Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 17083-17109 [2015-07192]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 80, Number 61 (Tuesday, March 31, 2015)]
[Notices]
[Pages 17083-17109]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2015-07192]


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NUCLEAR REGULATORY COMMISSION

[NRC-2015-0073]


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses and Combined Licenses Involving No Significant 
Hazards Considerations

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of 
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission 
(NRC) is publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued and grants the Commission the authority to issue and make 
immediately effective any 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 March 5, 2015 to March 18, 2015. The last 
biweekly notice was published on March 17, 2015.

DATES: Comments must be filed by April 30, 2015. A request for a 
hearing must be filed by June 1, 2015.

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 https://www.regulations.gov and search for Docket ID NRC-2015-0073. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. 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: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    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: Kay Goldstein, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-1506, email: Kay.Goldstein@nrc.gov.

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2015-0073 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 https://www.regulations.gov and search for Docket ID NRC-2015-0073.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection 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. 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-2015-0073, facility name, unit 
number(s), application date, and subject in your comment submission.
    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 https://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

[[Page 17084]]

submissions to remove such information before making the comment 
submissions available to the public or entering the comment submissions 
into ADAMS.

II. Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses and Combined Licenses and Proposed No Significant 
Hazards Consideration Determination

    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 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 unless the 
Commission finds an imminent danger to the health or safety of the 
public, in which case it will issue an appropriate order or rule under 
10 CFR part 2.

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

[[Page 17085]]

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 hearing.docket@nrc.gov, 
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 https://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 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 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 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's 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'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 https://www.nrc.gov/site-help/e-submittals.html, by email to 
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 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.
Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina
    Date of amendment request: November 24, 2014. A publicly-

[[Page 17086]]

available version is in ADAMS under Accession No. ML14330A327.
    Description of amendment request: The proposed amendments would 
modify the Technical Specifications (TS) to correct non-conservative 
setpoints. Specifically, modify the Allowable Value parameter and the 
Nominal Trip Setpoint for the TS 3.3.2 Table 3.3.2-1, ``Engineered 
Safety Feature Actuation System Instrumentation'' function for 
Auxiliary Feedwater Loss of Offsite Power (Function 6.d.) and for the 
TS 3.3.5 Loss of Voltage function in Surveillance Requirement (SR) 
3.3.5.2. As part of the change, the licensee is also proposing to add 
the applicable footnotes in accordance with TSTF-493, Revision 4, 
``Clarify Application of Setpoint Methodology for LSSS [limiting safety 
system set point] Functions.''
    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 and staff's changes/additions 
are provided in [ ]:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Duke Energy requests NRC review and approval to revise the 
Allowable Value parameter for the Technical Specification (TS) 3.3.2 
Table 3.3.2-1, ``Engineered Safety Feature Actuation System 
Instrumentation'' function for Auxiliary Feedwater Loss of Offsite 
Power (Function 6.d.) and for the TS 3.3.5 Loss of Voltage function 
in Surveillance Requirement (SR) 3.3.5.2 in order to make this 
parameter more restrictive. The existing parameter was determined to 
be non-conservative and this parameter is presently classified as 
Operable But Degraded in the Catawba Corrective Action Program. In 
addition, the Nominal Trip Setpoint parameter for this function is 
being slightly lowered in order to gain additional margin. Finally, 
as part of this License Amendment Request (LAR), applicable 
footnotes are also being added to the affected TS 3.3.2 function in 
accordance with TS Task Force Traveler [(TSTF)] TSTF-493, Revision 
4, ``Clarify Application of Setpoint Methodology for LSSS 
Functions.'' The more restrictive Allowable Value will preclude the 
potential for a double sequencing event to occur under the condition 
of a Loss of Coolant Accident (LOCA) load sequencer actuation with a 
pre-existing degraded voltage condition on the essential buses. 
These proposed changes will not increase the probability of 
occurrence of any design basis accident since the affected function, 
in and of itself, cannot initiate an accident. Should a LOCA occur, 
the proposed changes will ensure that the sequencer operates 
properly in order to mitigate the consequences of the event. 
Appropriate calculations were developed to substantiate the revised 
TS parameters proposed in this LAR. There will be no impact on the 
source term or pathways assumed in accidents previously evaluated. 
No analysis assumptions will be violated and there will be no 
adverse effects on onsite or offsite doses as the result of an 
accident. Adoption of the TSTF-493 footnotes for the respective SRs 
will ensure that the function's channels will continue to behave in 
accordance with safety analysis assumptions and the channel 
performance assumptions in the setpoint methodology.
    Therefore, the proposed amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendments do not change the methods governing 
normal plant operation; nor are the methods utilized to respond to 
plant transients altered. In addition, the proposed changes to the 
affected TS parameters and the adoption of the TSTF-493 footnotes 
will not create the potential for any new initiating events or 
transients to occur in the actual physical plant.
    Therefore, the proposed amendments do 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.
    Margin of safety is related to the confidence in the ability of 
the fission product barriers to perform their design functions 
during and following an accident. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. 
The proposed changes will assure the acceptable operation of the 
affected function under all postulated transient and accident 
conditions. This will ensure that all applicable design and safety 
limits are satisfied such that the fission product barriers will 
continue to perform their design functions.
    Therefore, the proposed amendments do not involve a significant 
reduction in a margin of safety.
    Based on the preceding discussion, Duke Energy concludes that 
the proposed amendments do not involve a significant hazards 
consideration under the standards set forth in 10 CFR 50.92(c), and, 
accordingly, a finding of ``no significant hazards consideration'' 
is justified.

    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: Lara S. Nichols, Associate General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Robert J. Pascarelli.
Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, 
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South 
Carolina
    Date of amendment request: March 14, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14078A037.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TS) for the Inservice Testing Program to 
reflect the current edition of the American Society of Mechanical 
Engineers (ASME) Code that is referenced in 10 CFR 50.55a(b).
    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 corrects a typographical error in TS 5.5.8, 
``Reactor Coolant Pump Flywheel Inspection Program,'' and revises TS 
5.5.9, ``lnservice Testing Program,'' for consistency with the 
requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing 
of pumps and valves which are classified as ASME Code Class 1, Class 
2 and Class 3. The proposed change incorporates revisions to the 
ASME Code that result in a net improvement in the measures for 
testing pumps and valves.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. The proposed change does not involve the addition or removal 
of any equipment, or any design changes to the facility.
    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 corrects a typographical error in TS 5.5.8, 
``Reactor Coolant Pump Flywheel Inspection Program,'' and revises TS 
5.5.9, ``lnservice Testing Program,'' for consistency with the 
requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing 
of pumps and valves which are classified as ASME Code Class 1, Class 
2 and Class 3. The proposed change incorporates revisions to the 
ASME Code that result in a net improvement in the measures for 
testing pumps and valves.
    The proposed change does not involve a modification to the 
physical configuration of

[[Page 17087]]

the plant (i.e., no new equipment will be installed), nor does it 
involve a change in the methods governing normal plant operation. 
The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released offsite and there is no increase in individual or 
cumulative occupational exposure.
    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 corrects a typographical error in TS 5.5.8, 
``Reactor Coolant Pump Flywheel Inspection Program,'' and revises TS 
5.5.9, ``lnservice Testing Program,'' for consistency with the 
requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing 
of pumps and valves which are classified as ASME Code Class 1, Class 
2 and Class 3. The proposed change incorporates revisions to the 
ASME Code that result in a net improvement in the measures for 
testing pumps and valves. The safety function of the affected pumps 
and valves will be maintained. 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: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Robert J. Pascarelli.
Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne 
County, Mississippi
    Date of amendment request: November 21, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14325A520.
    Description of amendment request: The amendment would change the 
GGNS Technical Specification (TS) 2.1.1, ``Reactor Core SLs [Safety 
Limits].'' Specifically, the change would revise the Minimum Critical 
Power Ratio (MCPR) SL stated in TS 2.1.1.2 for two-loop operation from 
greater than or equal to (>=) 1.11 to >= 1.15. Additionally, the change 
would revise the MCPR SL stated in TS 2.1.1.2 for single-loop operation 
from >= 1.14 to >= 1.15.
    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 Bases to TS 2.1.1.2 states that: ``The MCPR SL ensures 
sufficient conservatism in the operating MCPR limit that, in the 
event of an AOO [Anticipated Operational Occurrence] from the 
limiting condition of operation, at least 99.9% of the fuel rods in 
the core would be expected to avoid boiling transition.
    This condition is met in that the GGNS Cycle 20 (C20) MCPR SL 
evaluation was performed in accordance with Reference 4 [NEDE-24011-
P-A, ``General Electric Standard Application for Reactor Fuel 
(GESTAR-II'')]. The resulting values continue to ensure the 
conservatism described in the Bases to TS 2.1.1.2. The proposed 
changes also continue to ensure sufficient conservatism in the 
operating MCPR limit. The MCPR operating limits are presented and 
controlled in accordance with the GGNS Core Operating Limits Report 
(COLR).
    The requested Technical Specification change does not involve 
any plant modifications or operational changes that could affect 
system reliability or performance or that could affect the 
probability of operator error. The requested change does not affect 
any postulated accident precursors, any accident mitigating systems, 
or introduce any new accident initiation mechanisms.
    Therefore, the proposed change to increase the MCPR SL values 
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 any new modes of operation, 
any changes to setpoints, or any plant modifications. The proposed 
change to the MCPR SL accounts for requirements specified in the NRC 
Safety Evaluation limitations and conditions associated with NEDC-
33173P [``Applicability of GE Methods to Expanded Operating 
Domains''] and NEDC-33006P [``Licensing Topical Report--General 
Electric Boiling Water Reactor Maximum Extended Load Line Limit 
Analysis Plus'']. Compliance with the criterion for incipient 
boiling transition continues to be ensured. The core operating 
limits will continue to be developed using NRC approved methods. The 
proposed [MCPR SL] does not result in the creation of any new 
precursors to an accident.
    Therefore, the proposed change does not create 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 MCPR SLs have been evaluated in accordance with Global 
Nuclear Fuels NRC-approved cycle-specific safety limit methodology 
to ensure that during normal operation and during AOO's, at least 
99.9% of the fuel rods in the core are not expected to experience 
transition boiling. The proposed change to the [MCPR SL] accounts 
for requirements specified in the NRC Safety Evaluation limitations 
and conditions associated with NEDC-33173P and NEDC-33006P, which 
result in additional margin above that specified in the TS Bases.
    Therefore, the proposed change to the MCPR SL 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 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Meena K. Khanna.
Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1 (GGNS), Claiborne 
County, Mississippi
    Date of amendment request: November 21, 2014, as supplemented by 
letter dated February 18, 2015. Publicly-available versions are in 
ADAMS under Accession Nos. ML14325A752 and ML15049A536, respectively.
    Description of amendment request: The proposed amendment would 
revise GGNS's license basis to adopt a single fluence methodology.
    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 adopts a single flux methodology. While 
Chapter 15, Accident Analysis, of the Standard Review Plan (NUREG-
0800, Standard Review Plan for the Review of Safety Analysis Reports 
for Nuclear Power Plants) assumes the pressure vessel does not fail, 
the flux methodology is not an initiator to any accident previously 
evaluated. Accordingly, the proposed change

[[Page 17088]]

to the adoption of the flux methodology has no effect on the 
probability 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 adopts a flux methodology. The change does 
not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operations. The change does not alter 
assumptions made in the safety analysis regarding fluence.
    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 adopts a single fluence methodology. The 
proposed change does not alter the manner in which safety limits, 
limiting safety system settings or limiting conditions for operation 
are determined. The proposed change ensures that the methodology 
used for fluence is in compliance with RG 1.190 requirements.
    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: Joseph A. Aluise, Associate General 
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New 
Orleans, Louisiana 70113.
    NRC Branch Chief: Meena K. Khanna.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois
    Date of amendment request: August 19, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14231A902.
    Description of amendment request: The proposed amendment would 
increase the technical specification (TS) surveillance requirement (SR) 
3.7.9.2 allowable temperature to less than or equal to 
102[emsp14][deg]F [degree Fahrenheit].
    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 likelihood of a malfunction of any systems, structures or 
components (SSCs) supported by the UHS [ultimate heat sink] is not 
significantly increased by increasing the allowable Ultimate Heat 
Sink (UHS) temperature from <=100[emsp14][deg]F to 
<=102[emsp14][deg]F. The UHS provides a heat sink for process and 
operating heat from safety related components during a transient or 
accident, as well as during normal operation. The proposed change 
does not make any physical changes to any plant SSCs, nor does it 
alter any of the assumptions or conditions upon which the UHS is 
designed. The UHS is not an initiator of any analyzed accident. All 
equipment supported by the UHS has been evaluated to demonstrate 
that their performance and operation remains as described in the 
UFSAR [updated final safety analysis report] with no increase in 
probability of failure or malfunction.
    The SSCs credited to mitigate the consequences of postulated 
design basis accidents remain capable of performing their design 
basis function. The change in maximum UHS temperature has been 
evaluated using the UFSAR described methods to demonstrate that the 
UHS remains capable of removing normal operating and post-accident 
heat. The change in UHS temperature and resulting containment 
response following a postulated design basis accident has been 
demonstrated to not be impacted. Additionally, all the UHS supported 
equipment, credited in the accident analysis to mitigate an 
accident, has been shown to continue to perform their design 
function as described in the UFSAR.
    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 does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed change does not introduce any new modes of plant 
operation, change the design function of any SSC, change the mode of 
operation of any SSC, or change any actions required when the TS 
limit is exceeded. There are no new equipment failure modes or 
malfunctions created as affected SSCs continue to operate in the 
same manner as previously evaluated and have been evaluated to 
perform as designed at the increased UHS temperature and as assumed 
in the accident analysis. Additionally, accident initiators remain 
as described in the UFSAR and no new accident initiators are 
postulated as a result of the increase in UHS temperature.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the Proposed Change Involve a Significant Reduction in a 
Margin of Safety?
    Response: No.
    The proposed change continues to ensure that the maximum 
temperature of the cooling water supplied to the plant SSCs during a 
UHS design basis event remains within the evaluated equipment limits 
and capabilities assumed in the accident analysis. The proposed 
change does not result in any changes to plant equipment function, 
including setpoints and actuations. All equipment will function as 
designed in the plant safety analysis without any physical 
modifications. The proposed change does not alter a limiting 
condition for operation, limiting safety system setting, or safety 
limit specified in the Technical Specifications.
    The proposed change does not adversely impact the UHS inventory 
required to be available for the UFSAR described design basis 
accident involving the worst case 30-day period including losses for 
evaporation and seepage to support safe shutdown and cooldown of 
both Braidwood Station units. Additionally, the structural integrity 
of the UHS is not impacted and remains acceptable following the 
change, thereby ensuring that the assumptions for both UHS 
temperature and inventory remain valid.
    Therefore, since there is no adverse impact of this change on 
the Braidwood Station safety analysis, there is no reduction in the 
margin of safety of the plant.

