensing Basis,'' August 1998, and RG 1.177, ``An
Approach for Pant Specific Risk-Informed Decision Making: Technical
Specifications,'' August 1998.]. In addition, the analyses show that
the criteria of the three-tiered approach for allowing TS changes
are met. The risk impact of the proposed TS changes was assessed
following the three-tiered approach recommended in RG 1.177. A risk
assessment was performed to justify the proposed TS changes. The net
change to the margin of safety is insignificant.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Council--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
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: May 21, 2013.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) moderator temperature
coefficient (MTC) surveillance requirements associated with the
implementation of Topical Report WCAP-16011-P-A, ``Startup Test
Activity Reduction (STAR) Program,'' which describes the methods to be
used for the implementation of reduction in the startup testing
requirements. The changes are consistent with the Nuclear Regulatory
Commission (NRC)-approved Industry/Technical Specification Task Force
(TSTF) Standard Technical Specifications change TSTF-486, Revision 2 as
included in NUREG-1432, Revision 4.0, Standard Technical
Specifications--Combustion Engineering (CE) Plants.
The NRC staff published a notice of opportunity for comment in the
Federal Register on July 27, 2007 (72 FR 41360), on possible amendments
adopting TSTF-486 using the NRC's consolidated line-item improvement
process for amending licensees' TSs, which included a model safety
evaluation (SE) and model no significant hazards consideration (NSHC)
determination. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on September 6, 2007 (72 FR
51259), which included the resolution of public comments on the model
SE and model NSHC determination. The licensee affirmed in its
application dated May 21, 2013, that the proposed changes to the TSs
satisfy the intent of TSTF-486.
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 NSHC, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed changes for St. Lucie Units 1 and 2 revise the MTC
Technical Specification 4.1.1.4.1 and 4.1.1.4.2 for each Unit, to
implement the requirements of the topical report WCAP-16011-P-A,
STAR Program.
The MTC is not an initiator to any accident previously
evaluated. Therefore, there is no significant increase in the
probability of any accident previously evaluated. The MTC is an
input to the accident analyses used to predict plant behavior in the
event of an accident. The MTC limits specified in the Technical
Specifications/COLR [core operating limit report] remain unchanged.
WCAP-16011-P-A demonstrated, and the NRC concurred, that the
modified MTC verification is adequate to ensure that MTC stays
within the limits. The consequences of an accident after adopting
TSTF-486 are no different than the consequences of an accident prior
to adoption. Likewise, the deviations from the implementation of
TSTF-486 requirements being adopted in this license amendment do not
have any effect on the probability of occurrence or consequences of
accidents previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No new or different accidents will result from implementation of
the proposed changes. The changes do not involve a physical
alteration of the plant (i.e., no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different operating requirements or eliminate any existing
requirements. The changes do not alter limits and assumptions made
in the safety analysis. The proposed
[[Page 44174]]
changes are consistent with the safety analysis assumptions and
current plant operating practice. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
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?
TSTF-486 provides the means and requirements for CE-designed
plants to implement the previously approved WCAP-16011-P-A for MTC
verification at startup. MTC is a parameter controlled in the
licensee's TS/COLR, including surveillance requirements. As stated
previously, WCAP-16011-P-A describes methods to reduce the
requirements for startup testing. The proposed changes to the TS,
supported by TSTF-486, have been reviewed and found to be consistent
with WCAP-16011-P-A. The changes in the license amendment which
deviate from TSTF-486 requirements are justified to be acceptable
and do not affect the margin of safety. The MTC limits are
unaffected and an acceptable method will be used to verify the MTC
to be within its limit. Therefore, the proposed changes do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Jessie F. Quichocho.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: April 25, 2013.
Description of amendment request: The proposed license amendment
request would revise certain requirements from Section 5,
``Administrative Controls,'' of the Crystal River Unit 3 (CR-3)
Improved Technical Specifications (ITSs). The revisions would include
the following sections: 5.1 ``Responsibility;'' 5.2 ``Organization;''
5.6 ``Procedures, Programs and Manuals;'' 5.7 ``Reporting
Requirements;'' and 5.8 ``High Radiation Area,'' which are no longer
applicable, as CR-3 is in a permanently defueled condition.
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?
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?
No. This changes 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.
