[Federal Register Volume 79, Number 150 (Tuesday, August 5, 2014)]
[Notices]
[Pages 45484-45495]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-17949]


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NUCLEAR REGULATORY COMMISSION

[NRC-2014-0168]


Applications and Amendments to Facility Operating Licenses and 
Combined Licenses Involving Proposed No Significant Hazards 
Considerations and Containing Sensitive Unclassified Non-Safeguards 
Information and Order Imposing Procedures for Access to Sensitive 
Unclassified Non-Safeguards Information

AGENCY: Nuclear Regulatory Commission.

ACTION: License amendment request; opportunity to comment, request a 
hearing, and petition for leave to intervene; order.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) received and is 
considering approval of seven amendment requests. The amendment 
requests are for James A. Fitzpatrick Nuclear Power Plant; Pilgrim 
Nuclear Power Station; Calvert Cliffs Nuclear Power Plant; LaSalle 
County Station, Units 1 and 2 (two requests); Nine Mile Point Nuclear 
Station, Unit 2; Prairie Island Nuclear Power Plant, Units 1 and 2. For 
each amendment request, the NRC proposes to determine that they involve 
no significant hazards consideration. In addition, each amendment 
request contains sensitive unclassified non-safeguards information 
(SUNSI).

DATES: Comments must be filed by September 4, 2014. A request for a 
hearing must be filed by October 6, 2014. Any potential party as 
defined in Sec.  2.4 of Title 10 of the Code of Federal Regulations (10 
CFR), who believes access to SUNSI is necessary to respond to this 
notice must request document access by August 15, 2014.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0168. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: 3WFN-06-A44M, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.

[[Page 45485]]

    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear 
Reactor Regulation U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-5411, email: [email protected].

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2014-0168 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0168.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2014-0168 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

II. Background

    Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as 
amended (the Act), the NRC is publishing this notice. The Act requires 
the Commission to publish notice of any amendments issued, or proposed 
to be issued and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This notice includes notices of amendments containing SUNSI.

III. 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 Sec.  50.92 of Title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated, or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR Part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC's regulations are accessible electronically from the NRC Library on 
the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed within 60 days, 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

[[Page 45486]]

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 set forth 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, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, the 
Commission will make a final determination on the issue of no 
significant hazards consideration. The final determination will serve 
to decide when the hearing is held. If the final determination is that 
the amendment request involves no significant hazards consideration, 
the Commission may issue the amendment and make it immediately 
effective, notwithstanding the request for a hearing. Any hearing held 
would take place after issuance of the amendment. If the final 
determination is that the amendment request involves a significant 
hazards consideration, then any hearing held would take place before 
the issuance of any amendment.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten 10 days prior to the filing deadline, the participant should 
contact the Office of the Secretary by email at [email protected], 
or by telephone at 301-415-1677, to request (1) a digital 
identification (ID) certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at http://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic filing must be submitted to the E-Filing system 
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of 
a transmission, the E-Filing system time-stamps the document and sends 
the submitter an email notice confirming receipt of the document. The 
E-Filing system also distributes an email notice that provides access 
to the document to the NRC's Office of the General Counsel and any 
others who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at http://www.nrc.gov/site-help/e-submittals.html, by email to 
[email protected], or by a toll-

[[Page 45487]]

free call at 1-866-672-7640. The NRC Meta System Help Desk is available 
between 8 a.m. and 8 p.m., Eastern Time, Monday through Friday, 
excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, a request to intervene will require including information on 
local residence in order to demonstrate a proximity assertion of 
interest in the proceeding. With respect to copyrighted works, except 
for limited excerpts that serve the purpose of the adjudicatory filings 
and would constitute a Fair Use application, participants are requested 
not to include copyrighted materials in their submission.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii).
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
Fitzpatrick Nuclear Power Plant (JAF), Oswego County, New York

    Date of amendment request: May 1, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14143A316.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The 
amendment would revise Technical Specification (TS) 2.0, ``Safety 
Limits (SLs),'' by including new values for the Safety Limit Minimum 
Critical Power Ratio for both single and dual recirculation loop 
operation.
    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. The operation of JAF in accordance with the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The basis of the Safety Limit Minimum Critical Power Ratio 
(SLMCPR) is to ensure no mechanistic fuel damage is calculated to 
occur if the limit is not violated. The new SLMCPR values preserve 
the existing margin to transition boiling and probability of fuel 
damage is not increased. The derivation of the revised SLMCPR for 
JAF, for incorporation into the Technical Specifications and its use 
to determine plant and cycle-specific thermal limits, has been 
performed using NRC approved methods. These plant-specific 
calculations are performed each operating cycle and if necessary, 
will require future changes to these values based upon revised core 
designs. The revised SLMCPR values do not change the method of 
operating the plant and have no effect on the probability of an 
accident initiating event or transient.
    Based on the above, JAF has concluded that the proposed change 
will not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The operation of JAF in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed changes result only from a specific analysis for 
the JAF core reload design. These changes do not involve any new or 
different methods for operating the facility. No new initiating 
events or transients result from these changes.
    Based on the above, JAF has concluded that the proposed change 
will not create the possibility of a new or different kind of 
accident from those previously evaluated.
    3. The operation of JAF in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety.
    The new SLMCPR is calculated using NRC approved methods with 
plant and cycle specific parameters for the current core design. The 
SLMCPR value remains conservative enough to ensure that greater than 
99.9% of all fuel rods in the core will avoid transition boiling if 
the limit is not violated, thereby preserving the fuel cladding 
integrity. The operating MCPR limit is set appropriately above the 
safety limit value to ensure adequate margin when the cycle specific 
transients are evaluated. Accordingly, the margin of safety is 
maintained with the revised values.
    As a result, JAF has determined that the proposed change will 
not result in 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: Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Benjamin G. Beasley.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: January 31, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14042A166.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The 
amendment would revise the Cyber Security Plan (CSP) Milestone 8 full 
implementation date, as set forth in the CSP Implementation Schedule.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 45488]]

