Biweekly Notice; Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 5798-5816 [2015-01917]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 80, Number 22 (Tuesday, February 3, 2015)]
[Notices]
[Pages 5798-5816]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2015-01917]


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NUCLEAR REGULATORY COMMISSION

[NRC-2015-0015]


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 January 8, 2015, to January 21, 2015. The 
last biweekly notice was published on January 20, 2015.

DATES: Comments must be filed by March 5, 2015. A request for a hearing 
must be filed by April 6, 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-2015-0015. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: Carol.Gallagher@nrc.gov.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: 3WFN-06-A44M, 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: Beverly A. Clayton, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001; telephone: 301-415-3475, email: Beverly.Clayton@nrc.gov.

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2015-0015 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-2015-0015.
     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-2015-0015 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC posts all comment submissions at 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 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

[[Page 5799]]

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 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 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-

[[Page 5800]]

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.
Energy Northwest, Docket No. 50-397, Columbia Generating Station 
(Columbia), Benton County, Washington
    Date of amendment request: November 17, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14336A100.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications to revise values for the safety 
limit minimum critical power ratio (SLMCPR) due to core loading fuel 
management changes for the upcoming Columbia operating cycle.
    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 basis of the Safety Limit Minimum Critical Power Ratio 
(SLMCPR) is to ensure no mechanistic fuel damage is calculated to 
occur if the limit is not violated. The new SLMCPR values preserve 
the existing margin to transition boiling. The derivation of the 
revised SLMCPR for Columbia, for incorporation into the Technical 
Specifications and its use to determine plant and cycle-specific 
thermal limits, has been performed using NRC approved methods. The 
revised SLMCPR values do not change the method of operating the 
plant and have no effect on the probability of an accident 
initiating event or transient.
    Based on the above, Energy Northwest has concluded that the 
proposed change will not result in 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 analyzed?
    Response: No.
    The proposed changes result only from a specific analysis for 
the Columbia core reload design. These changes do not involve any 
new or different methods for operating the facility. No new 
initiating events or transients result from these changes.
    Based on the above, Energy Northwest has concluded that the 
proposed change will not create the possibility of a new or 
different kind of accident from those previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The new SLMCPR is calculated using NRC approved methods with 
plant and cycle specific parameters for the current core design. The 
SLMCPR value remains conservative enough to ensure that at least 
99.9% of all fuel rods in the core will avoid transition boiling if 
the limit is not violated,

[[Page 5801]]

thereby preserving the fuel cladding integrity. The operating limit 
minimum critical power ratio (MCPR) is established to ensure that no 
fuel damage results during anticipated operational occurrences 
(AOOs). Accordingly, the margin of safety is maintained with the 
revised values.
    As a result, Energy Northwest has determined that the proposed 
change will not result in a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 
1700 K Street NW., Washington, DC 20006-3817.
    Acting NRC Branch Chief: Eric R. Oesterle.
Florida Power & Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Units 1 and 2, St. Lucie County, Florida
    Date of amendment request: December 5, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14351A074.
    Description of amendment request: The amendment would revise 
Technical Specifications (TSs) Section 3.6.2.1, regarding containment 
spray and cooling systems, by eliminating second completion times 
limiting time from discovery of failure to meet a limiting condition 
for operation (LCO). The proposed revision is consistent with NRC-
approved Technical Specifications Task Force (TSTF) Traveler TSTF-439, 
Revision 2, ``Eliminate Second Completion Times Limiting Time from 
Discovery of Failure to Meet an LCO'' (Adams Accession No. 
ML051860296).
    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 any accident previously 
evaluated?
    Response: No.
    The proposed change that incorporated TSTF-439, Revision 2, 
[will eliminate] certain Completion Times from the TS. 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 revised 
Completion Times are no different [from] the consequences of the 
same accident during the existing Completion Times. As a result, the 
consequences of an accident previously evaluated are not affected by 
this change. The proposed change does not alter or prevent the 
ability of structures, systems, or components (SSCs) from performing 
their intended function to mitigate the consequences of an 
initiating event within the assumed acceptance limits.
    The proposed change does not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
Further, the proposed change does not increase the types or amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures. The proposed change is consistent with the 
[previous] safety analysis assumptions and resultant consequences. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    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. The proposed change does not alter any assumptions made 
in the safety analysis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed change to delete the second Completion Times 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 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.