    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: Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Travis L. Tate
Exelon Generation Company, LLC (EGC), Docket Nos. STN 50-454 and STN 
50-455, Byron Station, Units 1 and 2, Ogle County, Illinois
    Date of amendment request: November 24, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14328A800.
    Description of amendment request: The proposed amendment would 
revise Condition I and surveillance requirement (SR) 3.7.9.3 associated 
with technical specification (TS) Section 3.7.9, ``Ultimate Heat Sink 
(UHS),'' to reflect the current design basis flood level.
    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

[[Page 17089]]

consideration, which is presented below:

    EGC has evaluated whether or not a significant hazards 
consideration is involved with the proposed amendment by focusing on 
the three standards set forth in 10 CFR 50.92(c), ``Issuance of 
amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to revise TS 3.7.9, Condition I and SR 
3.7.9.3 will ensure the operability of the SX [service water] makeup 
pumps to meet TS 3.7.9 LCO [Limiting Condition for Operation] 
requirement. The proposed change does not result in any physical 
changes to safety related structures, systems, or components. The 
probability of a flood at the river screen house (RSH) is unchanged. 
Since the UHS itself is not an accident initiator, the proposed 
change does not impact the initiators or assumptions of analyzed 
accidents, nor do they impact the mitigation of accidents or 
transient events. Consequently, the proposed change does not 
increase the probability of occurrence for any accident previously 
evaluated.
    The proposed change will ensure that actions to verify 
operability of the deep well pumps will be taken prior to the 
potential for the SX makeup pumps to be adversely affected by the 
combined event flood high river level. Therefore, the UHS will be 
capable of performing its functions to mitigate accidents by serving 
as the heat sink for safety related equipment. Thus, the proposed 
change does not 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change to revise TS 3.7.9, Condition I and SR 
3.7.9.3 does not change the design function or operation of the SX 
makeup pumps. The proposed change does not change or introduce the 
possibility of any new or different type of equipment, modes of 
system operation, failure mechanisms, malfunctions, or accident 
initiators. The proposed change to lower the river level value at 
which action is taken to verify basin levels and deep well pumps are 
ready to perform the UHS makeup function in the place of the SX 
makeup pumps will not affect the operation or function of the UHS or 
the deep well pumps.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to revise TS 3.7.9, Condition I and SR 
3.7.9.3 reestablishes the margin between the design bases combined 
event flood level and TS 3.7.9, Condition I action level for high 
river level. The proposed change will ensure the operability of the 
SX makeup pumps to meet TS 3.7.9 LCO and do not affect the ability 
of the SX makeup pumps to provide the safety related source makeup 
to the UHS.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, EGC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and accordingly, a finding 
of no significant hazards consideration is justified.

    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: Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
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. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois
    Date of amendment request: December 22, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14357A085.
    Description of amendment request: The proposed amendment modifies 
the technical specifications (TSs) to add a new Limiting Condition for 
Operation (LCO) 3.10.8 to specifically permit inservice leakage and 
hydrostatic testing at reactor coolant system (RCS) temperatures 
greater than the average reactor coolant temperature for MODE 4 with 
the reactor shutdown. In addition, the proposed amendment includes an 
expanded scope of LCO 3.10.8 consistent with the NRC-approved Revision 
0 of Technical Specification Task Force (TSTF) Improved Standard 
Technical Specification Change Traveler, TSTF-484, ``Use of TS 3.10.1 
for Scram Time Testing Activities'' available in ADAMS under Accession 
No. ML062990425.
    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:

    EGC [Exelon Generation Company] has evaluated the proposed 
changes, using the criteria in 10 CFR 50.92, and has determined that 
the proposed changes do not involve a significant hazards 
consideration. The following information is provided to support a 
finding of no significant hazards consideration.
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes will not result in a significant change in 
the stored energy in the reactor vessel during the performance of 
the testing. The probability of an accident is not significantly 
increased because the proposed changes will not alter the method by 
which inservice leakage and hydrostatic testing is performed or 
significantly change the temperatures and pressures achieved to 
perform the test.
    The consequences of previously evaluated accidents are not 
significantly increased because the required testing conditions 
provide adequate assurance that the consequences of a steam leak 
will be conservatively bounded by the consequences of the postulated 
main system line break outside of primary containment. Under these 
proposed changes, the secondary containment, standby gas treatment 
system, and associated initiation instrumentation are required to be 
operable during the performance of inservice leakage and hydrostatic 
testing and would be capable of mitigating any airborne 
radioactivity or steam leaks that could occur. In addition, the 
required Emergency Core Cooling subsystems will be more than 
adequate to ensure that a significant increase in consequences will 
not occur by ensuring that the potential for failed fuel and a 
subsequent increase in coolant activity above Technical 
Specification limits are minimized.
    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.
    As the accumulated neutron fluence on the reactor vessel 
increases, the Pressure-Temperature Limits in TS 3.4.9 for DNPS 
[Dresden Nuclear Power Station] and QCNPS [Quad Cities Nuclear Power 
Station and TS [technical specification] 3.4.11 for LSCS [LaSalle 
County Station] may eventually require that inservice leakage and 
hydrostatic testing be conducted at RCS [reactor coolant system] 
temperatures greater than the average reactor coolant temperature 
for MODE 4 with the reactor shutdown. However, even with the 
required minimum reactor coolant temperatures less than or equal to 
the average reactor coolant temperature for MODE 4 with the reactor 
shutdown, maintaining RCS

[[Page 17090]]

temperatures within a small band during testing can be impractical. 
The proposed changes will not result in a significant change in the 
stored energy in the reactor vessel during the performance of the 
testing nor will it alter the way inservice leakage and hydrostatic 
testing is performed or significantly change the temperatures and 
pressures achieved to perform the testing.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes and additions result in increased system 
operability requirements above those that currently exist during the 
performance of inservice leakage and hydrostatic testing. The 
incremental increase in stored energy in the vessel during testing 
will be conservatively bounded by the consequences of the postulated 
main steam line break outside of primary containment and analyzed 
margins of safety are unchanged.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    EGC has reviewed the no significant hazards determination 
published on August 21, 2006 (71 FR 48561) [for Technical 
Specification Task Force traveler TSTF-484]. The no significant 
hazards determination was made available on October 27, 2006 (71 FR 
63050) as part of the CLIIP [Consolidated Line Item Improvement 
Process] Notice of Availability. EGC has concluded that the 
determination presented in the notice is applicable to DNPS, Units 2 
and 3; LSCS, Units 1 and 2; and QCNPS, Units 1 and 2; and the 
determination is hereby incorporated by reference to satisfy the 
requirements of 10 CFR 50.91(a).

    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: Bradley Fewell, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: Travis L. Tate.
Exelon Generation Company, LLC, Docket No. 50-373 and 50-374, LaSalle 
County Station (LSCS), Units 1 and 2, LaSalle County, Illinois
    Date of Amendment Request: January 12, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15012A544.
    Description of amendment request: The proposed amendment would 
delete the limiting condition for operation (LCO) Note for Technical 
Specification (TS) Section 3.5.1, ``ECCS [emergency core cooling 
system]--Operating.'' The current Note allows the licensee to consider 
the low pressure coolant injection (LPCI) subsystem associated with the 
residual heat removal (RHR) system to be OPERABLE under specified 
conditions.
    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.
    No physical changes to the facility will occur as a result of 
this proposed amendment. The proposed change will not alter the 
physical design. Current TS note could make LSCS susceptible to 
potential water hammer in the RHR system if in the SDC [shutdown 
cooling] Mode of RHR in Mode 3 when swapping from the SDC to LPCI 
mode of RHR. The proposed LAR [license amendment request] will 
eliminate the risk for cavitation of the pump and voiding in the 
suction piping, thereby avoiding potential to damage the RHR system, 
including water hammer.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not alter the physical design, safety 
limits, or safety analysis assumptions associated with the operation 
of the plant. Accordingly, the change does not introduce any new 
accident initiators, nor does it reduce or adversely affect the 
capabilities of any plant structure, system, or component to perform 
their safety function. Deletion of the TS note is appropriate 
because current TSs could put the plant at risk for potential 
cavitation of the pump and voiding in the suction piping, resulting 
in potential to damage the RHR system, including water hammer.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change conforms to NRC regulatory guidance 
regarding the content of plant Technical Specifications. The 
proposed change does not alter the physical design, safety limits, 
or safety analysis assumptions associated with the operation of the 
plant.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above evaluation, EGC [Exelon Generation Company, 
LLC] concludes that the proposed amendment does not involve a 
significant hazards consideration under the standards set forth in 
10 CFR 50.92(c), and, according a finding of no significant hazards 
consideration is justified.

    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: Tamra Domeyer, Exelon Generation Company, 
LLC, 4300 Winfield Road, Warrenville, IL, 60555.
    Branch Chief: Travis L. Tate.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio
    Date of amendment request: December 31, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14365A080.
    Description of amendment request: The proposed amendment would 
revise the frequency for the technical specification surveillance to 
verify that each containment spray system nozzle is unobstructed from a 
frequency of 10 years to an event-based frequency.
    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 containment spray system and its spray nozzles are not 
accident initiators and therefore the proposed change does not 
involve a significant increase in the probability of an accident. 
The revised surveillance requirement will require event-based 
frequency verification in lieu of a fixed frequency verification. 
The proposed change does not have a detrimental impact on the 
integrity of any plant structure, system, or component that may 
initiate an analyzed event. The proposed change will not alter the 
operation or otherwise increase the failure probability of any plant 
equipment that can initiate an analyzed accident. Because the system 
will continue to be available to perform its accident mitigation 
function, the consequences of accidents previously evaluated are not 
significantly increased.
    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?

[[Page 17091]]

    Response: No.
    The proposed change will not physically alter the plant (no new 
or different type of equipment will be installed) or change the 
methods governing normal plant operation. The proposed change does 
not introduce new accident initiators or impact assumptions made in 
the safety analysis. Testing requirements continue to demonstrate 
that the limiting conditions for operation are met and the system 
components are functional.
    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 safety function of the CSS [containment spray system] is to 
spray water into the containment atmosphere in the event of a loss-
of-coolant accident to prevent containment pressure from exceeding 
the design value and to remove fission products from the containment 
atmosphere.
    The CSS is not susceptible to corrosion-induced obstruction or 
obstruction from sources external to the system. Maintenance 
activities that unexpectedly introduce unretrievable foreign 
material into the system would require subsequent verification to 
ensure there is no nozzle blockage. The spray header nozzles are 
expected to remain unblocked and available in the event that a 
safety function is required. Therefore, the capacity of the system 
would remain unaffected. The proposed change does not relax any 
criteria used to establish safety limits and will not relax any 
safety system settings. The safety analysis acceptance criteria are 
not affected by this change.
    Therefore, the proposed change does not involve a significant 
reduction in the 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: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Travis L. Tate.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio
    Date of amendment request: December 19, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14353A349.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to adopt performance-based 
Type C testing for the reactor containment, which would allow for 
extended test intervals for Type C valves up to 75 months, and corrects 
an editorial issue in the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment adopts the NRG-accepted guidelines of 
[Nuclear Energy Institute] NEI 94-01, Revision 3-A, ``Industry 
Guideline for Implementing Performance-Based Option of 10 CFR part 
50, Appendix J,'' for [Davis-Besse Nuclear Power Station] DBNPS 
performance-based Type C containment isolation valve testing. 
Revision 3-A of NEI 94-01 allows, based on previous valve leak test 
performance, an extension of Type C containment isolation valve leak 
test intervals. Since the change involves only performance-based 
Type C testing, the proposed amendment does not involve either a 
physical change to the plant or a change in the manner in which the 
plant is operated or controlled.
    Implementation of these guidelines continues to provide adequate 
assurance that during design basis accidents, the components of the 
primary containment system will limit leakage rates to less than the 
values assumed in the plant safety analyses.
    The proposed amendment will not change the leakage rate 
acceptance requirements. As such, the containment will continue to 
perform its design function as a barrier to fission product 
releases.
    Therefore, the proposed amendment does not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment to revise the extended frequency 
performance-based Type C testing program does not change the design 
or operation of structures, systems, or components of the plant.
    The proposed amendment would continue to ensure containment 
operability and would ensure operation within the bounds of existing 
accident analyses. There are no accident initiators created or 
affected by the proposed amendment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed amendment to revise the extended frequency 
performance-based Type C testing program does not affect plant 
operations, design functions, or any analysis that verifies the 
capability of a structure, system, or component of the plant to 
perform a design function. In addition, this change does not affect 
safety limits, limiting safety system setpoints, or limiting 
conditions for operation. The specific requirements and conditions 
of the Technical Specification Containment Leakage Rate Testing 
Program exist to ensure that the degree of containment structural 
integrity and leak-tightness that is considered in the plant safety 
analysis is maintained.
    The overall containment leak rate limit specified by Technical 
Specifications is maintained, thus ensuring the margin of safety in 
the plant safety analysis is maintained. The design, operation, 
testing methods, and acceptance criteria for Type A, Type B, and 
Type C containment leakage tests specified in applicable codes and 
standards would continue to be met with the acceptance of this 
proposed change, since these are not affected by this revision to 
the performance-based containment testing program.
    Therefore, 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: Travis L. Tate.
Indiana Michigan Power Company (IandM), Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
    Date of amendment request: November 14, 2014, as supplemented by a 
letter dated February 12, 2015. Publicly-available versions are in 
ADAMS under Accession Nos. ML14324A209, and ML15050A247, respectively.)
    Description of amendment request: The proposed amendments would 
replace the current Donald C. Cook Nuclear Plant (CNP) Units 1 and 2 
technical specifications (TSs) limit on reactor coolant system (RCS) 
gross specific activity with a new limit on RCS noble gas specific 
activity. The noble gas specific activity limit would be based on a new 
DOSE EQUIVALENT XE-133 definition that would replace the current E-Bar 
average disintegration energy definition. In addition, the current DOSE 
EQUIVALENT I-131 definition would be revised to allow the use of 
additional thyroid dose conversion factors. The proposed RCS specific 
activity changes are consistent with NRC-approved Industry Technical