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 to the CSP Implementation Schedule is 
administrative in nature. This change does not alter accident 
analysis assumptions, add any initiators, or affect the function of 
plant systems or the manner in which systems are operated, 
maintained, modified, tested, or inspected. The proposed change does 
not require any plant modifications which affect the performance 
capability of the structures, systems, and components relied upon to 
mitigate the consequences of postulated accidents and has no impact 
on the probability or consequences of an 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 to the CSP Implementation Schedule is 
administrative in nature. This proposed change does not alter 
accident analysis assumptions, add any initiators, or affect the 
function of plant systems or the manner in which systems are 
operated, maintained, modified, tested, or inspected. The proposed 
change does not require any plant modifications which affect the 
performance capability of the structures, systems, and components 
relied upon to mitigate the consequences of postulated accidents and 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    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.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed change to 
the CSP Implementation Schedule is administrative in nature. In 
addition, the milestone date delay for full implementation of the 
CSP has no substantive impact because other measures have been taken 
which provide adequate protection during this period of time. 
Because there is no change to established safety margins as a result 
of this change, the proposed change does not involve a significant 
reduction in a margin of safety.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeanne Cho, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Benjamin G. Beasley.

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland

    Date of amendment request: September 24, 2013. A publicly-available 
version is in ADAMS under Accession Nos. ML13301A673 and ML13301A674.
    Description of amendments request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The 
amendments would modify the fire protection licensing basis to 
transition to the requirements of National Fire Protection Association 
(NFPA) standard 805, pursuant to 10 CFR 50.48(c).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The purpose of the proposed amendment is to permit Calvert 
Cliffs Units 1 and 2 to adopt a new fire protection licensing basis 
that complies with the requirements of 10 CFR 50.48(a) and (c) and 
the guidance in Regulatory Guide 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 Appendix R required fire 
protection features (69 FR 33536, June 16, 2004).
    Engineering analyses, which may include engineering evaluations, 
probabilistic safety assessments, and fire modeling calculations, 
have been performed to demonstrate that the performance-based 
requirements of NFPA 805 have been satisfied. The Updated Final 
Safety Analysis Report documents the analysis of design basis 
accidents at Calvert Cliffs Units 1 and 2. The proposed amendment 
does not affect accident initiators, nor does it alter design 
assumptions, conditions, or configurations of the facility that 
would increase the probability of accidents previously evaluated. 
Further, the changes to be made for fire hazard protection and 
mitigation do not adversely affect the ability of structures, 
systems or components to perform their design functions for accident 
mitigation, nor do they affect the postulated initiators or assumed 
failure modes for accidents described and evaluated in the UFSAR. 
Structures, systems or components required to safely shutdown the 
reactor and to maintain it in a safe shutdown condition will remain 
capable of performing their design function.
    NFPA 805, taken as a whole, provides an acceptable alternative 
for satisfying General Design Criterion 3 of Appendix A to 10 CFR 
50, meets the underlying intent of the NRC's existing fire 
protection regulations and guidance, and provides defense-in-depth. 
The goals, performance objectives and performance criteria specified 
in Chapter 1 of the standard ensure that, if there are any increases 
in core damage frequency or risk, the increase will be small and 
consistent with the intent of the Commission's Safety Goal Policy.
    The proposed amendment will not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of any accident previously 
evaluated and equipment required to mitigate an accident remains 
capable of performing the assumed function. The applicable 
radiological dose criteria will continue to be met.
    Based on the above discussion, it is concluded that the proposed 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any kind of accident previously 
evaluated?
    Response: No.
    The proposed change does not alter the requirements or functions 
for systems required during accident conditions. Implementation of 
the new fire protection licensing basis, which complies with the 
requirements of 10 CFR 50.48(a) and (c) and the guidance of 
Regulatory Guide 1.205, will not result in new or different 
accidents.
    The proposed amendment does not introduce new or different 
accident initiators, nor does it alter design assumptions, 
conditions, or configurations of the facility in such a manner as to 
introduce new or different accident initiators. The proposed 
amendment does not adversely affect the ability of structures, 
systems, or components to perform their design function. Structures, 
systems or components required to safely shutdown the reactor and 
maintain it in a safe shutdown condition remain capable of 
performing their design functions.
    The requirements of NFPA 805 address only fire protection and 
the impacts of fire on the plant that have previously been 
evaluated. Thus, implementation of the proposed amendment would not 
create the possibility of a new or different kind of accident beyond 
those already analyzed in the UFSAR. No new accident scenarios, 
transient precursors, failure mechanisms, or

[[Page 45489]]