    The NRC staff has reviewed the licensee's analysis and determined 
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: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Blvd., MS LAW/JB, 
Juno Beach, FL 33408-0420.
    NRC Branch Chief: Shana R. Helton.
Omaha Public Power District (OPPD), Docket No. 50-285, Fort Calhoun 
Station, Unit 1, Washington County, Nebraska
    Date of amendment request: December 26, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14365A123.
    Description of amendment request: The proposed amendment upgrades 
the Emergency Action Level (EAL) scheme by adopting NRC-endorsed 
Nuclear Energy Institute (NEI) 99-01, Revision 6, ``Methodology for the 
Development of Emergency Action Levels for Non-Passive Reactors,'' 
issued January 2011 (ADAMS Accession No. ML110240324).
    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 changes to OPPD's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, ``Development of 
Emergency Action Levels for Non-Passive Reactors,'' do not reduce 
the capability to meet the emergency planning requirements 
established in 10 CFR 50.47 and 10 CFR 50, Appendix E. The proposed 
changes do not reduce the functionality, performance, or capability 
of OPPD's ERO [emergency response organization] to respond in 
mitigating the consequences of any design basis accident.
    The probability of a reactor accident requiring implementation 
of Emergency Plan EALs has no relevance in determining whether the 
proposed changes to the EALs reduce the effectiveness of the 
Emergency Plans. As discussed in Section D, ``Planning Basis,'' of 
NUREG-0654, Revision 1, ``Criteria for Preparation and Evaluation of 
Radiological Emergency Response Plans and Preparedness in Support of 
Nuclear Power Plants'' [issued November 1980; ADAMS Accession No. 
ML040420012]:
    . . . The overall objective of emergency response plans is to 
provide dose savings (and in some cases immediate life saving) for a 
spectrum of accidents that could produce offsite doses in excess of 
Protective Action Guides (PAGs). No single specific accident 
sequence should be isolated as the one for which to plan because 
each accident could have different consequences, both in nature and 
degree. Further, the range of possible selection for a planning 
basis is very large, starting with a zero point of requiring no 
planning at all because significant offsite radiological accident 
consequences are unlikely to occur, to planning for the worst 
possible accident, regardless of its extremely low likelihood . . .
    Therefore, OPPD did not consider the risk insights regarding any 
specific accident

[[Page 5802]]