[[Page 17092]]

Specification Task Force (TSTF) Standard Technical Specification change 
traveler, TSTF-490, Revision 0, ``Deletion of E-Bar Definition and 
Revision to Reactor Coolant System Specific Activity Technical 
Specification,'' with deviations. Additionally, the proposed amendments 
would revise the CNP Units 1 and 2 licensing basis and TSs to adopt the 
alternative source term (AST) as allowed in 10 CFR 50.67. The proposed 
amendments represent full implementation of the AST as described in the 
NRC's Regulatory Guide 1.183, ``Alternative Radiological Source Terms 
for Evaluating Design Basis Accidents at Nuclear Power Reactors,'' 
Revision 0.
    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. The licensee concluded that the no significant hazards 
consideration determination published on March 19, 2007 (72 FR 12838), 
``Notice of Availability of the Model Safety Evaluation,'' is 
applicable. This determination is presented below, along with the 
licensee's analysis of the implementation of the AST.

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    Reactor coolant specific activity is not an initiator for any 
accident previously evaluated. The Completion Time when primary 
coolant gross activity is not within limit is not an initiator for 
any accident previously evaluated. The current variable limit on 
primary coolant iodine concentration is not an initiator to any 
accident previously evaluated. As a result, the proposed change does 
not significantly increase the probability of an accident. The 
proposed change will limit primary coolant noble gases to 
concentrations consistent with the accident analyses. The proposed 
change to the Completion Time has no impact on the consequences of 
any design basis accident since the consequences of an accident 
during the extended Completion Time are the same as the consequences 
of an accident during the Completion Time. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased.
    There are no physical changes to the plant being introduced by 
the proposed changes to the accident source term. Implementation of 
AST and the associated proposed TS changes and new atmospheric 
dispersion factors have no impact on the probability for initiation 
of any DBAs [Design Basis Accidents]. Once the occurrence of an 
accident has been postulated, the new accident source term and 
atmospheric dispersion factors are an input to analyses that 
evaluate the radiological consequences. The proposed changes do not 
involve a revision to the design or manner in which the facility is 
operated that could increase the probability of an accident 
previously evaluated in Chapter 14 of the UFSAR.
    Based on the AST analyses, there are no proposed changes to 
performance requirements and no proposed revision to the parameters 
or conditions that could contribute to the initiation of an accident 
previously discussed in Chapter 14 of the UFSAR. Plant-specific 
radiological analyses have been performed using the AST methodology 
and new X/Qs have been established. Based on the results of these 
analyses, it has been demonstrated that the CR [control room] and 
off-site dose consequences of the limiting events considered in the 
analyses meet the regulatory guidance provided for use with the AST, 
and the doses are within the limits established by 10 CFR 50.67.
    Therefore, it is concluded that the proposed amendment does not 
involve a significant increase in the probability or the 
consequences of an accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated.
    The proposed change in specific activity limits does not alter 
any physical part of the plant nor does it affect any plant 
operating parameter. The change does not create the potential for a 
new or different kind of accident from any previously calculated.
    No new modes of operation are introduced by the proposed 
changes. The proposed changes will not create any failure mode not 
bounded by previously evaluated accidents. Implementation of AST and 
the associated proposed TS changes and new X/Qs have no impact to 
the initiation of any DBAs. These changes do not affect the design 
function or modes of operation of structures, systems and components 
in the facility prior to a postulated accident. Since structures, 
systems and components are operated no differently after the AST 
implementation, no new failure modes are created by this proposed 
change. The alternative source term change itself does not have the 
capability to initiate accidents.
    Consequently, the proposed changes do 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.
    The proposed change revises the limits on noble gas 
radioactivity in the primary coolant. The proposed change is 
consistent with the assumptions in the safety analyses and will 
ensure the monitored values protect the initial assumptions in the 
safety analyses.
    The AST analyses have been performed using approved 
methodologies to ensure that analyzed events are bounding and safety 
margin has not been reduced. Also, new X/Qs, which are based on site 
specific meteorological data, were calculated in accordance with the 
guidance of RG 1.194 to utilize more recent data and improved 
calculational methodologies. The dose consequences of these limiting 
events are within the acceptance criteria presented in 10 CFR 50.67. 
Thus, by meeting the applicable regulatory limits for AST, there is 
no significant reduction in a margin of safety. Therefore, because 
the proposed changes continue to result in dose consequences within 
the applicable regulatory limits, the proposed amendment does not 
involve a significant reduction in margin of safety.

    The NRC staff has reviewed the 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 
requested involve no significant hazards consideration.
    Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel, 
One Cook Place, Bridgman, MI 49106.
    NRC Branch Chief: David L. Pelton.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446, 
Comanche Peak Nuclear Power Plant (CPNPP), Units 1 and 2, Somervell 
County, Texas
    Date of amendment request: January 28, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15036A032.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 5.5.16, ``Containment Leakage Rate Testing 
Program,'' for CPNPP, Units 1 and 2, to allow an increase in the 10 CFR 
part 50, appendix J, ``Primary Reactor Containment Leakage Testing for 
Water-Cooled Power Reactors,'' Type A Integrated Leak Rate Test (ILRT) 
interval from a 10-year frequency to a maximum of 15 years and the 
extension of the containment isolation valves leakage Type C tests from 
its current 60-month frequency to 75 months in accordance with Nuclear 
Energy Institute (NEI) 94-01, Revision 3-A, ``Industry Guidance for 
Implementing Performance Based Option of 10 CFR part 50, appendix J,'' 
July 2012 (ADAMS Accession No. ML12221A202), and conditions and 
limitations specified in NEI 94-01, Revision 2-A, ``Industry Guidance 
for Implementing Performance Based Option of 10 CFR part 50, appendix 
J,'' October 2008 (ADAMS Accession No. ML100620847), in addition to 
limitations and conditions of NEI 94-01, Revision 3-A. The proposed 
change would also delete the listing of one-time exceptions previously 
granted to ILRT frequencies.
    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:


[[Page 17093]]


    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment to the TS involves the extension of the 
CPNPP, Units 1 and 2 Type A containment test interval to 15 years 
and the extension of the Type C test interval to 75 months. The 
current Type A test interval of 120 months (10 years) would be 
extended on a permanent basis to no longer than 15 years from the 
last Type A test. The current Type C test interval of 60 months for 
selected components would be extended on a performance basis to no 
longer than 75 months. Extensions of up to nine months (total 
maximum interval of 84 months for Type C tests) are permissible only 
for non-routine emergent conditions. The proposed extension does not 
involve either a physical change to the plant or a change in the 
manner in which the plant is operated or controlled. The containment 
is designed to provide an essentially leak tight barrier against the 
uncontrolled release of radioactivity to the environment for 
postulated accidents. The containment and the testing requirements 
invoked to periodically demonstrate the integrity of the containment 
exist to ensure the plant's ability to mitigate the consequences of 
an accident, and do not involve the prevention or identification of 
any precursors of an accident. The change in dose risk for changing 
the Type A test frequency from three-per-ten years to once-per-
fifteen-years, measured as an increase to the total integrated dose 
risk for all internal events accident sequences for CPNPP, of 1.00E-
02 person rem/yr [roentgen equivalent man per year] to 6.51 person-
rem/yr for Unit 1 and 6.53 person-rem/yr for Unit 2 using the EPRI 
[Energy Power Research Institute] guidance with the base case 
corrosion included. Therefore, this proposed extension does not 
involve a significant increase in the probability of an accident 
previously evaluated.
    As documented in NUREG-1493 [, ``Performance-Based Containment 
Leak-Test Program: Draft Report for Comment,'' January 1995 (not 
publicly available)], Type B and C tests have identified a very 
large percentage of containment leakage paths, and the percentage of 
containment leakage paths that are detected only by Type A testing 
is very small. The CPNPP, Units 1 and 2 Type A test history supports 
this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and; (2) time based. Activity based failure mechanisms are defined 
as degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with ASME [American Society of Mechanical Engineers] 
Section XI, the Maintenance Rule, and TS requirements serve to 
provide a high degree of assurance that the containment would not 
degrade in a manner that is detectable only by a Type A test. Based 
on the above, the proposed extensions do not significantly increase 
the consequences of an accident previously evaluated.
    The proposed amendment also deletes exceptions previously 
granted to allow one-time extensions of the ILRT test frequency for 
both Units 1 and 2. These exceptions were for activities that have 
already taken place so their deletion is solely an administrative 
action that has no effect on any component and no impact on how the 
units are operated.
    Therefore, the proposed change does not result in 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 amendment to the TS involves the extension of the 
CPNPP, Unit 1 and 2 Type A containment test interval to 15 years and 
the extension of the Type C test interval to 75 months. The 
containment and the testing requirements to periodically demonstrate 
the integrity of the containment exist to ensure the plant's ability 
to mitigate the consequences of an accident do not involve any 
accident precursors or initiators. The proposed change does not 
involve a physical change to the plant (i.e., no new or different 
type of equipment will be installed) or a change to the manner in 
which the plant is operated or controlled.
    The proposed amendment also deletes exceptions previously 
granted to allow one-time extensions of the ILRT test frequency for 
both Units 1 and 2. These exceptions were for activities that would 
have already taken place by the time this amendment is approved; 
therefore, their deletion is solely an administrative action that 
does not result in any change in how the units are operated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment to TS 5.5.16 involves the extension of 
the CPNPP, Units 1 and 2 Type A containment test interval to 15 
years and the extension of the Type C test interval to 75 months for 
selected components. This amendment does not alter the manner in 
which safety limits, limiting safety system set points, or limiting 
conditions for operation are determined. The specific requirements 
and conditions of the TS Containment Leak Rate Testing Program exist 
to ensure that the degree of containment structural integrity and 
leak-tightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by TS 
is maintained.
    The proposed change involves only the extension of the interval 
between Type A containment leak rate tests and Type C tests for 
CPNPP, Units 1 and 2. The proposed surveillance interval extension 
is bounded by the 15-year ILRT Interval and the 75-month Type C test 
interval currently authorized within NEI 94-01, Revision 3-A. 
Industry experience supports the conclusion that Type B and C 
testing detects a large percentage of containment leakage paths and 
that the percentage of containment leakage paths that are detected 
only by Type A testing is small. The containment inspections 
performed in accordance with ASME Section Xl, TS and the Maintenance 
Rule serve to provide a high degree of assurance that the 
containment would not degrade in a manner that is detectable only by 
Type A testing. The combination of these factors ensures that the 
margin of safety in the plant safety analysis is maintained. The 
design, operation, testing methods and acceptance criteria for Type 
A, B, and C containment leakage tests specified in applicable codes 
and standards would continue to be met, with the acceptance of this 
proposed change, since these are not affected by changes to the Type 
A and Type C test intervals.
    The proposed amendment also deletes exceptions previously 
granted to allow one-time extensions of the ILRT test frequency for 
both Units 1 and 2. These exceptions were for activities that would 
have already taken place by the time this amendment is approved; 
therefore, their deletion is solely an administrative action and 
does not change how the units are operated and maintained.
    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: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1111 Pennsylvania Avenue NW., Washington, DC 20004.
    NRC Branch Chief: Michael T. Markley.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028, 
Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield 
County, South Carolina
    Date of amendment request: December 4, 2014. A publicly-available 
version is in ADAMs under Accession No. ML14339A637.
    Description of amendment request: The proposed change would amend 
Combined License (COL) Nos. NPF-93 and NPF-94 for the Virgil C. Summer 
Nuclear Station (VCSNS) Units 2 and 3 by changing the structure and 
layout of various areas of the annex building. The proposed amendment 
requires changes to the Updated Final Safety Analysis Report (UFSAR) in 
the form of departures from the incorporated plant-

