[Federal Register Volume 78, Number 141 (Tuesday, July 23, 2013)]
[Notices]
[Pages 44167-44179]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-17370]
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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 http://www.regulations.gov and search for Docket ID NRC-2103-0158. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected]. 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 http://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 http://www.nrc.gov/reading-rm/adams.html. To begin the
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC's
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. Documents may be viewed in
ADAMS by performing a search on the document date and docket number.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2013-0158 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 http://www.regulations.gov as well as entering the comment submissions into
ADAMS, and the NRC does not edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information in their comment submissions
that they do not want to be publicly disclosed. Your request should
state that the NRC will not edit comment submissions to remove such
information before making the comment submissions available to the
public or entering the comment submissions into ADAMS.
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses and Combined Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 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 http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or a presiding
officer designated by the Commission or by the Chief Administrative
Judge of the Atomic Safety and Licensing Board Panel, will rule on the
request and/or petition; and the Secretary or the Chief Administrative
Judge of the Atomic Safety and Licensing Board will issue a notice of a
hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the requestor/petitioner
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
requestor/petitioner shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/petitioner intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the requestor/petitioner to relief. A requestor/
petitioner who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held. If
the final determination is that the amendment request involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment. If the final determination is that the amendment
request involves a significant hazards consideration, then any hearing
held would take place before the issuance of any amendment.
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 [email protected], or by
telephone at 301-415-1677, to request (1) a digital information (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a request or
petition for hearing (even in instances in which the participant, or
its counsel or representative, already holds an NRC-issued digital ID
certificate). Based upon this information, the Secretary will establish
an electronic docket for the hearing in this proceeding if the
Secretary has not already established an electronic docket.
Information about applying for a digital ID certificate is
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in the NRC's ``Guidance for
Electronic Submission,'' which is available on the agency's public Web
site at http://www.nrc.gov/site-help/e-submittals.html. Participants
may attempt to use other software not listed on the Web site, but
should note that the NRC's E-Filing system does not support unlisted
software, and the NRC Meta System Help Desk will not be able to offer
assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through Electronic Information Exchange System, users
will be required to install a Web browser plug-in from the NRC Web
site. Further information on the Web-based submission form, including
the installation of the Web browser plug-in, is available on the NRC's
public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by email at
[email protected], or by a toll-free call at 1-866 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville,
Maryland, 20852, Attention: Rulemaking and Adjudications Staff.
Participants filing a document in this manner are responsible for
serving the document on all other participants. Filing is considered
complete by first-class mail as of the time of deposit in the mail, or
by courier, express mail, or expedited delivery service upon depositing
the document with the provider of the service. A presiding officer,
having granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the
Commission, or the presiding officer. Participants are requested not to
include personal privacy information, such as social security numbers,
home addresses, or home phone numbers in their filings, unless an NRC
regulation or other law requires submission of such information.
However, a request to intervene will require including information on
local residence in order to demonstrate a proximity assertion of
interest in the proceeding. With respect to copyrighted works, except
for limited excerpts that serve the purpose of the adjudicatory filings
and would constitute a Fair Use application, participants are requested
not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Requests for hearing,
petitions for leave to intervene, and motions for leave to file new or
amended contentions that are filed after the 60-day deadline will not
be entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the following three
factors in 10 CFR 2.309(c)(1): (i) the information upon which the
filing is based was not previously available, (ii) the information upon
which the filing is based is materially different from information
previously available, and (iii) the filing has been submitted in a
timely fashion based on the availability of the subsequent information.
For further details with respect to this license amendment
application, see the application for amendment which is available for
public inspection at the
[[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 http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who
encounter problems in accessing the documents located in ADAMS, should
contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737, or
by email to [email protected].
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 http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR's Reference staff at 1-800-397-4209, 301-415-4737 or by
email to [email protected].
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