initiation or progression in evaluating the proposed changes.
    The proposed changes do not involve any physical changes to 
plant equipment or systems, nor do they alter the assumptions of any 
accident analyses. The proposed changes do not adversely affect 
accident initiators or precursors nor do they alter the design 
assumptions, conditions, and configuration or the manner in which 
the plant is operated and maintained. The proposed changes do not 
adversely affect the ability of Structures, Systems, or Components 
(SSCs) to perform their intended safety functions in mitigating 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 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 to OPPD's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not involve any 
physical changes to plant systems or equipment. The proposed changes 
do not involve the addition of any new plant equipment. The proposed 
changes will not alter the design configuration, or method of 
operation of plant to be performed as required. The proposed changes 
do not create any new credible failure mechanisms, malfunctions, or 
accident initiators.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from those that have been 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to OPPD's EAL scheme to adopt the NRC-
endorsed guidance in NEI 99-01, Revision 6, do not alter or exceed a 
design basis or safety limit. There is no change being made to 
safety analysis assumptions, safety limits, or limiting safety 
system settings that would adversely affect plant safety as a result 
of the proposed change. There are no changes to setpoints or 
environmental conditions of any SSC or the manner in which any SSC 
is operated. Margins of safety are unaffected by the proposed 
changes to adopt the NEI 99-01, Revision 6, EAL scheme guidance. The 
applicable requirements of 10 CFR 50.47 and 10 CFR 50, Appendix E 
will continue to be met.
    Therefore, the proposed changes do not involve any 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 A. Repka, Esq., Winston & Strawn, 1700 
K Street NW., Washington, DC 20006-3817.
    Acting NRC Branch Chief: Eric R. Oesterle.
South Carolina Electric and Gas Company Docket Nos.: 52-027 and 52-028, 
Virgil C. Summer Nuclear Station, Units 2 and 3, Fairfield County, 
South Carolina
    Date of amendment request: July 17, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14202A088.
    Description of amendment request: The proposed changes would revise 
the Combined Licenses (COLs) by (1) providing additional detail to 
describe the mechanical connection between the internal containment 
structural module steel faceplates and the base concrete, (2) allowing 
for increases in the thickness of the structural wall module 
faceplates, (3) identifying changes to the wall thicknesses for 
portions of some internal containment structural wall modules, and (4) 
identifying the use of steel plates, structural shapes, reinforcement 
bars, or tie bars between the faceplates of the structural wall 
modules, where needed to meet applicable code requirements.
    Because this proposed change requires a departure from Tier 1 
information in the Westinghouse Advanced Passive 1000 Design Control 
Document (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 design function of the internal containment structures is to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in those structures. 
These structures are structurally designed to meet seismic Category 
I requirements as defined in Regulatory Guide 1.29.
    The changes to the design details for the structural modules do 
not have an adverse impact on the response of the nuclear island 
structures to safe shutdown earthquake ground motions or loads due 
to anticipated transients or postulated accident conditions, nor do 
they change the seismic Category I classification. Evaluations have 
been performed which determined that the proposed changes do not 
have a significant impact on the calculated loads for the affected 
structural modules, or critical locations, and no significant impact 
on the global seismic model. The changes to the design details for 
the structural modules do not impact the support, design, or 
operation of mechanical and fluid systems. There is no change to 
plant systems or the response of systems to postulated accident 
conditions. There is no change to the predicted radioactive releases 
due to postulated accident conditions. The plant response to 
previously evaluated accidents or external events is not adversely 
affected, nor does the change described create any new accident 
precursors.
    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 changes are to revise design details for the 
internal containment structural modules. The changes do not change 
the design requirements of the nuclear island structures, nor do 
they change the seismic Category I classification. The changes to 
the design details for the internal containment structural modules 
do not change the design function, support, design, or operation of 
mechanical and fluid systems. The changes to the design details for 
the internal containment structural modules do not result in a new 
failure mechanism for the nuclear island structures or introduce any 
new accident precursors. As a result, the design function of the 
nuclear island structures is not adversely affected by the proposed 
change.
    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 requested amendment proposes changes to the structural 
details associated with the in-containment structural modules. The 
purpose of these changes is to ensure that the requirements 
contained in the applicable construction codes are met. As discussed 
in UFSAR [Updated Final Analysis Report], Section 3.8.3.5, ``Design 
Procedures and Acceptance Criteria,'' the in-containment structural 
modules are designed in accordance with ACI [American Concrete 
Institute] 349 and AISC [American Institute of Steel Construction] 
N690. Thus, the identification of additional structural module 
connection details, the increase in structural module faceplate and 
wall thicknesses, and the addition of additional reinforcement in 
specific areas are proposed to ensure that the codes of record, and 
the associated margins contained therein, continue to be met as 
specified in the design basis. Structural and seismic analysis of 
the modified sections in accordance with the methodologies 
identified in the UFSAR has confirmed that the applicable 
requirements of ACI 349 and AISC N690 continue to be met for 
affected in-containment structural modules.
    As a result, the proposed changes do not adversely affect any 
safety related equipment