[[Page 17094]]

specific Design Control Document (DCD) Tier 2 information and involves 
changes to related plant-specific Tier 2* and Tier 1 information, with 
corresponding changes to the associated COL Appendix C information.
    Because, this proposed change requires a departure from Tier 1 
information in the Westinghouse Electric Company's Advanced Passive 
1000 DCD, the licensee also requested an exemption from the 
requirements of the Generic DCD Tier 1 in accordance with 10 CFR 
52.63(b)(1).
    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 proposed additions of a new nonsafety-related battery, 
battery room and battery equipment room, the room height increase, 
the floor thickness changes, the relocation of a non-structural 
internal wall, and the associated wall, room and corridor changes 
within the annex building do not adversely affect the fire loading 
analysis durations of the affected fire zones and areas (i.e., the 
calculated fire durations remain less than their design values). 
Thus, the fire loads analysis is not adversely affected (i.e., 
analysis results remain acceptable). The safe shutdown fire analysis 
is not affected. The proposed changes to the structural 
configuration, including anticipated equipment loading, room height, 
and floor thickness are accounted for in the updated structural 
configuration model that was used to analyze the Annex Building for 
safe shutdown earthquake (SSE) and other design loads and load 
combinations, thus the structural analysis is not adversely 
affected. The structural analysis description and results in the 
UFSAR are unchanged. The relocated internal Annex Building wall is 
non-structural, thus this change does not affect the structural 
analyses for the Annex Building. The proposed changes do not involve 
any accident initiating event or component failure, thus the 
probabilities of the accidents previously evaluated are not 
affected. The rooms affected by the proposed changes do not contain 
or interface with safety-related equipment, thus the proposed 
changes would not affect any safety-related equipment or accident 
mitigating function. The radioactive material source terms and 
release paths used in the safety analyses are unchanged, thus the 
radiological releases in the accident analyses are not affected.
    With the conversion of an annex building room to a battery room, 
the building volume serviced by nuclear island nonradioactive 
ventilation system decreases by approximate five percent. This 
reduced volume is used in the post-accident main control room dose 
portion of the UFSAR LOCA radiological analysis. However, the volume 
decrease is not sufficient to change the calculated main control 
room dose reported in the UFSAR, and control room habitability is 
not affected.
    Therefore, 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 accident previously 
evaluated?
    Response: No.
    The proposed additions of a new nonsafety-related battery, 
battery room and battery equipment room, the room height increase, 
the floor thickness changes, the relocation of a non-structural 
internal wall, and their associated wall, room and corridor changes 
do not change fire barrier performance, and the fire loading 
analyses results remain acceptable. The room height and floor 
thickness changes are consistent with the annex building 
configuration used in the building's structural analysis. The 
relocated internal wall is non-structural, thus the structural 
analyses for the annex building are not affected. The affected rooms 
and associated equipment do not interface with components that 
contain radioactive material. The affected rooms do not contain 
equipment whose failure could initiate an accident. The proposed 
changes do not create a new fault or sequence of events that could 
result in a radioactive material release.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed additions of a new nonsafety-related battery, 
battery room and battery equipment room, the room height increase, 
the floor thickness changes, the relocation of a non-structural 
internal wall, and their associated wall, room and corridor changes 
do not change the fire barrier performance of the affected fire 
areas. The affected rooms do not contain safety-related equipment, 
and the safe shutdown fire analysis is not affected. Because the 
proposed change does not alter compliance with the construction 
codes to which the annex building is designed and constructed, the 
proposed changes to the structural configuration, including 
anticipated equipment loading, room height, and floor thickness do 
not adversely affect the safety margins associated with the seismic 
Category II structural capability of the annex building.
    The floor areas and amounts of combustible material loads in 
affected fire zones and areas do not significantly change, such that 
their fire duration times remain within their two-hour design value, 
thus the safety margins associated with the fire loads analysis are 
not affected.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes, thus no margin of 
safety is reduced.
    Therefore, 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart.
South Carolina Electric and Gas Company, Docket Nos.: 52-027 and 52-
028, Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield 
County, South Carolina
    Date of amendment request: February 10, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15041A698.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-93 and NPF-94 for the Virgil C. Summer 
Nuclear Station (VCSNS) Units 2 and 3 by revising Tier 2* information 
contained within the Human Factors Engineering Design Verification, 
Task Support Verification and Integrated System Validation plans. These 
documents are incorporated by reference into the VCSNS Units 2 and 3 
Updated Final Safety Analysis Report and will additionally require 
changes to be made to affected Tier 2 information.
    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 proposed amendment includes changes to Integrated System 
Validation (ISV) activities, which are performed on the AP1000 plant 
simulator to validate the adequacy of the AP1000 human systems 
interface design and confirm that it meets human factors engineering 
principles. The proposed changes involve administrative details 
related to performance of the ISV, and no plant hardware or 
equipment is affected whose failure could initiate an accident, or 
that interfaces with a component that could initiate an accident, or 
that contains radioactive material. Therefore, these changes have no 
effect on any accident initiator in the Updated Final Safety 
Analysis Report (UFSAR), nor do they affect the radioactive material 
releases in the UFSAR accident analysis.

[[Page 17095]]

    Therefore, 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 accident previously 
evaluated?
    Response: No.
    The proposed amendment includes changes to ISV activities, which 
are performed on the AP1000 plant simulator to validate the adequacy 
of the AP1000 human system interface design and confirm that it 
meets human factors engineering principles. The proposed changes 
involve administrative details related to performance of the ISV, 
and no plant hardware or equipment is affected whose failure could 
initiate an accident, or that interfaces with a component that could 
initiate an accident, or that contains radioactive material. 
Although the ISV may identify a need to initiate changes to add, 
modify, or remove plant structures, systems, or components, these 
changes will not be made directly as part of the ISV.
    Therefore, 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 a margin of safety?
    Response: No.
    The proposed amendment includes changes to ISV activities, which 
are performed on the AP1000 plant simulator to validate the adequacy 
of the AP1000 human system interface design and confirm that it 
meets human factors engineering principles. The proposed changes 
involve administrative details related to performance of the ISV, 
and do not affect any safety-related equipment, design code 
compliance, design function, design analysis, safety analysis input 
or result, or design/safety margin. No safety analysis or design 
basis acceptance limit/criterion is challenged or exceeded by the 
proposed changes, thus no margin of safety is reduced.
    Therefore, 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 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    NRC Branch Chief: Lawrence Burkhart.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-321 and 50-
366, Hatch Nuclear Plant, Units 1 and 2, Appling County, Georgia
    Date of amendment request: October 10, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14288A226.
    Description of amendment request: The licensee requested 21 
revisions to the Technical Specifications. The licensee states the 
changes were chosen to increase the consistency between the Hatch 
Technical Specifications, the Improved Standard Technical 
Specifications, and the Technical Specifications of other plants in the 
Southern Nuclear Operating Company fleet. A list of the requested 
revisions is included in Enclosure 1 of the application.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration for each of the 24 changes requested, which is presented 
below:

2.1 TSTF-30-A, Revision 3, ``Extend the Completion Time for Inoperable 
Isolation Valve to a Closed System to 72 Hours.''

    Specification 3.6.1.3, ``Primary Containment Isolation Valves 
(PCIVs),'' Action C, TS page 3.6-9, is revised to provide a 72 hour 
Completion Time for penetration flow paths with one inoperable PCIV 
with a closed system.
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change extends the Completion Time to isolate an 
inoperable primary containment isolation valve (PCIV) from 4 hours 
to 72 hours when the PCIV is associated with a closed system. The 
PCIVs are not an initiator of any accident previously evaluated. The 
consequences of a previously evaluated accident during the extended 
Completion Time are the same as the consequences during the existing 
Completion Time.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change extends the Completion Time to isolate an 
inoperable primary containment isolation valve (PCIV) from 4 hours 
to 72 hours when the PCIV is associated with a closed system. The 
PCIVs serve to mitigate the potential for radioactive release from 
the primary containment following an accident. The design and 
response of the PCIVs to an accident are not affected by this 
change. The revised Completion Time is appropriate given the 
isolation capability of the closed system.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.2 TSTF-45-A, Revision 2, ``Exempt Verification of CIVs that are 
Locked, Sealed or Otherwise Secured''

    The proposed change revises SRs 3.6.1.3.2 and 3.6.1.3.3 in 
Specification 3.6.1.3, ``Primary Containment Isolation Valves 
(PCIVs),'' to exempt manual PCIVs and blind flanges which are 
locked, sealed, or otherwise secured in position from position 
verification requirements. The proposed change also revises SR 
3.6.4.2.1 in Specification 3.6.4.2, ``Secondary Containment 
Isolation Valves (SCIVs),'' to exempt manual SCIVs and blind flanges 
which are locked, sealed, or otherwise secured in position from 
position verification requirements.
    Signification Hazards Consideration: SNC has evaluated whether 
or not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change exempts manual primary containment isolation 
valves and blind flanges located inside and outside of containment, 
and manual secondary containment isolation valves and blind flanges, 
that are locked, sealed, or otherwise secured in position from the 
periodic verification of valve position required by Surveillance 
Requirements 3.6.1.3.2, 3.6.1.3.3, and 3.6.4.2.1. The exempted 
valves and devices are verified to be in the correct position upon 
being locked, sealed, or secured. Because the valves and devices are 
in the condition assumed in the accident analysis, the proposed 
change will not affect the initiators or mitigation of any accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.

[[Page 17096]]

    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change exempts manual primary containment isolation 
valves and blind flanges located inside and outside of containment, 
and manual secondary containment isolation valves and blind flanges, 
that are locked, sealed, or otherwise secured in position from the 
periodic verification of valve position required by Surveillance 
Requirements 3.6.1.3.2, 3.6.1.3.3, and 3.6.4.2.1. These valves and 
devices are administratively controlled and their operation is a 
non-routine event. The position of a locked, sealed or secured blind 
flange or valve is verified at the time it is locked, sealed or 
secured, and any changes to their position is performed under 
administrative controls. Industry experience has shown that these 
valves are generally found to be in the correct position. Since the 
change impacts only the frequency of verification for blind flange 
and valve position, the proposed change will provide a similar level 
of assurance of correct position as the current frequency of 
verification.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.3 TSTF-46-A, Revision 1, ``Clarify the CIV Surveillance to Apply Only 
to Automatic Isolation Valves''

    The proposed change modifies SR 3.6.1.3.5 in Specification 
3.6.1.3, ``Primary Containment Isolation Valves (PCIVs),'' and SR 
3.6.4.2.2, in Specification 3.6.4.2, ``Secondary Containment 
Isolation Valves (SCIVs),'' including their associated Bases, to 
delete the requirement to verify the isolation time of ``each power 
operated'' containment isolation valve and only require verification 
of each ``power operated automatic isolation valve.''
    Signification Hazards Consideration: SNC has evaluated whether 
or not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the requirements in Technical 
Specification Surveillance Requirements (SRs) 3.6.1.3.5 and 
3.6.4.2.2, and their associated Bases, to delete the requirement to 
verify the isolation time of ``each power operated'' PCIV and SCIV 
and only require verification of closure time for each ``automatic 
power operated isolation valve.'' The closure times for PCIVs and 
SCIVs that do not receive an automatic closure signal are not an 
initiator of any design basis accident or event, and therefore the 
proposed change does not increase the probability of any accident 
previously evaluated. The PCIVs and SCIVs are used to respond to 
accidents previously evaluated. Power operated PCIVs and SCIVs that 
do not receive an automatic closure signal are not assumed to close 
in a specified time. The proposed change does not change how the 
plant would mitigate 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not result in a change in the manner in 
which the PCIVs and SCIVs provide plant protection or introduce any 
new or different operational conditions. Periodic verification that 
the closure times for PCIVs and SCIVs that receive an automatic 
closure signal are within the limits established by the accident 
analysis will continue to be performed under SRs 3.6.1.3.5 and 
3.6.4.2.2. The change does not alter assumptions made in the safety 
analysis, and is consistent with the safety analysis assumptions and 
current plant operating practice. There are also no design changes 
associated with the proposed changes, and the change does not 
involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed).
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change provides clarification that only PCIVs and 
SCIVs that receive an automatic isolation signal are within the 
scope of SRs 3.6.1.3.5 and 3.6.4.2.2. The proposed change does not 
result in a change in the manner in which the PCIVs and SCIVs 
provide plant protection. Periodic verification that closure times 
for PCIVs and SCIVs that receive an automatic isolation signal are 
within the limits established by the accident analysis will continue 
to be performed. The proposed change does not affect the safety 
analysis acceptance criteria for any analyzed event, nor is there a 
change to any safety analysis limit. The proposed change does not 
alter the manner in which safety limits, limiting safety system 
settings or limiting conditions for operation are determined, nor is 
there any adverse effect on those plant systems necessary to assure 
the accomplishment of protection functions. The proposed change will 
not result in plant operation in a configuration outside the design 
basis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.4 TSTF-222-A, Revision 1, ``Control Rod Scram Time Testing''