limiting single failures will be introduced, and there will be no 
adverse effect or challenges imposed on any safety related system as 
a result of the proposed amendment.
    Based on the above discussion, it is concluded that the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The purpose of the proposed amendment is to permit Calvert 
Cliffs Units 1 and 2 to adopt a new fire protection licensing basis 
which complies with the requirements on 10 CFR 50.48(a) and (c) and 
the guidance in Regulatory Guide 1.205. The NRC considers that NFPA 
805 provides an acceptable methodology and performance criteria for 
licensees to identify for protection systems and features that are 
an acceptable alternative to the 10 CFR 50 Appendix R required fire 
protection features (69 FR 33536, June 16, 2004).
    The overall approach of NFPA 805 is consistent with the key 
principals for evaluating license basis changes, as described in 
Regulatory Guide 1.174, is consistent with the defense-in-depth 
philosophy, and maintains sufficient safety margins. Engineering 
analyses, which may include engineering evaluations, probabilistic 
safety assessments, and fire modeling calculations, have been 
performed to demonstrate that the performance based methods do not 
result in a significant reduction in the margin of safety.
    The proposed amendment 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 to mitigate accidents in the UFSAR. The proposed 
amendment does not adversely affect the ability of structures, 
systems or components to perform their design function. Structures, 
systems or components required to safely shutdown the reactor and to 
maintain it in a safe shutdown condition remain capable of 
performing their design function.
    Based on the above discussion, it is concluded that the proposed 
amendment 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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Exelon Generation, 200 
Exelon Way, Kennett Square, PA 19348.
    NRC Branch Chief: Benjamin G. Beasley.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station (LSCS), Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: July 12, 2012, as supplemented by 
letters dated September 17, 2012, January 18, 2013, February 11, 2013, 
October 4, 2013, and February 20, 2014. Publicly-available versions are 
in ADAMS under Accession Nos. ML12200A330, ML122690041, ML13022A476, 
ML13042A405, ML13282A339, and ML14066A250.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment would modify Technical Specification 3.7.3, ``Ultimate Heat 
Sink,'' by changing the maximum allowable temperature of the ultimate 
heat sink from a fixed limit of 101.25 degrees Fahrenheit to a variable 
limit between 101.25 and 104 degrees Fahrenheit depending on the time 
of day. The proposed amendment was initially published in the Federal 
Register Biweekly notice on April 2, 2013 (78 FR 19746).
    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 makes no physical changes to the plant, nor 
does it alter any of the assumptions or conditions upon which the 
UHS [ultimate heat sink] is designed. These assumptions and 
conditions as described in the LSCS UFSAR [updated final safety 
analysis report] include failure of the cooling lake dike, a loss of 
offsite power and a DBA [design-basis accident] LOCA [loss-of-
coolant accident] on one unit, and a normal shutdown of the other 
unit.
    The accidents analyzed in the UFSAR are assumed to be initiated 
by the failure of plant structures, systems, or components (SSCs). 
An inoperable UHS is not an initiator of any analyzed events as 
described in the UFSAR. The impact on the structural integrity of 
the UHS due to a potential increase water temperature prior to and 
during the UHS design basis event has been evaluated, and does not 
increase the probability of the failure of the cooling lake dike. 
The proposed temperature limit for cooling water supplied to the 
plant from the CSCS [core standby cooling system] Pond could reduce 
the commercial capability of the LSCS units; however, it does not 
result in an increase in the probability of occurrence for any of 
the events described in the UFSAR.
    The basis provided in Regulatory Guide 1.27, ``Ultimate Heat 
Sink for Nuclear Power Plants,'' Revision 2, dated January 1976, was 
employed for the temperature analysis of the LSCS UHS to implement 
General Design Criteria 2, ``Design bases for protection against 
natural phenomena,'' and 44, ``Cooling water,'' of Appendix A to 10 
CFR 50 [Title 10 of the Code of Federal Regulations Part 50]. 
Revision 1 of this Regulatory Guide was employed for the original 
design and licensing basis of the LSCS UHS, and Revision 2 of this 
Regulatory Guide was used for the subsequent evaluation, which 
investigated the potential for changing the average water 
temperature of the cooling water supplied to the plant from the CSCS 
Pond from a fixed temperature limit to a limit based on the time of 
day. The meteorological conditions chosen for the LSCS UHS analysis 
utilized a critical period consisting of the most severe 33 hour 
transit time followed by the subsequent 31 calendar days based on 
historical data. The heat loads selected for the UHS analysis 
considered failure of the cooling lake dike, a loss of offsite power 
and a DBA LOCA on one unit, and a normal shutdown of the other unit. 
The LSCS cooling lake is conservatively assumed to be unavailable at 
the start of the event. The analysis shows that with an initial UHS 
temperature less than or equal to the proposed time-of-day-based 
limit, the required safety-related heat loads can be adequately 
cooled for 30 days while continuing to ensure safety-related cooling 
water temperature remains less than the design temperature for LSCS, 
Units 1 and 2.
    Based on the above, it has been demonstrated that the change of 
the initial temperature limit for cooling water supplied to the 
plant from the CSCS Pond to less than or equal to a temperature 
based on the time of day will not impede the ability of the 
equipment and components cooled by the UHS during a UHS design basis 
event to perform their safety functions.
    There is no impact of this change on LSCS safety analyses 
including the consequences of all postulated events since all 
required safety-related equipment continues to perform as designed. 
The effects of the proposed change on the ability of the UHS to 
assure that a 30-day supply of water is available considering losses 
due to evaporation, seepage, and firefighting have been considered. 
Sufficient inventory remains available to mitigate the design basis 
event for the LSCS UHS for the required 30-day period.
    Therefore, the proposed activity 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 physically alter the operation, 
testing, or maintenance of any plant SSCs beyond operating with a 
UHS temperature limit based on the time of day. The proposed change 
is supported by appropriate design analysis. Moreover, the UHS 
temperature does not initiate accident

[[Page 45490]]