[[Page 5803]]

or other design functions, design code compliance, 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.
    Therefore, the requested 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 J. Burkhart.
South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 1, 
Fairfield County, South Carolina
    Date of amendment request: December 19, 2014. A publicly-available 
version is in ADAMS Package Accession No. ML14363A422.
    Description of amendment request: The licensee proposes to expand 
the emergency planning zone (EPZ) boundary, to revise the evacuation 
time estimates (ETA) analysis, and revise the alert and notification 
system (ANS) design reports to encompass the expanded EPZ boundary.
    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 changes, which include expansion of the EPZ 
boundary and revision of the ETE analysis and ANS design reports to 
encompass the expanded EPZ boundary, do not impact the physical 
function of plant structures, systems, or components (SSC) or the 
manner in which SSCs perform their design function. The proposed 
changes neither adversely affect accident initiators or precursors, 
nor alter design assumptions. The proposed changes do not alter or 
prevent the ability of SSCs to perform their intended function to 
mitigate the consequences of an initiating event within assumed 
acceptance limits. No operating procedures or administrative 
controls that function to prevent or mitigate accidents are affected 
by the proposed changes. Therefore, the proposed changes 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 changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be installed 
or removed) or a change in the method of plant operation. The 
proposed changes will not introduce failure modes that could result 
in a new accident, and the change does not alter assumptions made in 
the safety analysis. The proposed changes, which include expansion 
of the EPZ boundary and revision of the ETE analysis and ANS design 
reports to encompass the expanded EPZ boundary, are not initiators 
of any accidents. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is associated with the ability of the fission 
product barriers (i.e., fuel cladding, reactor coolant system 
pressure boundary, and containment structure) to limit the level of 
radiation dose to the public. The proposed changes, which include 
expansion of the EPZ boundary and revision of the ETE analysis and 
ANS design reports to encompass the expanded EPZ boundary, do not 
impact operation of the plant or its response to transients or 
accidents. The proposed changes do not alter requirements of the 
Technical Specifications or the Unit 1 Operating License. The 
proposed changes do not involve a change in the method of plant 
operation and no accident analyses will be affected by the proposed 
changes.
    Additionally, the proposed changes will 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 these proposed changes. The proposed changes will not result in 
plant operation in a configuration outside the design basis. The 
proposed changes do not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition.

    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: J. Hagood Hamilton, Jr., South Carolina 
Electric & Gas Company, Post Office Box 764, Columbia, SC 29218.
    NRC Branch Chief: Robert J. Pascarelli.
Southern Nuclear Operating Company, Inc. Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, 
Georgia
    Date of amendment request: January 8, 2015. A publicly-available 
version is in ADAMS under Accession No. ML15008A466.
    Description of amendment request: The proposed change would amend 
Combined License Nos. NPF-91 and NPF-92 for the VEGP, Units 3 and 4 by 
departing from the plant-specific Design Control Document (DCD) Tier 1 
(and corresponding Combined License Appendix C information) and Tier 2 
material by making changes to specify the use of latching control 
relays in lieu of breakers to de-energize the control rod drive 
mechanism (CRDM) motor generator (MG) set generator field on a diverse 
actuation system (DAS) signal.
    Because this proposed change requires a departure from Tier 1 
information in the Westinghouse 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 change to use field control relays in lieu of field 
circuit breakers to de-energize the CRDM MG Set excitation field 
does not result in a change to the basic MG Set design function, 
which is to supply reliable electrical power to the CRDMs while 
providing a trip function on a DAS signal, allowing the control rods 
to drop. The Probabilistic Risk Assessment (PRA) is not adversely 
affected. No safety-related structure, system, or component (SSC) or 
function is adversely affected. The change does not involve nor 
interface with any SSC accident initiator or initiating sequence of 
events, and thus, the probabilities of the accidents evaluated in 
the UFSAR are not affected. Because the change maintains the CRDM MG 
set trip function used to mitigate an accident, the consequences of 
the accidents evaluated in the UFSAR are not affected.
    Therefore, there is no 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.