    Specification 3.1.4, ``Control Rod Scram Times,'' SRs 3.1.4.1 
and 3.1.4.4, are revised to only require scram time testing of 
control rods that are in an affected core cell. The SR 3.1.4.1 
Frequency ``Prior to exceeding 40% RTP after fuel movement within 
the reactor vessel,'' is eliminated and a new Frequency is added to 
SR 3.1.4.4 which states, ``Prior to exceeding 40% RTP after fuel 
movement within the affected core cell.''
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change clarifies the intent of Surveillance testing 
in Specification 3.1.4, ``Control Rod Scram Times.'' The existing 
Specification wording requires control rod scram time testing of all 
control rods whenever fuel is moved within the reactor pressure 
vessel, even though the Technical Specification Bases state that 
control rod scram time testing is only required in the affected core 
cells. The Frequency of Surveillances 3.1.4.1 and 3.1.4.4 are 
revised to implement the Bases statement in the Specifications. The 
proposed change does not affect any plant equipment, test methods, 
or plant operation, and are not initiators of any analyzed accident 
sequence. The control rods will continue to perform their function 
as designed. Operation in accordance with the proposed Technical 
Specifications will ensure that all analyzed accidents will continue 
to be mitigated as previously analyzed.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods

[[Page 17097]]

governing normal plant operation. The changes do not alter the 
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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change clarifies the intent of Surveillance testing 
in Specification 3.1.4, ``Control Rod Scram Times.'' The existing 
Specification wording requires control rod scram time testing of all 
control rods whenever fuel is moved within the reactor pressure 
vessel, even though the Technical Specification Bases state that the 
control rod scam time testing is only required in the affected core 
cells. The proposed change will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. Control rod scram time testing will be performed following 
any fuel movement that could affect the scram time.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.5 TSTF-264-A, Revision 0, ``3.3.9 and 3.3.10--Delete Flux Monitors 
Specific Overlap Requirement SRs''

    The proposed change revises Specification 3.3.1.1, ``RPS 
Instrumentation,'' by deleting Surveillances 3.3.1.1.6 and 
3.3.1.1.7, which verify the overlap between the source range monitor 
(SRM) and the intermediate range monitor (IRM), and between the IRM 
and the average power range monitor (APRM).
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates two Surveillances Requirements 
(SRs) (SRs 3.3.1.1.6 and 3.3.1.1.7) which verify the overlap between 
the source range monitor (SRM) and intermediate range monitor (IRM) 
and between the IRM and the average power range monitor (APRM). The 
testing requirement is incorporated in the existing Channel Check 
Surveillance (SR 3.3.1.1.1). The proposed change does not affect any 
plant equipment, test methods, or plant operation, and are not 
initiators of any analyzed accident sequence. The SRM, IRM, and APRM 
will continue to perform their function as designed. Operation in 
accordance with the proposed Technical Specifications will ensure 
that all analyzed accidents will continue to be mitigated as 
previously analyzed.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change eliminates SRs 3.3.1.1.6 and 3.3.1.1.7 which 
verify the overlap between the SRM and IRM and between the IRM and 
the APRM. The testing requirement is incorporated in the existing 
Channel Check Surveillance (SR 3.3.1.1.1). The proposed change will 
not affect the operation of plant equipment or the function of any 
equipment assumed in the accident analysis. Instrument channel 
overlap will continue to be verified under the existing Channel 
Check surveillance.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.6 TSTF-269-A, Revision 2, ``Allow Administrative Means of Position 
Verification for Locked or Sealed Valves''

    The proposed change modifies Specification 3.6.1.3, ``Primary 
Containment Isolation Valves,'' and Specification 3.6.4.2, 
``Secondary Containment Isolation Valves.'' The specifications 
require penetrations with an inoperable isolation valve to be 
isolated and periodically verified to be isolated. A Note is added 
to Specification 3.6.1.3, Actions A and C, and Specification 
3.6.4.2, Action A, to allow isolation devices that are locked, 
sealed, or otherwise secured to be verified by use of administrative 
means.
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change modifies Specification 3.6.1.3, ``Primary 
Containment Isolation Valves,'' and Specification 3.6.4.2, 
``Secondary Containment Isolation Valves.'' The specifications 
require penetrations with an inoperable isolation valve to be 
isolated and periodically verified to be isolated. A Note is added 
to Specification 3.6.1.3, Actions A and C, and Specification 
3.6.4.2, Action A, to allow isolation devices that are locked, 
sealed, or otherwise secured to be verified by use of administrative 
means. The proposed change does not affect any plant equipment, test 
methods, or plant operation, and are not initiators of any analyzed 
accident sequence. The inoperable containment penetrations will 
continue to be isolated, and hence perform their isolation function. 
Operation in accordance with the proposed Technical Specifications 
will ensure that all analyzed accidents will continue to be 
mitigated as previously analyzed.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. The primary and secondary containment isolation valves 
will continue to be operable or will be isolated as required by the 
existing specifications.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.7 TSTF-273-A, Revision 2, ``Safety Function Determination Program 
Clarifications''

    The proposed Technical Specification (TS) changes add 
explanatory text to the Bases for limiting condition for operation 
(LCO) 3.0.6 clarifying the ``appropriate LCO for loss of function,'' 
and that consideration does not have to be made for a loss of power 
in determining loss of function. Explanatory text is also added to 
the programmatic description of the Safety Function Determination 
Program (SFDP) in Specification 5.5.12 to provide clarification of 
these same issues.
    Signification Hazards Consideration: SNC has evaluated whether 
or not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the

[[Page 17098]]

three standards set forth in 10 CFR 50.92, ``Issuance of 
amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed Technical Specification (TS) changes add 
explanatory text to the programmatic description of the Safety 
Function Determination Program (SFDP) in Specification 5.5.10 to 
clarify in the requirements that consideration does not have to be 
made for a loss of power in determining loss of function. The Bases 
for limiting condition for operations (LCO) 3.0.6 are revised to 
provide clarification of the ``appropriate LCO for loss of 
function,'' and that consideration does not have to be made for a 
loss of power in determining loss of function. The changes are 
editorial and administrative in nature, and therefore do not 
increase the probability of any accident previously evaluated. No 
physical or operational changes are made to the plant. The proposed 
change does not change how the plant would mitigate 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes are editorial and administrative in nature 
and do not result in a change in the manner in which the plant 
operates. The loss of function of any specific component will 
continue to be addressed in its specific TS LCO and plant 
configuration will be governed by the required actions of those 
LCOs. The proposed changes are clarifications that do not degrade 
the availability or capability of safety related equipment, and 
therefore do not create the possibility of a new or different kind 
of accident from any accident previously evaluated. There are no 
design changes associated with the proposed changes, and the changes 
do not involve a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed). The changes do not 
alter assumptions made in the safety analysis, and are consistent 
with the safety analysis assumptions and current plant operating 
practice. Due to the administrative nature of the changes, they 
cannot be an accident initiator.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to TS 5.5.10 are clarifications and are 
editorial and administrative in nature. No changes are made the LCOs 
for plant equipment, the time required for the TS Required Actions 
to be completed, or the out of service time for the components 
involved. The proposed changes do not affect the safety analysis 
acceptance criteria for any analyzed event, nor is there a change to 
any safety analysis limit. The proposed changes do not alter the 
manner in which safety limits, limiting safety system settings or 
limiting conditions for operation are determined, nor is there any 
adverse effect on those plant systems necessary to assure the 
accomplishment of protection functions. The proposed changes will 
not result in plant operation in a configuration outside the design 
basis.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.8 TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode Restriction 
Notes''

    The proposed change revises several Specification 3.8.1, ``AC 
Sources--Operating,'' Surveillance Notes to allow full or partial 
performance of the SRs to re-establish Operability provided an 
assessment determines the safety of the plant is maintained or 
enhanced. These Surveillances currently have Notes prohibiting their 
performance in Modes 1 or 2, or in Modes 1, 2, or 3.
    SR 3.8.1.6 (ISTS SR 3.8.1.8), which tests the transfer of 
Alternating (AC) sources from normal to alternate offsite circuits, 
contains a Note prohibiting performance in Mode 1 or 2. The Note is 
modified to state that performance is normally prohibited in Mode 1 
or 2 but may be performed to re-establish Operability provided an 
assessment determines the safety of the plant is maintained or 
enhanced.
    SR 3.8.1.7 (ISTS SR 3.8.1.9), which tests the ability of the 
emergency diesel generator (DG) to reject a load greater than or 
equal to its associated single largest post-accident load, contains 
a Note prohibiting performance in Mode 1 or 2. An exception is 
provided for the swing DG. The Note is modified to state that 
performance is normally prohibited in Mode 1 or 2 but may be 
performed to re- establish Operability provided an assessment 
determines the safety of the plant is maintained or enhanced.
    SR 3.8.1.8 (ISTS SR 3.8.1.10), which tests emergency DG 
operation following a load rejection of greater than or equal to 
2775 kW, contains a Note prohibiting performance in Mode 1 or 2. The 
Note is modified to state that performance is normally prohibited in 
Mode 1 or 2 but portions of the SR may be performed to re- establish 
Operability provided an assessment determines the safety of the 
plant is maintained or enhanced.
    SR 3.8.1.9 (ISTS SR 3.8.1.11), which tests the response to a 
loss of offsite power signal, contains a Note prohibiting 
performance in Mode 1, 2, or 3. The Note is modified to state that 
performance is normally prohibited in Mode 1, 2, or 3, but portions 
of the SR may be performed to re-establish Operability provided an 
assessment determines the safety of the plant is maintained or 
enhanced.
    SR 3.8.1.10 (ISTS SR 3.8.1.12), which tests response to an 
Emergency Core Cooling System (ECCS) initiation signal, contains a 
Note prohibiting performance in Mode 1 or 2. The Note is modified to 
state that performance is normally prohibited in Mode 1 or 2, but 
the SR may be performed to re-establish Operability provided an 
assessment determines the safety of the plant is maintained or 
enhanced.
    SR 3.8.1.11 (ISTS SR 3.8.1.13), which tests that each DGs 
automatic trips are bypassed on a loss of voltage signal concurrent 
with an ECCS initiation signal, contains a Note prohibiting 
performance in Mode 1, 2, or 3. The Note is modified to state that 
performance is normally prohibited in Mode 1, 2, or 3, but the SR 
may be performed to re-establish Operability provided an assessment 
determines the safety of the plant is maintained or enhanced.
    SR 3.8.1.12 (ISTS SR 3.8.1.14), which performs a 24 hour loaded 
test run of the DG, contains a Note prohibiting performance in Mode 
1 or 2. The Note is modified to state that performance is normally 
prohibited in Mode 1 or 2, but the SR may be performed to re-
establish Operability provided an assessment determines the safety 
of the plant is maintained or enhanced.
    SR 3.8.1.14 (ISTS SR 3.8.1.16), which verifies transfer from DG 
to offsite power, contains a Note prohibiting performance in Mode 1, 
2, or 3. The Note is modified to state that performance is normally 
prohibited in Mode 1, 2, or 3, but portions of the SR may be 
performed to re-establish Operability provided an assessment 
determines the safety of the plant is maintained or enhanced.
    SR 3.8.1.15 (ISTS SR 3.8.1.17), which verifies than a DG 
operating in test mode will return to ready-to-load condition and 
energize the emergency load from offsite power on receipt of an ECCS 
initiation signal, contains a Note prohibiting performance in Mode 
1, 2, or 3. The Note is modified to state that performance is 
normally prohibited in Mode 1, 2, or 3, but portions of the SR may 
be performed to re-establish Operability provided an assessment 
determines the safety of the plant is maintained or enhanced.
    SR 3.8.1.16 (ISTS SR 3.8.1.18), which verifies the interval 
between each sequenced load, contains a Note prohibiting performance 
in Mode 1, 2, or 3. The Note is modified to state that performance 
is normally prohibited in Mode 1, 2, or 3, but the SR may be 
performed to re-establish Operability provided an assessment 
determines the safety of the plant is maintained or enhanced.
    SR 3.8.1.17 (ISTS SR 3.8.1.19), which verifies the response to a 
loss of offsite power signal and Engineered Safety Features (ESF) 
actuation signal, contains a Note prohibiting performance in Mode 1, 
2, or 3. The Note is modified to state that performance is normally 
prohibited in Mode 1, 2, or 3, but portions of the SR may be 
performed to re-establish Operability provided an assessment 
determines the safety of the plant is maintained or enhanced.
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:

[[Page 17099]]

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change modifies Mode restriction Notes on eleven 
emergency diesel generator (DG) Surveillances to allow performance 
of the Surveillance in whole or in part to re-establish emergency DG 
Operability. The emergency DGs and their associated emergency loads 
are accident mitigating features, and are not an initiator of any 
accident previously evaluated. As a result the probability of any 
accident previously evaluated is not increased. The proposed change 
allows Surveillance testing to be performed in whole or in part to 
re-establish Operability of an emergency DG. The consequences of an 
accident previously evaluated during the period that the emergency 
DG is being tested to re-establish Operability are no different from 
the consequences of an accident previously evaluated while the 
emergency DG is inoperable. As a result, the consequences of any 
accident previously evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The purpose of Surveillances is to verify that equipment is 
capable of performing its assumed safety function. The proposed 
change will only allow the performance of the Surveillances to re-
establish Operability and the proposed changes may not be used to 
remove an emergency DG from service. The proposed changes also 
require an assessment to verify that plant safety will be maintained 
or enhanced by performance of the Surveillance in the normally 
prohibited Modes.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.9 TSTF-284-A, Revision 3, ``Add `Met vs. Perform' to Technical 
Specification 1.4, Frequency''