precursors. The impact of increased UHS temperature can affect the 
commercial operation of the plant, but the proposed change would not 
create any accident not considered in the LSCS UFSAR.
    This proposed change will not alter the manner in which 
equipment operation is initiated, nor will the functional demands on 
credited equipment be changed. No alteration in the procedures that 
ensure the LSCS units remain within analyzed limits is proposed, and 
no change is being made to procedures relied upon to respond to an 
off-normal event. As such, no new failure modes are being 
introduced. The proposed change does not alter assumptions made in 
the LSCS safety analysis.
    Changing the temperature of cooling water supplied to the plant 
from the CSCS Pond (i.e., the UHS) as proposed has no impact on 
plant accident response. The proposed temperature limits do not 
introduce new failure mechanisms for SSCs. An engineering analysis 
performed to support the change in temperature of cooling water 
supplied to the plant from the CSCS Pond provides the basis to 
conclude that the equipment is adequately designed for operation as 
proposed.
    All systems that are important to safety will continue to be 
operated and maintained within their design bases, and the proposed 
change will continue to ensure that all associated systems and 
components are operated reliably within their design capabilities.
    The proposed change will ensure the maximum temperature of the 
cooling water supplied to the plant during the UHS design basis 
event remains less than the current safety-related cooling water 
design temperature for LSCS, Units 1 and 2. Therefore, there is no 
impact of this change on the LSCS safety analyses including 
inventory and cooling requirements for safety-related systems using 
the UHS as their cooling water supply.
    All systems will continue to be operated within their design 
capabilities, no new failure modes are introduced, nor is there any 
adverse impact on plant equipment; therefore, the proposed change 
does not result in 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 margin of safety is determined by the design and 
qualification of the plant equipment, the operation of the plant 
within analyzed limits, and the point at which protective or 
mitigative actions are initiated. The proposed change does not 
impact any of these factors. There are no required design changes or 
equipment performance parameter changes associated with the proposed 
change. No protection setpoints are affected as a result of this 
change. The proposed change in the limit for the temperature of 
cooling water supplied to the plant from the CSCS Pond will not 
change the operational characteristics of the design of any 
equipment or system. All accident analysis assumptions and 
conditions will continue to be met.
    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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Travis L. Tate.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station (LSCS), Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: December 20, 2013, as supplemented by 
letter dated February 26, 2014. Publicly-available versions are in 
ADAMS under Accession Nos. ML13358A354 and ML14057A549.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The proposed 
amendment would modify LSCS, Unit 1, pressure and temperature curves in 
Technical Specification 3.4.11, ``RCS [Reactor Coolant System] Pressure 
and Temperature (P/T) Limits.''
    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 makes no physical changes to the plant. The 
proposed amendment incorporates the recent ISP [integrated 
surveillance program] results into the NRC-approved methodology of 
the GE Hitachi Nuclear Energy Licensing Topical Report NEDC-33178P-
A, Revision 1, for the preparation of the LSCS, Unit 1 P/T [pressure 
and temperature] limit curves. In 10 CFR 50, Appendix G, 
requirements are established to protect the integrity of the Reactor 
Coolant Pressure Boundary in nuclear power plants. Implementing the 
NRC-approved methodology for calculating P/T limit curves Evaluation 
of Proposed Changes provide an equivalent level of assurance that 
Reactor Coolant Pressure Boundary integrity will be maintained, as 
specified in 10 CFR 50, Appendix G.
    The proposed changes do not adversely affect accident initiators 
or precursors, and do not negatively alter the design assumptions, 
conditions, or configuration of the plant or the manner in which the 
plant is operated and maintained. The ability of structures, 
systems, and components to perform their intended safety functions 
is not altered or prevented by the proposed changes, and the 
assumptions used in determining the radiological consequences of 
previously evaluated accidents are not affected.
    Therefore, the proposed activity 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 revised P/T limits do not alter or involve any design basis 
accident initiators. Reactor Coolant Pressure Boundary integrity 
will continue to be maintained in accordance with 10 CFR 50, 
Appendix G, and the assumed accident performance of plant 
structures, systems and components will not be affected. These 
changes do not involve any physical alteration of the plant (i.e., 
no new or different type of equipment will be installed), and 
installed equipment is not being operated in a new or different 
manner. Thus, no new failure modes are introduced.
    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 changes do not affect the function of the Reactor 
Coolant Pressure Boundary or its response during plant transients. 
By calculating the P/T limits using NRC-approved methodology, 
adequate margins of safety relating to Reactor Coolant Pressure 
Boundary integrity are maintained. The proposed changes do not alter 
the manner in which safety limits, limiting safety system settings, 
or limiting conditions for operation are determined. There are no 
changes to setpoints at which protective actions are initiated, and 
the operability requirements for equipment assumed to operate for 
accident mitigation are not 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Travis L. Tate.

[[Page 45491]]