[[Page 5804]]

    There is no safety-related SSC or function adversely affected by 
this proposed change to use control relays instead of breakers to 
de-energize the CRDM MG set generator field on demand. This proposed 
change does not change any equipment qualification or fission 
product barrier. The change does not result in a new failure mode, 
malfunction or sequence of events that could affect safety or 
safety-related equipment. This activity will not allow for a new 
fission product release path, result in a new fission product 
barrier failure mode, or create a new sequence of events that would 
result in significant fuel cladding failures.
    Therefore, this activity 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.
    There is no safety-related SSC or function adversely affected by 
this proposed change to use relays instead of breakers to control 
the CRDM MG set generator field. The function to trip the MG set 
generator field on a DAS signal, allowing the control rods to drop, 
is not adversely affected by the use of relays as the device to de-
energize the generator field. The proposed change does not affect 
any safety-related design code, 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 requested change, 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: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Branch Chief: Lawrence Burkhart.
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 amendment request: November 24, 2014. A publicly-available 
version is in ADAMS under Accession Package No. ML14335A689.
    Description of amendment request: The licensee requested 24 
revisions to the Technical Specifications. Twenty two revisions adopt 
various previously NRC approved Technical Specifications Task Force 
Travelers and two revisions are not associated with Travelers. 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:

Request No. 1: TSTF-27-A, Revision 3, ``Revise SR Frequency for 
Minimumn Temperature for Criticality''

    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 Surveillance Frequency for 
monitoring RCS temperature to ensure the minimum temperature for 
criticality is met. The Frequency is changed from a 30 minute 
Frequency when certain conditions are met to a periodic Frequency 
that it is controlled in accordance with the Surveillance Frequency 
Control Program. The measurement of RCS [reactor coolant system] 
temperature is not an initiator of any accident previously 
evaluated. The minimum RCS temperature for criticality is not 
changed. As a result, the mitigation of any accident previously 
evaluated is not affected.
    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 the Surveillance Frequency for 
monitoring RCS temperature to ensure the minimum temperature for 
criticality is met. The current, condition based Frequency 
represents a distraction to the control room operator during the 
critical period of plant startup. RCS temperature is closely 
monitored by the operator during the approach to criticality and 
temperature is recorded on charts and computer logs. Allowing the 
operator to monitor temperature as needed by the situation and 
logging RCS temperature at a periodic Frequency that it is 
controlled in accordance with the Surveillance Frequency Control 
Program is sufficient to ensure that the LCO [limiting condition for 
operation] is met while eliminating a diversion of the operator's 
attention.
    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.

Request No. 2: TSTF-46-A, Revision 1, ``Clarify the CIV Surveillance to 
Apply Only to Automatic Isolation Valves''

    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 SR 3.6.3.4, and the associated Bases, to delete the 
reference to verifying the isolation time of ``each power operated'' 
containment isolation valve (CIV) and only require verification of 
each ``automatic power operated containment isolation valve.'' The 
closure times for CIVs 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 CIVs are used to respond 
to accidents previously evaluated. Power operated CIVs 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 CIVs provide plant protection or introduce any new or 
different operational conditions. Periodic verification that the 
closure times for CIVs that receive an automatic closure signal are 
within the limits established by the accident analysis will continue 
to be performed under SR 3.6.3.4. 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 CIVs that 
receive an automatic

[[Page 5805]]

isolation signal are within the scope of the SR 3.6.3.4. The 
proposed change does not result in a change in the manner in which 
the CIVs provide plant protection. Periodic verification that 
closure times for CIVs 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.

    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.

