Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 44167-44179 [2013-17370]

Agencies

• NUCLEAR REGULATORY COMMISSION

[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]


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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 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