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York

    Date of amendment request: November 1, 2013, as supplemented by 
letters dated January 21, February 14, February 25, March 10, May 14, 
and June 13, 2014. A publicly-available version is in ADAMS under 
Accession Nos. ML13316B107, ML13316B109, ML13316B110, ML14023A654, 
ML14051A138, ML14064A321, ML14064A322, ML14064A323, ML14064A324, 
ML14071A466, ML14139A416, and ML14169A034.
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The license 
amendment request was originally noticed in the Federal Register (FR) 
on June 6, 2014 (79 FR 32763-32765). This notice is being reissued in 
its entirety to include the revised description of the amendment 
request and revised analysis of the issue of no significant hazards 
consideration submitted by the licensee in its June 13, 2014 
submission. The proposed amendment includes changes to the NMP2 
Technical Specifications (TSs) necessary to: (1) Implement the Maximum 
Extended Load Line Limit Analysis Plus (MELLLA+) expanded operating 
domain; (2) change the stability solution to Detect and Suppress 
Solution--Confirmation Density (DSS-CD); (3) use the TRACG04 analysis 
code; and (4) increase the Safety Limit Minimum Critical Power Ratio 
(SLMCPR) for two recirculation loops in operation.
    The following is a list of the proposed changes to the NMP2 TSs:
     Revise Safety Limit (SL) 2.1.1.2 by increasing the SLMCPR 
for two recirculation loops in operation from >=1.07 to >=1.09.
     Revise the acceptance criterion in TS 3.1.7, ``Standby 
Liquid Control (SLC) System,'' Surveillance Requirement (SR) 3.1.7.7 by 
increasing the discharge pressure from >=1,327 pounds per square inch 
gauge (psig) to >=1,335 psig.
     Change the Required Actions for Condition F of TS 3.3.1.1, 
``Reactor Protection System (RPS) Instrumentation.''
     Change Condition G of TS 3.3.1.1.
     Add new Conditions J and K to TS 3.3.1.1.
     Correct an editorial error in Note 3 to TS SR 3.3.1.1.13 
(i.e., ``ORRM'' is changed to ``OPRM'' [Oscillation Power Range 
Monitor]).
     Eliminate TS SR 3.3.1.1.16 and references to it in TS 
Table 3.3.1.1-1, ``Reactor Protection System Instrumentation.''
     Change the allowable value (AV) for TS Table 3.3.1.1-1, 
Function 2.b, Average Power Range Monitor (APRM)--Flow Biased Simulated 
Thermal Power (STP)--Upscale from ``<= 0.55W + 60.5% [Rated Thermal 
Power] RTP and <= 115.5% RTP'' to ``<= 0.61W + 63.4% RTP and <= 115.5% 
RTP.''
     Add a new note to TS Table 3.3.1.1-1, Function 2.b that 
requires the Flow Biased Simulated Thermal Power--Upscale scram 
setpoint to be reset to the values defined by the Core Operating Limits 
Report (COLR) to implement the Automated Backup Stability Protection 
(BSP) Scram Region in accordance with Required Action F.2 of TS 
3.3.1.1.
     Add a new note to TS Table 3.3.1.1-1, Function 2.e, 
Oscillation Power Range Monitor (OPRM)--Upscale to denote that 
following implementation of DSS-CD, DSS-CD is not required to be armed 
while in the DSS-CD Armed Region during the first reactor startup and 
during the first controlled shutdown that passes completely through the 
DSS-CD Armed Region. However, DSS-CD is considered operable and capable 
of automatically arming for operation at recirculation drive flow rates 
above the DSS-CD Armed Region.
     Change the mode of applicability for TS Table 3.3.1.1-1, 
Function 2.e, OPRM-Upscale from Mode 1 to >=18% RTP.
     Change the allowable value for TS Table 3.3.1.1-1, 
Function 2.e from ``As specified in the COLR'' to ``NA [not 
applicable].''
     TS Limiting Condition for Operation (LCO) 3.4.1, 
``Recirculation Loops Operating,'' is modified to prohibit operation in 
the Maximum Extended Load Line Limit Analysis (MELLLA) domain or 
MELLLA+ expanded operating domain as defined in the COLR when in 
operation with a single recirculation loop.
     Add Required Action B.2 to TS 3.4.1 to identify that 
intentional operation in the MELLLA domain or MELLLA+ domain as defined 
in the COLR is prohibited when a recirculation loop is declared ``not 
in operation'' due to a recirculation loop flow mismatch not within 
limits.
     Revise TS 5.6.5.a.4 to replace ``Reactor Protection System 
Instrumentation Setpoint for the OPRM--Upscale Function Allowable Value 
for Specification 3.3.1.1'' with ``The Manual Backup Stability 
Protection (BSP) Scram Region (Region I), the Manual BSP Controlled 
Entry Region (Region II), the modified APRM Simulated Thermal Power--
High setpoints used in the OPRM (Function 2.e), Automated BSP Scram 
Region, and the BSP Boundary for Specification 3.3.1.1.''
     Add TS 5.6.8, ``OPRM Report,'' to define the contents of 
the report required by new Required Action F.3 of TS 3.3.1.1.
    The NRC's approval of the requested operating domain expansion will 
allow NMP2 to implement operational changes that will increase 
operational flexibility for power maneuvering, compensate for fuel 
depletion, and maintain efficient power distribution in the reactor 
core without the need for more frequent rod pattern changes. MELLLA+ 
supports operation of NMP2 at Current Licensed Thermal Power (CLTP) of 
3,988 Megawatts--Thermal (MWth) with core flow as low as 85% 
of rated core flow. By operating in the MELLLA+ domain, a significantly 
lower number of control rod movements will be required than in the 
present operating domain. This represents a significant improvement in 
operating flexibility. It also provides safer operation, because 
reducing the number of control rod manipulations: (a) Minimizes the 
likelihood of fuel failures, and (b) reduces the likelihood of 
accidents initiated by reactor maneuvers required to achieve an 
operating condition where control rods can be withdrawn.
    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. Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The probability (frequency of occurrence) of Design Basis 
Accidents occurring is not affected by implementing the MELLLA+ 
operating domain and DSS-CD stability solution, because NMP2 
continues to comply with the regulatory and design basis criteria 
established for plant equipment. A SLS [standby liquid control 
system] failure is not a precursor of any previously evaluated 
accident in the NMP2 USAR [updated safety analysis report]. The 
increase to the SLMCPR for two recirculation loops in operation does 
not increase the probability of an evaluated accident. Consequently, 
there is no change in the probability of a previously evaluated 
accident.
    The spectrum of postulated transients was investigated and shown 
to remain within the NRC approved acceptance limits. Fuel integrity 
is maintained by meeting existing design and regulatory limits. 
Further, a probabilistic risk assessment demonstrates that the 
calculated core damage frequency and the large early release 
frequency do not significantly change due to operation in the 
MELLLA+ domain.

[[Page 45492]]