    The change inserts a discussion paragraph into Specification 
1.4, and two new examples are added to facilitate the use and 
application of SR Notes that utilize the terms ``met'' and 
``perform.''
    Signification Hazards Consideration: SNC has evaluated whether 
or not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes insert a discussion paragraph into 
Specification 1.4, and several new examples are added to facilitate 
the use and application of Surveillance Requirement (SR) Notes that 
utilize the terms ``met'' and ``perform''. The changes also modify 
SRs in multiple Specifications to appropriately use ``met'' and 
``perform'' exceptions. The changes are administrative in nature 
because they provide clarification and correction of existing 
expectations, and therefore the proposed change does not increase 
the probability of any accident previously evaluated. No physical or 
operational changes are made to the plant. The proposed change does 
not significantly change how the plant would mitigate 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes are administrative in nature and do not 
result in a change in the manner in which the plant operates. The 
proposed changes do not degrade the availability or capability of 
safety related equipment, and therefore do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. There are no design changes associated with 
the proposed changes, and the changes do not involve a physical 
alteration of the plant (i.e., no new or different type of equipment 
will be installed). The changes do not alter assumptions made in the 
safety analysis, and are consistent with the safety analysis 
assumptions and current plant operating practice. Due to the 
administrative nature of the changes, they cannot be an accident 
initiator.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes are administrative in nature and do not 
result in a change in the manner in which the plant operates. The 
proposed changes provide clarification and correction of existing 
expectations that do not degrade the availability or capability of 
safety related equipment, or alter their operation. The proposed 
changes do not affect the safety analysis acceptance criteria for 
any analyzed event, nor is there a change to any safety analysis 
limit. The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined, nor is there any adverse effect on those 
plant systems necessary to assure the accomplishment of protection 
functions. The proposed changes will not result in plant operation 
in a configuration outside the design basis.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.10 TSTF-295-A, Revision 0, ``Modify Note 2 to Actions of PAM Table to 
Separate Condition Entry for Each Penetration''

    Specification 3.3.3.1, ``Post Accident Monitoring (PAM) 
Instrumentation,'' Function 6, is renamed from ``Primary Containment 
Isolation Valve Position'' to ``Penetration Flow Path Primary 
Containment Isolation Valve Position.''
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change clarifies the separate condition entry Note 
in Specification 3.3.3.1, ``Post Accident Monitoring (PAM) 
Instrumentation,'' for Function 6, ``Primary Containment Isolation 
Valve Position,'' and Function 9, ``Suppression Pool Water 
Temperature.'' The proposed change does not affect any plant 
equipment, test methods, or plant operation, and are not initiators 
of any analyzed accident sequence. The actions taken for inoperable 
PAM channels are not changed. Operation in accordance with the 
proposed Technical Specifications will ensure that all analyzed 
accidents will continue to be mitigated as previously analyzed.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods

[[Page 17100]]

governing normal plant operation. The changes do not alter the 
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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. The PAM channels will continue to be operable or the 
existing, appropriate actions will be followed.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.11 TSTF-306-A, Revision 2, ``Add Action to LCO 3.3.6.1 to Give Option 
to Isolate the Penetration''

    The proposed change revises Specification 3.3.6.1, ``Primary 
Containment Isolation Instrumentation.'' An Actions Note is added 
allowing penetration flow paths to be unisolated intermittently 
under administrative controls. The traversing incore probe (TIP) 
isolation system is also segregated into a separate Function, 
allowing 12 hours to place the channel in trip and 24 hours to 
isolate the penetration. A new Condition G is added for the new TIP 
isolation system Function. Condition G is referenced from Required 
Action C.1 when Conditions A or B are not met. The subsequent 
Actions are renumbered.
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises Specification 3.3.6.1, ``Primary 
Containment Isolation Instrumentation.'' An Actions Note is added 
allowing penetration flow paths to be unisolated intermittently 
under administrative controls. The traversing incore probe (TIP) 
isolation system is segregated into a separate Function, allowing 12 
hours to place the channel in trip and 24 hours to isolate the 
penetration. A new Action G is added which is referenced by the new 
TIP isolation system Function. The subsequent Actions are 
renumbered. The proposed change does not affect any plant equipment, 
test methods, or plant operation, and are not initiators of any 
analyzed accident sequence. The allowance to unisolate a penetration 
flow path will not have a significant effect on mitigation of any 
accident previously evaluated because the penetration flow path can 
be isolated, if needed, by a dedicated operator. The option to 
isolate a TIP System penetration will ensure the penetration will 
perform as assumed in the accident analysis. Operation in accordance 
with the proposed Technical Specifications will ensure that all 
analyzed accidents will continue to be mitigated as previously 
analyzed.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change will not affect the operation of plant 
equipment or the function of any equipment assumed in the accident 
analysis. The allowance to unisolate a penetration flow path will 
not have a significant effect on a margin of safety because the 
penetration flow path can be isolated manually, if needed. The 
option to isolate a TIP System penetration will ensure the 
penetration will perform as assumed in the accident analysis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.12 TSTF-308-A, Revision 1, ``Determination of Cumulative and 
Projected Dose Contributions in RECP''

    The proposed change revises Specification 5.5.4, ``Radioactive 
Effluent Controls Program,'' paragraph e, to describe the original 
intent of the dose projections.
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises Specification 5.5.4, ``Radioactive 
Effluent Controls Program,'' paragraph e, to describe the original 
intent of the dose projections. The cumulative and projection of 
doses due to liquid releases are not an assumption in any accident 
previously evaluated and have no effect on the mitigation of any 
accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises Specification 5.5.4, ``Radioactive 
Effluent Controls Program,'' paragraph e, to describe the original 
intent of the dose projections. The cumulative and projection of 
doses due to liquid releases are administrative tools to assure 
compliance with regulatory limits. The proposed change revises the 
requirement to clarify the intent, thereby improving the 
administrative control over this process. As a result, any effect on 
the margin of safety should be minimal.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.13 TSTF-318-A, Revision 0, ``Revise 3.5.1 for One LPCI Pump 
Inoperable in Each of Two ECCS Divisions''

    The proposed change adds a provision to Condition A of 
Specification 3.5.1, ``ECCS--Operating,'' to allow one Low Pressure 
Coolant Injection (LPCI) pump to be inoperable in each subsystem for 
a period of seven days.
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change adds a provision to Condition A of Technical 
Specification (TS) 3.5.1 to allow one Low Pressure Coolant Injection 
(LPCI) pump to be inoperable in each subsystem for a period of seven 
days. The change to allow one LPCI pump to be inoperable in both 
subsystems is more reliable than what is currently allowed by 
Condition A, which requires entry into

[[Page 17101]]

shutdown limiting condition for operation (LCO) 3.0.3 under these 
conditions. The LPCI mode of the Residual Heat Removal system is not 
assumed to be initiator of any analyzed event sequence. The 
consequences of an accident previously evaluated under the proposed 
allowance are no different than the consequences under the existing 
requirements.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change adds a provision to Condition A of Technical 
Specification TS 3.5.1 to allow one LPCI pump to be inoperable in 
each subsystem for a period of seven days. The change to allow one 
LPCI pump to be inoperable in both subsystems is more reliable than 
what is currently allowed by Condition A, which requires entry into 
shutdown LCO 3.0.3 under these conditions. The proposed change does 
not affect any safety analysis assumptions.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.14 TSTF-322-A, Revision 2, ``Secondary Containment and Shield 
Building Boundary Integrity SRs'

    The proposed change revises Specification 3.6.4.1, ``Secondary 
Containment,'' SRs 3.6.4.1.3 and 3.6.4.1.4 to clarify the intent of 
the Surveillances.
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises Specification 3.6.4.1, ``Secondary 
Containment,'' Surveillance Requirements (SRs) 3.6.4.1.3 and 
3.6.4.1.4 to clarify the intent of the Surveillances. The secondary 
containment and the standby gas treatment (SGT) system are not 
initiators of any accident previously evaluated. Operation in 
accordance with the proposed Technical Specifications will ensure 
that all analyzed accidents will continue to be mitigated as 
previously analyzed.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change is an clarification of the intent of the 
surveillances to ensure that the secondary containment is not 
inappropriately declared inoperable when a SGT subsystem is 
inoperable. The safety functions of the secondary containment and 
the SGT system are not affected. This change is a correction that 
ensures that the intent of the secondary containment surveillances 
is clear.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.15 TSTF-323-A, Revision 0, ``EFCV Completion Time to 72 hours''

    The proposed change revises Specification 3.6.1.3, ``Primary 
Containment Isolation Valves,'' Action C, to provide a 72 hour 
Completion Time instead of a 12 hour Completion Time to isolate an 
inoperable excess flow check valve (EFCV).
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises Specification 3.6.1.3, ``Primary 
Containment Isolation Valves,'' Action C, to provide a 72 hour 
Completion Time instead of a 12 hour Completion Time to isolate an 
inoperable excess flow check valve (EFCV). The primary containment 
isolation valves (PCIVs) are not an initiator of any accident 
previously evaluated. The consequences of a previously evaluated 
accident during the extended Completion Time are the same as the 
consequences during the existing Completion Time.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change extends the Completion Time to isolate an 
inoperable primary containment penetration equipped with an excess 
flow check valve from 12 hours to 72 hours. The PCIVs serve to 
mitigate the potential for radioactive release from the primary 
containment following an accident. The design and response of the 
PCIVs to an accident are not affected by this change. The revised 
Completion Time is appropriate given the EFCVs are on penetrations 
that have been found to have acceptable barrier(s) in the event that 
the single isolation valve fails.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.16 TSTF-374-A, Revision 0, ``Revision to TS 5.5.13 and Associated TS 
Bases for Diesel Fuel Oil''

    The proposed change revises Specification 5.5.9, ``Diesel Fuel 
Oil Testing Program,'' to remove references to the specific American 
Society for Testing and Materials (ASTM) Standard from the 
Administrative Controls Section of TS, and places them in a 
licensee-controlled document. Also, alternate criteria are added to 
establish the acceptability of new fuel oil for use prior to and 
following the addition to storage tanks.
    Signification Hazards Consideration: SNC has evaluated whether 
or not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 17102]]

    Response: No.
    The proposed changes remove the references to specific ASTM 
standards from the Administrative Controls Section of the Technical 
Specifications (TS) and place them in a licensee controlled 
document. Requirements to perform testing in accordance with the 
applicable ASTM standards is retained in the TS as are requirements 
to perform testing of both new and stored diesel fuel oil. Future 
changes to the licensee controlled document will be evaluated 
pursuant to the requirements of 10 CFR 50.59 to ensure that these 
changes do not result in more than a minimal increase in the 
probability or consequences of an accident previously evaluated. In 
addition, tests used to establish the acceptability of new fuel oil 
for use prior to and following the addition to storage tanks has 
been expanded to recognize more rigorous testing of water and 
sediment content. Relocating the specific ASTM standard references 
from the TS to a licensee controlled document and allowing a water 
and sediment content test to be performed to establish the 
acceptability of new fuel oil will not affect nor degrade the 
ability of the emergency diesel generators (EDGs) to perform their 
specified safety function. Fuel oil quality will continue to be 
tested and maintained to ASTM requirements. Diesel fuel oil testing 
is not an initiator of any accident previously evaluated, and the 
proposed changes do not adversely affect any accident initiators or 
precursors, or alter design assumptions, conditions, and 
configuration of the facility, or the manner in which the plant is 
operated. The proposed changes do not adversely affect the ability 
of structures, systems, and components to perform their intended 
safety function to mitigate the consequences of an initiating event 
within the assumed acceptance limits.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes remove the references to specific ASTM 
standards from the Administrative Controls Section of TS and place 
them in a licensee controlled document. In addition, the tests used 
to establish the acceptability of new fuel oil for use prior to and 
following the addition to storage tanks has been expanded to allow a 
water and sediment content test to be performed to establish the 
acceptability of new fuel oil. 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. The requirements retained in the TS will continue to 
require testing of new and stored diesel fuel oil to ensure the 
proper functioning of the EDGs.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes remove the references to specific ASTM 
standards from the Administrative Controls Section of TS and place 
them in a licensee controlled document. Instituting the proposed 
changes will continue to ensure the use of applicable ASTM standards 
to evaluate the changes to the licensee-controlled document are 
performed in accordance with the provisions of 10 CFR 50.59. This 
approach provides an effective level of regulatory control and 
ensures that diesel fuel oil testing is conducted such that there is 
no significant reduction in a margin of safety. The margin of safety 
provided by the EDGs is unaffected by the proposed changes since TS 
requirements will continue to ensure fuel oil is of the appropriate 
quality. The proposed changes provide the flexibility needed to 
improve fuel oil sampling and analysis methodologies while 
maintaining sufficient controls to preserve the current margins of 
safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.17 TSTF-400-A, Revision 1, ``Clarify SR on Bypass of DG Automatic 
Trips''

    The proposed change revises Specification 3.8.1, ``AC Sources--
Operating,'' Surveillance 3.8.1.11, to clarify that the intent of 
the SR is to test the non-critical emergency DG automatic trips.
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change clarifies the purpose of Surveillance Requirement 
(SR) 3.8.1.11, which is to verify that non-critical automatic 
emergency diesel generator (DG) trips are bypassed in an accident. 
The non-critical automatic DG trips and their bypasses are not 
initiators of any accident previously evaluated. Therefore, the 
probability of any accident is not significantly increased. 
Additionally, the function of the emergency DG in mitigating 
accidents is not changed. The revised SR continues to ensure the 
emergency DG will operate as assumed in the accident analysis.
    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.
    This change clarifies the purpose of SR 3.8.1.11, which is to 
verify that non-critical automatic emergency DG trips are bypassed 
in an accident. The proposed change does not involve a physical 
alteration of the plant (no new or different type of equipment will 
be installed), or a change in the methods governing normal plant 
operation. Thus, this 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.
    This change clarifies the purpose of SR 3.8.1.11, which is to 
verify that non-critical automatic DG trips are bypassed in an 
accident. This change clarifies the purpose of the SR, which is to 
verify that the emergency DG is capable of performing the assumed 
safety function. The safety function of the emergency DG is 
unaffected, so the change does not affect the margin of safety.
    Therefore, this change does not involve a significant reduction 
in a margin of safety.
    Based on the above, SNC concludes that the proposed change 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