Request No. 3: TSTF-87-A, Revision 2, ``Revise ``RTBs Open'' and ``CRDM 
Deenergized'' Actions to ``Incapable of Rod Withdrawal''

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change revises the Required Actions for LCO 3.4.5, ``RCS 
Loops--Mode 3,'' Conditions C.2 and D.1, from ``De-energize all 
control rod drive mechanisms,'' to ``Place the Rod Control System in 
a condition incapable of rod withdrawal.'' It also revises LCO 
3.4.9, ``Pressurizer,'' Required Action A. 1, from requiring the 
Reactor Trip Breakers to be open after reaching MODE 3 to ``Place 
the Rod Control System in a condition incapable of rod withdrawal,'' 
and to require full insertion of all rods. Inadvertent rod 
withdrawal can be an initiator for design basis accidents or events 
during certain plant conditions, and therefore must be prevented 
under those conditions. The proposed Required Actions for LCO 3.4.5 
and LCO 3.4.9 satisfy the same intent as the current Required 
Actions, which is to prevent inadvertent rod withdrawal when an 
applicable Condition is not met, and is consistent with the 
assumptions of the accident analysis. As a result, the proposed 
change does not increase the probability of any accident previously 
evaluated. The proposed change does not change how the plant would 
mitigate an accident previously evaluated as in both the current and 
proposed requirements, rod withdrawal is prohibited.
    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 provides less specific, but equivalent, 
direction on the manner in which inadvertent control rod withdrawal 
is to be prevented when the Conditions of LCO 3.4.5 and LCO 3.4.9 
are not met. Rod withdrawal will continue to be prevented when the 
applicable Conditions of LCO 3.4.5 and LCO 3.4.9 are met. There are 
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). The change 
does not alter assumptions made in the safety analysis, and is 
consistent with 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 provides the operational flexibility of 
allowing alternate, but equivalent, methods of preventing rod 
withdrawal when LCO 3.4.5 and LCO 3.4.9 are not met. 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.

    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.

Request No. 4: TSTF-245-A, Revision 1, ``AFW Train Operable When in 
Service''

    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 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to 
clarify the operability of an AFW train when it is aligned for 
manual steam generator level control. The AFW System is 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 AFW System is used to respond to accidents 
previously evaluated. The proposed change does not affect the design 
of the AFW System, and no physical 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 change does not result in a change in the manner in 
which the AFW System provides plant protection. The AFW System will 
continue to supply water to the steam generators to remove decay 
heat and other residual heat by delivering at least the minimum 
required flow rate to the steam generators. There are 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). The change does not 
alter assumptions made in the safety analysis, and is consistent 
with the safety analysis assumptions and current plant operating 
practice. Manual control of AFW level control valves is not 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 change provides the operational flexibility of 
allowing an AFW train(s) to be considered operable when it is not in 
the normal standby alignment and is temporarily incapable of 
automatic initiation, such as during alignment and operation for 
manual steam generator level control, provided it is capable of 
being manually realigned to the AFW heat removal mode of operation. 
The proposed change does not result in a change in the manner in 
which the AFW System provides plant protection. The AFW System will 
continue to supply water to the steam generators to remove decay 
heat and other residual heat by delivering at least the minimum 
required flow rate to the steam generators. 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.

    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

[[Page 5806]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

Request No. 5: TSTF-247-A, Revision 0, ``Provide Separate Condition 
Entry for Each PORV and Block Valve''

    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 3.4.11, ``Pressurizer PORVs [power operated relief 
valves],'' to clarify that separate Condition entry is allowed for 
each block valve. Additionally, the Actions are modified to no 
longer require that the PORVs be placed in manual operation when 
both block valves are inoperable and cannot be restored to operable 
status within the specified Completion Time. This preserves the 
overpressure protection capabilities of the PORVs. The pressurizer 
block valves are used to isolate their respective PORV in the event 
it is experiencing excessive leakage, and are not an initiator of 
any design basis accident or event. Therefore the proposed change 
does not increase the probability of any accident previously 
evaluated. The PORV and block valves are used to respond to 
accidents previously evaluated. The proposed change does not affect 
the design of the PORV and block valves, and no physical 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 change does not result in a change in the manner in 
which the PORV and block valves provide plant protection. The PORVs 
will continue to provide overpressure protection, and the block 
valves will continue to provide isolation capability in the event a 
PORV is experiencing excessive leakage. There are 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). The change does not 
alter assumptions made in the safety analysis, and is consistent 
with the safety analysis assumptions and current plant operating 
practice. Operation of the PORV block valves is not 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 provide clarification that separate 
Condition entry is allowed for each block valve. Additionally, the 
Actions are modified to no longer require that the PORVs be placed 
in manual operation when both block valves are inoperable and cannot 
be restored to operable status within the specified Completion Time. 
This preserves the overpressure protection capabilities of the 
PORVs. The proposed change does not result in a change in the manner 
in which the PORV and block valves provide plant protection. The 
PORVs will continue to provide overpressure protection, and the 
block valves will continue to provide isolation capability in the 
event a PORV is experiencing excessive leakage. 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.