    Challenges to the reactor coolant pressure boundary were 
evaluated for the MELLLA+ operating domain conditions (pressure, 
temperature, flow, and radiation) and were found to meet their 
acceptance criteria for allowable stresses and overpressure margin.
    Challenges to the containment were evaluated and the containment 
and its associated cooling systems continue to meet the current 
licensing basis. The calculated post LOCA [loss-of-coolant accident] 
suppression pool temperature remains acceptable.
    The SLS is used to mitigate the consequences of an Anticipated 
Transient Without SCRAM (ATWS) special event and is used to limit 
the radiological dose during a Loss of Coolant Accident (LOCA). The 
proposed changes do not affect the capability of the SLS to perform 
these two functions in accordance with the assumptions of the 
associated analyses. The ATWS evaluation with the proposed changes 
incorporated demonstrated that all the ATWS acceptance criteria are 
met. The ability of the SLS to mitigate radiological dose in the 
event of a LOCA by maintaining suppression pool pH >=7.0 is not 
affected by these changes.
    This proposed change to the SLMCPR for two recirculation loops 
in operation does not result in any modification to the design or 
operation of the systems that are used in mitigation of accidents. 
Limits have been established, consistent with NRC approved methods, 
to ensure that fuel performance during normal, transient, and 
accident conditions is acceptable. The proposed change to the SLMCPR 
for two recirculation loops in operation continues to conservatively 
establish this safety limit such that the fuel is protected during 
normal operation and during any plant transients or anticipated 
operational occurrences.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    Equipment that could be affected by implementing the MELLLA+ 
operating domain and DSS-CD stability solution was evaluated. No new 
operating mode, safety-related equipment lineup, accident scenario, 
or equipment failure mode was identified. The full spectrum of 
accident considerations was evaluated and no new or different kind 
of accident was identified. The MELLLA+ operating domain and DSS-CD 
stability solution use developed technology and apply it within the 
capabilities of existing plant safety-related equipment in 
accordance with the regulatory criteria (including NRC approved 
codes, standards and methods). No new accident or event precursor 
was identified.
    The long-term stability solution is being changed from the 
currently approved Option III solution to DSS-CD. DSS-CD is designed 
to identify the power oscillation upon inception and initiate 
control rod insertion (scram) to terminate the oscillations prior to 
any significant amplitude growth exceeding the applicable safety 
limits. DSS-CD is based on the same hardware design as Option III. 
However, it introduces an enhanced detection algorithm that detects 
the inception of power oscillations and generates an earlier power 
suppression trip signal. The existing Option III algorithms are 
retained (with generic setpoints) to provide defense-in-depth 
protection for unanticipated reactor instability events.
    Structures, systems, and components (SSCs) previously required 
for the mitigation of a transient remain capable of fulfilling their 
intended design functions. The proposed changes do not adversely 
affect safety-related systems or components and do not challenge the 
performance or integrity of any safety-related system. The physical 
change's to the SLS is limited to the increase in the SLS pump 
discharge pressure acceptance criterion. The proposed changes do not 
otherwise affect the design or operation of the SLS.
    This proposed change to the SLMCPR for two recirculation loops 
in operation does not result in any modification to the design or 
operation of the systems that are used in the mitigation of 
accidents. The proposed change to the SLMCPR for two recirculation 
loops in operation assures that safety criteria are maintained.
    The proposed changes do not adversely affect any current system 
interfaces or create any new interfaces that could result in an 
accident or malfunction of a different kind than was 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. Will the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The MELLLA+ operating domain affects only design and operational 
margins. Challenges to the fuel, reactor coolant pressure boundary, 
and containment were evaluated for the MELLLA+ operating domain 
conditions. Fuel integrity is maintained by meeting existing design 
and regulatory limits. The calculated loads on affected SSCs, 
including the reactor coolant pressure boundary, will remain within 
their design specifications for design basis event categories. No 
NRC acceptance criterion is exceeded.
    Comprehensive analyses of the proposed changes have concluded 
that relevant design and safety acceptance criteria will be met 
without a significant reduction in margins of safety. The analyses 
have demonstrated that the NMP2 SSCs are capable of safely 
performing at MELLLA+ conditions. The analyses identified and 
defined the major input parameters to the Nuclear Steam Supply 
System (NSSS), analyzed NSSS design transients, and evaluated the 
capabilities of the NSSS fluid systems, NSSS/Balance of Plant (BOP) 
interfaces, NSSS control systems, and NSSS and BOP components, as 
appropriate. Radiological consequences of design basis events remain 
within regulatory limits and are not increased significantly. The 
analyses confirmed that NSSS and BOP SSCs are capable of achieving 
MELLLA+ conditions without significant reduction in margins of 
safety.
    Analyses have shown that the integrity of primary fission 
product barriers will not be significantly affected as a result of 
change in the operating domain. Calculated loads on SSCs important 
to safety have been shown to remain within design allowables with 
MELLLA+ conditions for all design basis event categories. Plant 
response to transients and accidents do not result in exceeding 
acceptance criteria. As appropriate, the evaluations that 
demonstrate acceptability of MELLLA+ have been performed using 
methods that have either been reviewed and approved by the NRC 
staff, or that are in compliance with regulatory review guidance and 
standards established for maintaining adequate margins of safety. 
These evaluations demonstrate that there are no significant 
reductions in the margins of safety.
    The SLS is used to mitigate the consequences of an ATWS event 
and is used to limit the radiological dose during a LOCA. The 
proposed changes do not affect the capability of the SLS to perform 
these two functions in accordance with the assumptions of the 
associated analyses. The ATWS evaluation with the proposed changes 
incorporated demonstrated that all the ATWS acceptance criteria are 
met. The ability of the SLS to mitigate radiological dose in the 
event of a LOCA by maintaining suppression pool pH >=7.0 is not 
affected by these changes.
    This proposed change to the SLMCPR for two recirculation loops 
in operation provides a margin of safety by ensuring that no more 
than 0.1% of fuel rods are expected to be in boiling transition if 
the MCPR limit is not violated. The proposed change will ensure the 
appropriate level of fuel protection is maintained. Additionally, 
operational limits are established based on the proposed SLMCPR to 
ensure that the SLMCPR is not violated during all modes of 
operation. This will ensure that the fuel design safety criteria are 
met (i.e., that at least 99.9% of the fuel rods do not experience 
transition boiling during normal operation as well as anticipated 
operational occurrences).
    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 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: Gautam Sen, Senior Counsel, Constellation 
Energy Nuclear Group, LLC, 100 Constellation Way, Suite 200C, 
Baltimore, MD 21202.
    NRC Branch Chief: Benjamin Beasley.

[[Page 45493]]