2.18 TSTF-439-A, Revision 2, ``Eliminate Second Completion Times 
Limiting Time From Discovery of Failure To Meet an LCO''

    Specifications 3.1.7, ``Standby Liquid Control (SLC) System;'' 
3.6.4.3, ``Standby Gas Treatment (SGT) System;'' 3.8.1, ``AC 
Sources--Operating;'' and 3.8.7, ``Distribution Systems--
Operating,'' contain Required Actions with a second Completion Time 
to establish a limit on the maximum time allowed for any combination 
of Conditions that result in a single continuous failure to meet the 
LCO. These Completion Times (henceforth referred to as ``second 
Completion Times'') are joined by an ``AND'' logical connector to 
the Condition-specific Completion Time and state ``X days from 
discovery of failure to meet the LCO'' (where ``X'' varies by 
specification). The proposed change deletes these second Completion 
Times from the affected Required Actions. It also revises ISTS 
Example 1.3-3 to remove the discussion of second Completion Times 
and to revise the discussion in that Example to state that 
alternating between Conditions in such a manner that operation could 
continue indefinitely without restoring systems to meet the LCO is 
inconsistent with the basis of the Completion Times and is 
inappropriate. Therefore, the licensee shall have administrative 
controls to limit the maximum time allowed for any combination of 
Conditions that result in a single contiguous occurrence of failing 
to meet the LCO.
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.

[[Page 17103]]

    The proposed change eliminates certain Completion Times from the 
Technical Specifications. Completion Times are not an initiator to 
any accident previously evaluated. As a result, the probability of 
an accident previously evaluated is not affected. The consequences 
of an accident during the remaining Completion Time are no different 
than the consequences of the same accident during the removed 
Completion Times. As a result, the consequences of an accident 
previously evaluated are not affected by this change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to delete the second Completion Time 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 changes will not result in plant operation in a 
configuration outside of the design basis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.19 TSTF-458-T, Revision 0, ``Removing Restart of Shutdown Clock for 
Increasing Suppression Pool Temperature''

    The proposed change revises Specification 3.6.2.1, ``Suppression 
Pool Average Temperature,'' Required Actions D and E, to eliminate 
redundant requirements.
    Significant Hazards Consideration SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises Specification 3.6.2.1, ``Suppression 
Pool Average Temperature,'' Required Actions D and E, to eliminate 
redundant requirements when suppression pool temperature is above 
the Limiting Conditions for Operation (LCO) limit. Suppression pool 
temperature is not an initiator to any accident previously 
evaluated. Suppression pool temperature may affect the mitigation of 
accidents previously evaluated. The proposed change reduces the time 
allowed to operate with suppression pool temperature above the 
limit. The consequences of an accident under the proposed change are 
no different than under the current requirements.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises Specification 3.6.2.1, ``Suppression 
Pool Average Temperature,'' Required Actions D and E, to eliminate 
redundant requirements when suppression pool temperature is above 
the LCO limit. The proposed change reduces the time allowed to 
operate with suppression pool temperature above the limit. The 
proposed revision will not adversely affect the margin of safety as 
it corrects the Actions to provide appropriate compensatory measures 
when suppression pool temperature is greater than the limit.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

2.20 TSTF-464-T, Revision 0, ``Clarify the Control Rod Block 
Instrumentation Required Action''

    The proposed change revises Specification 3.3.2.1, Required 
Action C.2.1.2 from ``Verify by administrative methods that startup 
with RWM inoperable has not been performed in the last calendar 
year'' to ``Verify by administrative methods that startup with RWM 
inoperable has not been performed in the last 12 months.''
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises a Required Action to limit startup 
with the Rod Worth Minimizer (RWM) inoperable from once per calendar 
year to once per 12 months. The RWM is used to minimize the 
possibility and consequences of a control rod drop accident. This 
change clarifies the intent of the limitation, but does not affect 
the requirement for the RWM to be operable. As, over time, the 
number of startups with the RWM inoperable will not increase, the 
probability of any accident previously evaluated is not 
significantly increased. As the RWM is still required to be 
operable, the consequences of an any accident previously evaluated 
are not significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises a Required Action to limit startup 
with the Rod Worth Minimizer inoperable from once per calendar year 
to once per 12 months. No new or different accidents result from 
utilizing the proposed change. 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 significant change in the methods governing 
normal plant operation. The changes do not alter assumptions made in 
the safety analysis. The proposed 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 previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change revises a Required Action to limit startup 
with the Rod Worth Minimizer (RWM) inoperable from once per calendar 
year to once per 12 months. This clarifies the intent of the 
Required Action. The number of startups with RWM inoperable is not 
increased.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed change 
presents no significant hazards consideration under the standards 
set forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

2.21 ISTS Adoption #1--Revise the 5.5.7 Introductory Paragraph To Be 
Consistent With the ISTS

    The proposed change revises the introductory paragraph of 
Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP),'' 
to be consistent with the ISTS. Specific requirements to perform 
testing after

[[Page 17104]]

structural maintenance on the HEPA filter or charcoal adsorber 
housing or following painting, fire or chemical release, and after 
every 720 hours of operation are relocated to the licensee- 
controlled program.
    The existing wording states, ``The VFTP will establish the 
required testing of Engineered Safety Feature (ESF) filter 
ventilation systems at the frequencies specified in Regulatory Guide 
1.52, Revision 2, Sections C.5.c and C.5.d, or: (1) After any 
structural maintenance on the HEPA filter or charcoal adsorber 
housings, (2) following painting, fire or chemical release in any 
ventilation zone communicating with the system, or 3) after every 
720 hours of charcoal adsorber operation.''
    The proposed wording states, ``A program shall be established to 
implement the following required testing of Engineered Safety 
Feature (ESF) filter ventilation systems at the frequencies 
specified in Regulatory Guide 1.52, Revision 2, Sections C.5.c and 
C.5.d, and in accordance with Regulatory Guide 1.52, Revision 2.''
    Significant Hazards Consideration: SNC has evaluated whether or 
not a significant hazards consideration is involved with the 
proposed amendment(s) by focusing on the three standards set forth 
in 10 CFR 50.92, ``Issuance of amendment,'' as discussed below:
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the introductory paragraph of 
Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP),'' 
to be consistent with the ISTS. Specific requirements to perform 
testing after structural maintenance on the HEPA filter or charcoal 
adsorber housing or following painting, fire or chemical release, 
and after every 720 hours of operation are retained as a reference 
to Regulatory Guide requirements and general requirements in 
Surveillance Requirement (SR) 3.0.1. Implementation of these 
requirements will be in the licensee-controlled VFTP. The VFTP will 
be maintained in accordance with 10 CFR 50.59. Since any changes to 
the VFTP will be evaluated under 10 CFR 50.59, no significant 
increase in the probability or consequences of an accident 
previously evaluated will be allowed.
    Therefore, this proposed change does not represent a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration to 
the plant (i.e., no new or different type of equipment will be 
installed) or a change to the methods governing normal plant 
operation. The changes do not alter the 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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change revises the introductory paragraph of 
Specification 5.5.7, ``Ventilation Filter Testing Program (VFTP),'' 
to be consistent with the ISTS. The proposed change will not reduce 
a margin of safety because it has no effect on any safety analysis 
assumption. In addition, no requirements are being removed, but are 
being replaced with references to an NRC Regulatory Guide and the 
requirements of SR 3.0.1.
    Therefore, this proposed change does not involve a significant 
reduction in a margin of safety.
    Based on the above, SNC concludes that the proposed amendment 
does not involve a significant hazards consideration under the 
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding 
of ``no significant hazards consideration'' is justified.

    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: Jennifer M. Buettner, Associate General 
Counsel, Southern Nuclear Operating Company, 40 Inverness Center 
Parkway, Birmingham, AL 35201
    NRC Branch Chief: Robert J. Pascarelli.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama
    Date of amendment request: December 11, 2014 (ADAMS Accession No. 
ML14349A694).
    Description of amendment request: The amendment would revise 
Section 3.8.3, ``Diesel Fuel Oil, Lube Oil, and Starting Air,'' of the 
Technical Specifications (TSs) by replacing the current volume 
requirements with the number of continuous days the diesel generators 
(DGs) are required to run. The numerical volumes will be maintained in 
the licensee-controlled TSs Bases document so they may be modified 
under licensee control. The resulting requirements will specify an 
inventory of stored diesel fuel oil and lube oil sufficient for a 7-day 
supply for each DG. This proposed amendment is consistent with NRC's 
approved Technical Specifications Task Force (TSTF) Improved Standard 
Technical Specifications Change Traveler TSTF-501, Revision 1, 
``Relocate Stored Fuel Oil and Lube Oil Volume Values to Licensee 
Control.'' The availability of this TSs improvement was announced in 
the Federal Register on May 26, 2010 (75 FR 29588). The licensee also 
proposed additional changes to Section 3.8.3 and Section 5.5.9, 
``Diesel Fuel Oil Testing Program,'' to support other related changes 
proposed by TSTF-501, Revision 1. These additional changes concern fuel 
oil quality and associated surveillance requirements (SRs).
    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 changes to TS Section 3.8.3, Conditions A and B, 
and to SR 3.8.3.1 and SR 3.8.3.2 remove the volume of diesel fuel 
oil and lube oil required to support 7-day operation of each onsite 
diesel generator, and the volume equivalent to a 6-day supply, from 
the TS and replace them with the associated number of days. The 
numerical volumes will be maintained under licensee control. The 
specific volume of fuel oil equivalent to a 7 and 6-day supply is 
calculated using the NRC-approved methodology described in 
Regulatory Guide 1.137, Revision 1, ``Fuel-Oil Systems for Standby 
Diesel Generators'' and ANSI [American National Standards 
Institute]-N195 1976, ``Fuel Oil Systems for Standby Diesel-
Generators.'' The specific volume of lube oil equivalent to a 7-day 
and 6-day supply is based on the diesel generator manufacturer's 
consumption values for the run time of the diesel generator. Because 
the requirement to maintain a 7-day supply of diesel fuel oil and 
lube oil is not changed and is consistent with the assumptions in 
the accident analyses, and the actions taken when the volume of fuel 
oil and lube oil are less than a 6-day supply have not changed, 
neither the probability nor the consequences of any accident 
previously evaluated will be affected.
    The addition of a new Condition D provides a required action and 
completion time if new fuel oil properties are not within limits. 
The new SR 3.8.3.5 requires checking for and removing water from the 
7-day storage tank every 31 days. The revised Section 5.5.9 adds 
testing requirements for new fuel oil to be completed prior to the 
addition of the new fuel oil to the 7-day storage tank, as well as 
additional testing to be completed prior or within 31 days of the 
addition. These requirements are more restrictive testing 
requirements and provide corrective action to be taken if the 
testing limits are not met. They are taken from the current NRC 
approved NUREG-1433, Revision 4, ``Standard Technical 
Specifications, General Electric BWR/4 Plants.'' Improved, more 
restrictive testing standards will neither change the probability or 
the consequences of any accident previously evaluated be affected.
    Therefore, the proposed changes do not involve a significant 
increase in the

[[Page 17105]]

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 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. The change 
does not alter assumptions made in the safety analysis but ensures 
that the diesel generator operates as assumed in the accident 
analysis. The proposed change is consistent with the safety analysis 
assumptions.
    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 to Section 3.8.3, Conditions A and B, and 
to SR 3.8.3.1 and SR 3.8.3.2 remove the numerical volume of diesel 
fuel oil and lube oil required to support 7-day operation of each 
onsite diesel generator, and the numerical volume equivalent to a 6-
day supply from the TS and replaces them with the associated number 
of days. The numerical volumes will be maintained under licensee 
control. As the bases for the existing limits on diesel fuel oil 
volume and lube oil volume are not changed, no change is made to the 
accident analysis assumptions and no margin of safety is reduced as 
part of this change.
    The new, more restrictive, testing requirements, and the 
provision for corrective action to be taken if the testing limits 
are not met, are taken from the current NRC approved NUREG-1433, 
Revision 4, ``Standard Technical Specifications, General Electric 
BWR/4 Plants.'' These changes do not revise the accident analysis 
assumptions and no margin of safety is reduced as part of these 
changes.
    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Branch Chief: Shana R. Helton.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas
    Date of amendment request: November 20, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14330A247.
    Description of amendment request: The amendment would revise the 
Technical Specification (TS) requirements to address NRC Generic Letter 
2008-01, ``Managing Gas Accumulation in Emergency Core Cooling, Decay 
Heat Removal, and Containment Spray Systems,'' as described in 
Technical Specification Task Force (TSTF) Traveler TSTF-523, Revision 
2, ``Generic Letter 2008-01, Managing Gas Accumulation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises or adds SRs [surveillance 
requirements] that require verification that the Emergency Core 
Cooling System (ECCS), the Residual Heat Removal (RHR) System, and 
the Containment Spray System are not rendered inoperable due to 
accumulated gas and to provide allowances which permit performance 
of the revised verification. Gas accumulation in the subject systems 
is not an initiator of any accident previously evaluated. As a 
result, the probability of any accident previously evaluated is not 
significantly increased. The proposed SRs ensure that the subject 
systems continue to be capable to perform their assumed safety 
function and are not rendered inoperable due to gas accumulation. 
Thus, the consequences of any accident previously evaluated are not 
significantly increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change revises or adds SRs that require 
verification that the ECCS, the RHR System, and the Containment 
Spray System are not rendered inoperable due to accumulated gas and 
to provide allowances which permit performance of the revised 
verification. The proposed 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 proposed change does not impose any new 
or different requirements that could initiate an accident. The 
proposed change does not alter assumptions made in the safety 
analysis and is consistent with the safety analysis assumptions.
    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 revises or adds SRs that require 
verification that the ECCS, the RHR System, and the Containment 
Spray System are not rendered inoperable due to accumulated gas and 
to provide allowances which permit performance of the revised 
verification. The proposed change adds new requirements to manage 
gas accumulation in order to ensure the subject systems are capable 
of performing their assumed safety functions. The proposed SRs are 
more comprehensive than the current SRs and will ensure that the 
assumptions of the safety analysis are protected. The proposed 
change does not adversely affect any current plant safety margins or 
the reliability of the equipment assumed in the safety analysis. 
Therefore, there are no changes being made to any safety analysis 
assumptions, safety limits or limiting safety system settings that 
would adversely affect plant safety as a result of the proposed 
change.
    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: Jay Silberg, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
    NRC Branch Chief: Michael T. Markley.