    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.

Request No. 6: TSTF-248-A, Revision 0, ``Revise Shutdown Margin 
Definition for Stuck Rod Exception''

    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 the definition of Shutdown Margin 
to eliminate the requirement to assume the highest worth control rod 
is fully withdrawn when calculating Shutdown Margin if it can be 
verified by two independent means that all control rods are 
inserted. The method for calculating shutdown margin is not an 
initiator of any accident previously evaluated. If it can be 
verified by two independent means that all control rods are 
inserted, the calculated Shutdown Margin, without the conservatism 
of assuming the highest worth control rod is withdrawn, is accurate 
and consistent with the assumptions in the accident analysis. As a 
result, the mitigation of any accident previously evaluated is not 
affected.
    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 modifies the definition of Shutdown Margin 
to eliminate the requirement to assume the highest worth control rod 
is fully withdrawn when calculating Shutdown Margin if it can be 
verified by two independent means that all control rods are 
inserted. The additional margin of safety provided by the assumption 
that the highest worth control rod is fully withdrawn is unnecessary 
if it can be independently verified that all controls rods are 
inserted.
    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.

Request No. 7: TSTF-266-A, Revision 3, ``Eliminate the Remote Shutdown 
System Table of Instrumentation and Controls''

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change removes the list of Remote Shutdown System 
instrumentation and controls from the Technical Specifications and 
places them in the Bases. The Technical Specifications continue to 
require that the instrumentation and controls be operable. The 
location of the list of Remote Shutdown System instrumentation and 
controls is not an initiator to any accident previously evaluated. 
The proposed change will have no effect on the mitigation of any 
accident previously evaluated because the instrumentation and 
controls continue to be required to be operable.
    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

[[Page 5807]]

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 removes the list of Remote Shutdown System 
instrumentation and controls from the Technical Specifications and 
places it in the Bases. The review performed by the NRC when the 
list of Remote Shutdown System instrumentation and controls is 
revised will no longer be needed unless the criteria in 10 CFR 50.59 
are not met such that prior NRC review is required. The Technical 
Specification requirement that the Remote Shutdown System be 
operable, the definition of operability, the requirements of 10 CFR 
50.59, and the Technical Specifications Bases Control Program are 
sufficient to ensure that revision of the list without prior NRC 
review and approval does not introduce a significant safety risk.
    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.

Request No. 8: TSTF-272-A, Revision 1, ``Refueling Boron Concentration 
Clarification''

    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 the Applicability of Specification 
3.9.1, ``Boron Concentration,'' to clarify that the boron 
concentration limits are only applicable to the refueling canal and 
the refueling cavity when those volumes are attached to the Reactor 
Coolant System (RCS). The boron concentration of water volumes not 
connected to the RCS are not an initiator of an accident previously 
evaluated. The ability to mitigate any accident previously evaluated 
is not affected by the boron concentration of water volumes not 
connected to the RCS.
    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 modifies the Applicability of Specification 
3.9.1, ``Boron Concentration,'' to clarify that the boron 
concentration limits are only applicable to the refueling canal and 
the refueling cavity when those volumes are attached to the RCS. 
Technical Specification SR 3.0.4 requires that Surveillances be met 
prior to entering the Applicability of a Specification. As a result, 
the boron concentration of the refueling cavity or the refueling 
canal must be verified to satisfy the LCO prior to connecting those 
volumes to the RCS. The margin of safety provided by the refueling 
boron concentration is not affected by this change as the RCS boron 
concentration will continue to satisfy the LCO.
    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.