Northern States Power Company--Minnesota, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota; and Northern States 
Power Company (NSPC)--Minnesota, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of amendment request: November 27, 2013, as supplemented by 
letter dated May 5, 2014. Publicly-available versions are in ADAMS 
under Accession Nos. ML13333B674 and ML14126A727).
    Description of amendment request: This amendment request contains 
sensitive unclassified non-safeguards information (SUNSI). The license 
amendment request pertains to the Cyber Security Plan (CSP) 
implementation schedule change in the completion date for Milestone 8. 
Milestone 8 pertains to the date that full implementation of the CSP 
for all safety, security, and emergency preparedness functions will be 
achieved.
    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 amendment proposes a change to the NSPM Cyber Security Plan 
(CSP) Milestone 8 (M8) full implementation date.
    The revision of the full implementation date for the NSPM CSP 
does not involve modifications to any safety-related structures, 
systems or components (SSCs). Rather, the implementation schedule 
provides a timetable for fully implementing the NSPM CSP. The CSP 
describes how the requirements of 10 CFR 73.54 are to be implemented 
to identify, evaluate, and mitigate cyber-attacks up to and 
including the design basis cyber-attack threat, thereby achieving 
high assurance that the facility's digital computer and 
communications systems and networks are protected from cyber-
attacks. The revision of the NSPM CSP Implementation Schedule will 
not alter previously evaluated design basis accident analysis 
assumptions, add any accident initiators, modify the function of the 
plant safety-related SSCs, or affect how any plant safety-related 
SSCs are operated, maintained, modified, tested, or inspected.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The amendment proposes a change to the NSPM CSP Milestone 8 (M8) 
full implementation date.
    The revision of the full implementation date for the NSPM CSP 
does not involve modifications to any safety-related structures, 
systems or components (SSCs). The implementation of the NSPM CSP 
does not introduce new equipment that could create a new or 
different kind of accident, and no new equipment failure modes are 
created. No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of this proposed 
amendment.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The amendment proposes a change to the NSPM CSP Milestone 8 (M8) 
full implementation date.
    The revision of the full implementation date for the NSPM CSP 
does not involve modifications to any safety-related structures, 
systems or components (SSCs). The margin of safety is associated 
with the confidence in the ability of the fission product barriers 
(i.e., fuel cladding, reactor coolant pressure boundary, and 
containment structure) to limit the level of radiation to the 
public. The proposed amendment does not alter the way any safety-
related SSC functions and does not alter the way the plant is 
operated. The Cyber Security Plan provides assurance that safety-
related SSCs are protected from cyber-attacks. The proposed 
amendment does not introduce any new uncertainties or change any 
existing uncertainties associated with any safety limit. The 
proposed amendment has no effect on the structural integrity of the 
fuel cladding, reactor coolant pressure boundary, or containment 
structure. Based on the above considerations, the proposed amendment 
does not degrade the confidence in the ability of the fission 
product barriers to limit the level of radiation to the public.
    Therefore the proposed change does not involve a 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: Peter M. Glass, Assistant General Counsel, 
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
    NRC Branch Chief: Robert D. Carlson.
Order Imposing Procedures for Access to Sensitive Unclassified Non-
Safeguards Information for Contention Preparation
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
Fitzpatrick, Nuclear Power Plant, Oswego County, New York
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts
Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York
Northern States Power Company--Minnesota, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota; and
Northern States Power Company, Minnesota, Docket Nos. 50-282 and 50-
306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue 
County, Minnesota

    A. This Order contains instructions regarding how potential parties 
to this proceeding may request access to documents containing SUNSI.
    B. Within 10 days after publication of this notice of hearing and 
opportunity to petition for leave to intervene, any potential party who 
believes access to SUNSI is necessary to respond to this notice may 
request such access. A ``potential party'' is any person who intends to 
participate as a party by demonstrating standing and filing an 
admissible contention under 10 CFR 2.309. Requests for access to SUNSI 
submitted later than 10 days after publication of this notice will not 
be considered absent a showing of good cause for the late filing, 
addressing why the request could not have been filed earlier.
    C. The requester shall submit a letter requesting permission to 
access SUNSI to the Office of the Secretary, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, and provide a copy to the Associate General 
Counsel for Hearings, Enforcement and Administration, Office of the 
General Counsel, Washington, DC 20555-0001. The expedited delivery or 
courier mail address for both offices is: U.S. Nuclear Regulatory 
Commission, 11555 Rockville Pike, Rockville, Maryland 20852. The email 
address for the Office of the Secretary and the Office of the General 
Counsel are

[[Page 45494]]

[email protected] and [email protected], respectively.\1\ The 
request must include the following information:
---------------------------------------------------------------------------

    \1\ While a request for hearing or petition to intervene in this 
proceeding must comply with the filing requirements of the NRC's 
``E-Filing Rule,'' the initial request to access SUNSI under these 
procedures should be submitted as described in this paragraph.
---------------------------------------------------------------------------

    (1) A description of the licensing action with a citation to this 
Federal Register notice;
    (2) The name and address of the potential party and a description 
of the potential party's particularized interest that could be harmed 
by the action identified in C.(1); and
    (3) The identity of the individual or entity requesting access to 
SUNSI and the requester's basis for the need for the information in 
order to meaningfully participate in this adjudicatory proceeding. In 
particular, the request must explain why publicly-available versions of 
the information requested would not be sufficient to provide the basis 
and specificity for a proffered contention.
    D. Based on an evaluation of the information submitted under 
paragraph C.(3) the NRC staff will determine within 10 days of receipt 
of the request whether:
    (1) There is a reasonable basis to believe the petitioner is likely 
to establish standing to participate in this NRC proceeding; and
    (2) The requestor has established a legitimate need for access to 
SUNSI.
    E. If the NRC staff determines that the requestor satisfies both 
D.(1) and D.(2) above, the NRC staff will notify the requestor in 
writing that access to SUNSI has been granted. The written notification 
will contain instructions on how the requestor may obtain copies of the 
requested documents, and any other conditions that may apply to access 
to those documents. These conditions may include, but are not limited 
to, the signing of a Non-Disclosure Agreement or Affidavit, or 
Protective Order \2\ setting forth terms and conditions to prevent the 
unauthorized or inadvertent disclosure of SUNSI by each individual who 
will be granted access to SUNSI.
---------------------------------------------------------------------------

    \2\ Any motion for Protective Order or draft Non-Disclosure 
Affidavit or Agreement for SUNSI must be filed with the presiding 
officer or the Chief Administrative Judge if the presiding officer 
has not yet been designated, within 30 days of the deadline for the 
receipt of the written access request.
---------------------------------------------------------------------------