III. Previously Published Notices 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 following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

[[Page 17106]]

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington
    Date of amendment request: August 22, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14237A729.
    Brief description of amendment request: The proposed amendment 
would revise the technical specification (TS) surveillance requirement 
(SR) for the ultimate heat sink (UHS) to clarify that spray pond level 
is the average of the level in both ponds. The design of the ultimate 
heat sink is such that it is difficult to meet the current SR when only 
one standby service water (SW) pump is in operation without overflowing 
a spray pond resulting in a net loss of water inventory, which may 
challenge the ability of the UHS to provide sufficient inventory for 30 
days. However, if the SR is not met, a plant shutdown is required.
    Date of publication of individual notice in Federal Register: 
September 5, 2014 (79 FR 53085).
    Expiration date of individual notice: October 6, 2014 (public 
comments); November 4, 2014 (hearing requests).

IV. 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 can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan
    Date of amendment request: April 23, 2013, as supplemented by 
letters dated June 19, and October 13, 2014.
    Brief description of amendment: The amendment revised the Fermi 2 
technical specification (TS) surveillance requirements (SRs) associated 
with SR 3.8.4.2 and SR 3.8.4.5 to add a battery resistance limit; SR 
3.8.6.3 to change the average electrolyte temperature of representative 
cells, and SR 3.8.4.8 to change the frequency of battery capacity 
testing.
    Date of issuance: March 16, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 199. A publicly-available version is in ADAMS under 
Accession No. ML15057A297; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-43: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2014 (79 FR 
42542). The supplemental letters dated June 19, and October 13, 2014, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 16, 2015.
    No significant hazards consideration comments received: No.
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc., 
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana
    Date of amendment request: June 13, 2013, as supplemented by 
letters dated August 28 and November 3, 2014, and January 22, 2015.
    Brief description of amendment: The amendment revised the Technical 
Specifications to risk-inform requirements regarding selected Required 
Action end states by adopting Technical Specification Task Force 
(TSTF)-423, Revision 1, ``Technical Specifications End States, NEDC-
32998-A,'' with some deviations as approved by the NRC staff. This 
technical specification improvement is part of the Consolidated Line 
Item Improvement Process (CLIIP). In addition, it approves a change to 
the facility operating license for the River Bend Station, Unit 1. The 
change deletes two license conditions that are no longer applicable and 
adds a new license condition for maintaining commitments required for 
the approval of this TSTF into the Updated Safety Analysis Report.
    Date of issuance: February 17, 2015.
    Effective date: As of the date of issuance and shall be implemented 
90 days from the date of issuance.
    Amendment No.: 185. A publicly-available version is in ADAMS under 
Accession No. ML14106A167; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-47: The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: August 20, 2013 (78 FR 
51226). The supplemental letters dated August 28, and November 3, 2014, 
and January 22, 2015, provided additional information that clarified 
the application, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 17, 2015.
    No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit 3, Westchester County, New York
    Date of amendment request: February 4, 2014, as supplemented by 
letter dated December 9, 2014.
    Brief description of amendment: The amendment revised Technical 
Specification 5.5.15, ``Containment Leakage Rate Testing Program,'' to 
allow a permanent extension of the Type A primary containment 
integrated leak

[[Page 17107]]

rate test frequency from once every 10 years to once every 15 years.
    Date of issuance: March 13, 2015.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 256. A publicly-available version is in ADAMS under 
Accession No. ML15028A308; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-64: The amendment revised the 
Facility Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2014 (79 FR 
38587). The supplemental letter provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's original 
proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 13, 2015.
    No significant hazards consideration comments received: No
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point 
Nuclear Generating Unit 3, Westchester County, New York
    Date of amendment request: April 1, 2014.
    Brief description of amendment: The amendment revised Technical 
Specification Figures 3.4.3-1, ``Heatup Limitations for Reactor Coolant 
System,'' 3.4.3-2, ``Cooldown Limitations for Reactor Coolant System,'' 
and 3.4.3-3, ``Hydrostatic and Inservice Leak Testing Limitations for 
Reactor Coolant System'' to address vacuum fill operations in the TSs. 
The proposed changes clarify that the figures are applicable for vacuum 
fill conditions where pressure limits are considered to be met for 
pressures that are below 0 pounds per square inch gauge (psig) (i.e., 
up to and including full vacuum conditions). Vacuum fill operations for 
the RCS can result in system pressures below 0 psig.
    Date of issuance: March 6, 2015.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 255. A publicly-available version is in ADAMS under 
Accession No. ML15050A144; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. DPR-64: The amendment revised the 
Facility Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: October 28, 2014 (79 FR 
64223).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 6, 2015.
    No significant hazards consideration comments received: No
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts
    Date of amendment request: April 5, 2013, as supplemented by letter 
dated March 20, 2014.
    Brief description of amendment: This amendment revised Technical 
Specification (TS) 2.1.1 and 2.1.2, ``Safety Limits,'' by reducing the 
reactor steam dome pressure from 785 pounds per square inch gauge 
(psig) to 685 psig to resolve the Pressure Regulator Failure-Open 
transient.
    Date of issuance: March 12, 2015.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days of issuance.
    Amendment No.: 242. A publicly-available version is in ADAMS under 
Accession No. ML14272A070; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-35: Amendment revised 
the License and TS.
    Date of initial notice in Federal Register: August 6, 2013 (78 FR 
47788). The supplement dated March 20, 2014, provided additional 
information that clarified the application, did not expand the scope of 
the application as originally noticed, and did not change the NRC 
staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 2015.
    No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket No. 50-220, Nine Mile Point 
Nuclear Station, Unit 1, Oswego County, New York
    Date of application for amendment: March 8, 2013, as supplemented 
by letter dated May 16, 2013, July 8, July16, August 29, 2014, and 
January 22, 2015. The public versions of these documents are available 
in ADAMS at the Accession Nos. ML13073A103, ML13144A068, ML14203A050, 
ML14199A384, ML14251A233, and ML15026A132, respectively.
    Brief description of amendment: The amendment to the Nine Mile 
Point Unit 1 (NMP1) Renewed Facility Operating License DPR-63 modified 
Technical Specification (TS) Table 3.6.2i, ``Diesel Generator 
Initiation,'' by revising the existing 4.16kV Power Board (PB) 102/103 
Emergency Bus Undervoltage (Degraded Voltage) Operating Time value and 
by updating the Set Point heading title. The TS revisions are being 
made to resolve the green non-cited violation (NCV) associated with the 
vital bus degraded voltage protection time delay documented in NRC 
Inspection Report (IR) 05000220/201101, ``Nine Mile Point Nuclear 
Station--NRC Unresolved Item Follow-up Inspection Report,'' dated 
January 23, 2012 (ADAMS Accession No. ML12023A119), specifically, 
NCV05000220/20 11011-01, ``Vital Bus Degraded Voltage Time Delay Not 
Maintained within LOCA Analysis Assumptions.''
    Date of issuance: March 12, 2015.
    Effective date: effective as of the date of its issuance and shall 
be implemented within 60 days.
    Amendment No.: 217.
    Renewed Facility Operating License No. DPR-63: Amendment revised 
the License and Technical Specifications.
    Date of initial notice in Federal Register: June 11, 2013, (78 FR 
35062). The supplements dated May 16, 2013, July 8, July16, August 29, 
2014, and January 22, 2015, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the NRC staff's initial 
proposed no significant hazards consideration determination noticed in 
the Federal Register on June 11, 2013 (78 FR 35062).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 2015.
    No significant hazards consideration comments received: No
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania
    Date of application for amendments: July 11, 2014, as supplemented 
by letter dated December 1, 2014.
    Brief description of amendments: The amendments incorporate several 
administrative changes to the Facility Operating Licenses (FOLs) and 
the Technical Specifications (TSs) such as deleting historical items 
that are no longer applicable, correcting errors, and removing 
references that are no longer valid.
    Date of issuance: March 11, 2015.

[[Page 17108]]

    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendments Nos.: 296 and 299. A publicly-available version is in 
ADAMS under Accession No. ML14363A227; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the FOLs and the TSs.
    Date of initial notice in Federal Register: September 2, 2014 (79 
FR 52062). The supplemental letter dated December 1, 2014, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 11, 2015.
    No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Units 1 and 2 (BVPS-1 and 2), 
Beaver County, Pennsylvania
    Date of amendment request: October 18, 2013, as supplemented by 
letters dated June 26, 2014, September 21, 2014, and February 4, 2015.
    Brief description of amendments: The amendment changes the Beaver 
Valley Power Station Technical Specifications (TS). Specifically, this 
change request involves the adoption of an approved change to the 
standard TS for Westinghouse plants (NUREG-1431), to allow relocation 
of specific TS surveillance frequencies to a licensee-controlled 
program. The proposed change is described in TS Task Force (TSTF) 
Traveler, TSTF-425, Revision 3, ``Relocation Surveillance Frequencies 
to Licensee Control--RITSTF [Risk-Informed Technical Specifications 
Task Force] Initiative 5b'' (Agencywide Documents Access and Management 
System (ADAMS) Accession No. ML090850642). A Notice of Availability was 
published in the Federal Register on July 6, 2009 (74 FR 31996).
    The proposed change relocates surveillance frequencies to a 
licensee-controlled program, the Surveillance Frequency Control 
Program. This change is applicable to licensees using probabilistic 
risk guidelines contained in NRC-approved NEI 04-10, Revision 1, 
``Risk-Informed Technical Specifications Initiative 5b, Risk-Informed 
Method for Control of Surveillance Frequencies'' (ADAMS Accession No. 
ML071360456).
    Date of issuance: March 6, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 292 and 179. A publicly-available version is in 
ADAMS under Accession No. ML14322A461; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-66 and NPF-73: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2014 (79 FR 
3416). The supplemental letters dated June 26, 2014, September 21, 
2014, and February 4, 2015, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 6, 2015.
    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 amendment request: November 21, 2013, and supplemented by 
the letters dated March 5 and June 30, 2014.
    Brief description of amendment: The amendment authorizes changes to 
the VEGP Units 3 and 4 Updated Final Safety Analysis Report to revise 
the details of the effective thermal conductivity resulting from the 
oxidation of the inorganic zinc component of the containment vessel 
coating system.
    Date of issuance: February 26, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 31. A publicly-available version is in ADAMS under 
Accession No. ML15028A358; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: March 18, 2014 (79 FR 
15150).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 26, 2015.
    No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama
    Date of application for amendment: September 25, 2012; as 
supplemented on December 20, 2012; September 16, October 30, and 
November 12, 2013; April 23, May 23, July 3, August 11, August 29, and 
October 13, 2014; and January 16, 2015.
    Brief description of amendments: The amendment authorizes the 
transition of the Joseph M. Farley Nuclear Plant, Units 1 and 2, fire 
protection program to a risk-informed, performance-based program based 
on National Fire Protection Association (NFPA) 805, ``Performance-Based 
Standard for Fire Protection for Light Water Reactor Electric 
Generating Plants, 2001 Edition'' (NFPA 805), in accordance with 10 CFR 
50.48(c).
    Date of issuance: March 10, 2015.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: Unit 1-196, Unit 2-192. A publicly-available 
version is in ADAMS under Accession No. ML14308A048, documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-2 and NPF-8: The amendments 
revised the Renewed Facility Operating Licenses and Technical 
Specifications.
    Date of initial notice in Federal Register: March 12, 2013 (78 FR 
15750). The supplemental letters dated September 16, October 30, and 
November 12, 2013; April 23, May 23, July 3, August 11, August 29, and 
October 13, 2014; and January 16, 2015, provided additional information 
that clarified the application, did not expand the scope of the 
application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination as 
published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 10, 2015.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 23rd day of March 2015.


[[Page 17109]]


    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2015-07192 Filed 3-30-15; 8:45 am]
BILLING CODE 7590-01-P