3. Does the proposed change involve a significant reduction in a
margin of safety?
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?
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?
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?
No. The changes proposed here for control room staffing do not
directly involve any limits or parameters and therefore cannot
affect ant 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?
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?
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?
No. This change does not directly involve any physical limits or
parameters and therefore cannot affect any margin of safety.
D. ITS Section 5.1.2.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?
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?
No. This change in the description of functional responsibility
of organizational positions cannot create a new or different
[[Page 44175]]
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?
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?
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?
No. This change does not introduce any changes to the function
of any plant structures, systems, or components there it cannot
create a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
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?
No. This change, in conjunction with new paragraph 5.2.2.e,
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?
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?
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?
No. This change continues to ensure that the minimum shift
compliment of qualified 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?
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?
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?
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?
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?
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 deleted, since CR-3 is no longer authorized to
have fuel in the reactor.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. This administrative change 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?
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?
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.d (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?
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?
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?
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.e (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?
No. Certified Fuel Handlers are specifically trained and
qualified to safely handle irradiated fuel. Applying these
qualifications to the supervision of fuel movement ensures
[[Page 44176]]
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?
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?
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?
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?
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?
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?
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 accident from any accident previously evaluated?
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?
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?
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?
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?
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?
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?
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?
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?
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?
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?
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?
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?
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?
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?
[[Page 44177]]
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?
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?
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 5.6.2.10: 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?
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?
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?
No. This change does not directly involve any limits or
parameters and therefore cannot affect any margin of safety.
T. 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?
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?
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?
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.
U. 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?
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?
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?
No. This change does not directly involve any limits or
parameters and therefore cannot affect any margins of safety.
V. 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?
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?
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?
No. Since CR-3 can never load a core into the reactor again,
eliminating this control program cannot affect any margin of safety.
W. 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?
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?
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?
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.
X. 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?
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?
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?
No. This change does not directly involve any limits or
parameters and therefore cannot affect any margin of safety.
[[Page 44178]]
Y. ITS Section 5.7.2: Special Reports
This section is being eliminated.
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. Eliminating reporting requirements for programs that are no
longer required or conditions that cannot exist 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?
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?
No. Eliminating reporting requirements that are no longer
required cannot affect any margin of safety.
Z. 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?
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?
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?
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.
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 https://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 pdr.resource@nrc.gov.
Northern States Power Company--Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2,
Goodhue County, Minnesota
Date of application for amendments: July 25, 2012.
Brief description of amendments: The amendments revise Technical
Specifications (TSs) 3.4.19--``Steam Generator (SG) Tube Integrity,''
5.5.8--``Steam Generator (SG) Program,'' and 5.6.7--``Steam Generator
Tube Inspection Report'' to apply the appropriate program attributes to
the Unit 2 replacement steam generators that are planned for
installation in fall 2013. The amendments also revise the PINGP Units 1
and 2 TSs to adopt the program improvements in Technical Specifications
Task Force Traveler (TSTF) 510, Revision 2, ``Revision to Steam
Generator Program Inspection Frequencies and Tube Sample Selection.''
Date of issuance: July 2, 2013.
Effective date: As of the date of issuance and shall be implemented
within 60 days after reactor startup following Unit 2 steam generator
replacements.
Amendment Nos.: 208 and 195.
Renewed Facility Operating License Nos. DPR-42 and DPR-60:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 14, 2012 (77
FR 56881).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 2, 2013.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company. Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Burke County,
Georgia
Date of application for amendments: March 20, 2013.
Brief description of amendment: The amendment authorizes a
departure from the Vogtle Electric Generating Plant Units 3 and 4
plant-specific Design Control Document (DCD) material incorporated into
the Updated Final Safety Analysis Report (UFSAR) by revising the
structural analysis requirements to provide alternative requirements
for development of headed reinforcement bars (T-heads) within the
nuclear island structures above the basemat elevation.
Date of issuance: May 22, 2013.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: Unit 3-9 and Unit 4-9.
[[Page 44179]]
Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised
the Facility Combined Licenses.
Date of initial notice in Federal Register: April 16, 2013 (78 FR
22573).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 22, 2013.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 15th day of July 2013.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2013-17370 Filed 7-22-13; 8:45 am]
BILLING CODE 7590-01-P