    F. Filing of Contentions. Any contentions in these proceedings that 
are based upon the information received as a result of the request made 
for SUNSI must be filed by the requestor no later than 25 days after 
the requestor is granted access to that information. However, if more 
than 25 days remain between the date the petitioner is granted access 
to the information and the deadline for filing all other contentions 
(as established in the notice of hearing or opportunity for hearing), 
the petitioner may file its SUNSI contentions by that later deadline. 
This provision does not extend the time for filing a request for a 
hearing and petition to intervene, which must comply with the 
requirements of 10 CFR 2.309.
    G. Review of Denials of Access.
    (1) If the request for access to SUNSI is denied by the NRC staff 
after a determination on standing and need for access, the NRC staff 
shall immediately notify the requestor in writing, briefly stating the 
reason or reasons for the denial.
    (2) The requester may challenge the NRC staff's adverse 
determination by filing a challenge within 5 days of receipt of that 
determination with: (a) The presiding officer designated in this 
proceeding; (b) if no presiding officer has been appointed, the Chief 
Administrative Judge, or if he or she is unavailable, another 
administrative judge, or an administrative law judge with jurisdiction 
pursuant to 10 CFR 2.318(a); or (c) officer if that officer has been 
designated to rule on information access issues.
    H. Review of Grants of Access. A party other than the requester may 
challenge an NRC staff determination granting access to SUNSI whose 
release would harm that party's interest independent of the proceeding. 
Such a challenge must be filed with the Chief Administrative Judge 
within 5 days of the notification by the NRC staff of its grant of 
access.
    If challenges to the NRC staff determinations are filed, these 
procedures give way to the normal process for litigating disputes 
concerning access to information. The availability of interlocutory 
review by the Commission of orders ruling on such NRC staff 
determinations (whether granting or denying access) is governed by 10 
CFR 2.311.\3\
---------------------------------------------------------------------------

    \3\ Requesters should note that the filing requirements of the 
NRC's E-Filing Rule (72 FR 49139; August 28, 2007) apply to appeals 
of NRC staff determinations (because they must be served on a 
presiding officer or the Commission, as applicable), but not to the 
initial SUNSI request submitted to the NRC staff under these 
procedures.
---------------------------------------------------------------------------

    I. The Commission expects that the NRC staff and presiding officers 
(and any other reviewing officers) will consider and resolve requests 
for access to SUNSI, and motions for protective orders, in a timely 
fashion in order to minimize any unnecessary delays in identifying 
those petitioners who have standing and who have propounded contentions 
meeting the specificity and basis requirements in 10 CFR Part 2. 
Attachment 1 to this Order summarizes the general target schedule for 
processing and resolving requests under these procedures.
    It is so ordered.

    Dated at Rockville, Maryland, this 24th day of July 2014.

    For the Commission.
Richard J. Laufer,
Acting Secretary of the Commission.

   Attachment 1--General Target Schedule for Processing and Resolving
Requests for Access to Sensitive Unclassified Non-Safeguards Information
                           in This Proceeding
------------------------------------------------------------------------
        Day                            Event/activity
------------------------------------------------------------------------
0.................  Publication of Federal Register notice of hearing
                     and opportunity to petition for leave to intervene,
                     including order with instructions for access
                     requests.
10................  Deadline for submitting requests for access to
                     Sensitive Unclassified Non-Safeguards Information
                     (SUNSI) with information: Supporting the standing
                     of a potential party identified by name and
                     address; describing the need for the information in
                     order for the potential party to participate
                     meaningfully in an adjudicatory proceeding.
60................  Deadline for submitting petition for intervention
                     containing: (i) Demonstration of standing; and (ii)
                     all contentions whose formulation does not require
                     access to SUNSI (+25 Answers to petition for
                     intervention; +7 petitioner/requestor reply).
20................  U.S. Nuclear Regulatory Commission (NRC) staff
                     informs the requester of the staff's determination
                     whether the request for access provides a
                     reasonable basis to believe standing can be
                     established and shows need for SUNSI. (NRC staff
                     also informs any party to the proceeding whose
                     interest independent of the proceeding would be
                     harmed by the release of the information.) If NRC
                     staff makes the finding of need for SUNSI and
                     likelihood of standing, NRC staff begins document
                     processing (preparation of redactions or review of
                     redacted documents).

[[Page 45495]]

 
25................  If NRC staff finds no ``need'' or no likelihood of
                     standing, the deadline for petitioner/requester to
                     file a motion seeking a ruling to reverse the NRC
                     staff's denial of access; NRC staff files copy of
                     access determination with the presiding officer (or
                     Chief Administrative Judge or other designated
                     officer, as appropriate). If NRC staff finds
                     ``need'' for SUNSI, the deadline for any party to
                     the proceeding whose interest independent of the
                     proceeding would be harmed by the release of the
                     information to file a motion seeking a ruling to
                     reverse the NRC staff's grant of access.
30................  Deadline for NRC staff reply to motions to reverse
                     NRC staff determination(s).
40................  (Receipt +30) If NRC staff finds standing and need
                     for SUNSI, deadline for NRC staff to complete
                     information processing and file motion for
                     Protective Order and draft Non-Disclosure
                     Affidavit. Deadline for applicant/licensee to file
                     Non-Disclosure Agreement for SUNSI.
A.................  If access granted: Issuance of presiding officer or
                     other designated officer decision on motion for
                     protective order for access to sensitive
                     information (including schedule for providing
                     access and submission of contentions) or decision
                     reversing a final adverse determination by the NRC
                     staff.
A + 3.............  Deadline for filing executed Non-Disclosure
                     Affidavits. Access provided to SUNSI consistent
                     with decision issuing the protective order.
A + 28............  Deadline for submission of contentions whose
                     development depends upon access to SUNSI. However,
                     if more than 25 days remain between the
                     petitioner's receipt of (or access to) the
                     information and the deadline for filing all other
                     contentions (as established in the notice of
                     hearing or opportunity for hearing), the petitioner
                     may file its SUNSI contentions by that later
                     deadline.
A + 53............  (Contention receipt +25) Answers to contentions
                     whose development depends upon access to SUNSI.
A + 60............  (Answer receipt +7) Petitioner/Intervenor reply to
                     answers.
>A + 60...........  Decision on contention admission.
------------------------------------------------------------------------

[FR Doc. 2014-17949 Filed 8-4-14; 8:45 am]
BILLING CODE 7590-01-P