Request No. 9: TSTF-273-A, Revision 2, ``Safety Function Determination 
Program Clarifications''

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed TS changes add explanatory text to the programmatic 
description of the Safety Function Determination Program (SFDP) in 
Specification 5.5.15 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 LCO 3.0.6 is 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 5.5.15 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.

    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.

Request No. 10: TSTF-283-A, Revision 3, ``Modify Section 3.8 Mode 
Restriction Notes''

    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 four 
diesel generator (DG) Surveillances to allow performance of the 
Surveillance in whole or in part to reestablish DG Operability. The 
emergency diesel generators 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 
reestablish

[[Page 5808]]

Operability of a DG. The consequences of an accident previously 
evaluated during the period that the DG is being tested to 
reestablish Operability are no different from the consequences of an 
accident previously evaluated while the 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 it's assumed safety function. The proposed 
change will only allow the performance of the Surveillances to 
reestablish Operability and the proposed changes may not be used to 
remove a DG from service. In addition, the proposed change will 
potentially shorten the time that a DG is unavailable because 
testing to reestablish Operability can be performed without a plant 
shutdown. 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.

    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.

Request No. 11: TSTF-284-A, Revision 3, ``Add `Met vs. Perform' to 
Technical Specification 14, Frequency''

    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 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 provide clarification and correction of existing 
expectations 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 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.

    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.

Request No. 12: TSTF-308-A, Revision 1, ``Determination of Cumulative 
and Projected Dose Contributions in RECP''

    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.

    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.

Request No. 13: TSTF-312-A, Revision 1, ``Administrative Control of 
Containment Penetrations''

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would allow containment penetrations to be 
unisolated under administrative controls during core

[[Page 5809]]

alterations or movement of irradiated fuel assemblies within 
containment. The status of containment penetration flow paths (i.e., 
open or closed) is not an initiator for any design basis accident or 
event, and therefore the proposed change does not increase the 
probability of any accident previously evaluated. The proposed 
change does not affect the design of the primary containment, or 
alter plant operating practices such that the probability of an 
accident previously evaluated would be significantly increased. The 
proposed change does not significantly change how the plant would 
mitigate an accident previously evaluated, and is bounded by the 
fuel handling accident (FHA) 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 amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    Allowing penetration flow paths to be open is not an initiator 
for any accident. The proposed change to allow open penetration flow 
paths will not affect plant safety functions or plant operating 
practices such that a new or different accident could be created. 
There are 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). The 
change does not alter assumptions made in the safety analysis, and 
is consistent with the safety analysis assumptions and current plant 
operating practice.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    TS 3.9.3 provides measures to ensure that the dose consequences 
of a postulated FHA inside containment are minimized. The proposed 
change to LCO 3.9.3 will allow penetration flow path(s) to be open 
during refueling operations under administrative control. These 
administrative controls will provide assurance that prompt closure 
of open penetrations flow paths can and will be achieved in the 
event of an FHA inside containment, and will minimize dose 
consequences. The proposed change is bounded by the existing FHA 
analysis. 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.

    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.

Request No. 14: TSTF-314-A, Revision 0, ``Require Static and Transient 
FQ Measurement''

    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 Required Actions of 
Specification 3.1.4, ``Rod Group Alignment Limits,'' and 
Specification 3.2.4, ``Quadrant Power Tilt Ratio,'' to require 
measurement of both the steady state and transient portions of the 
Heat Flux Hot Channel Factor, FQ(Z). This change will 
ensure that the hot channel factors are within their limits when the 
rod alignment limits or quadrant power tilt ratio are not within 
their limits. The verification of hot channel factors is not an 
initiator of any accident previously evaluated. The verification 
that both the steady state and transient portion of FQ(Z) 
are within their limits will ensure this initial assumption of the 
accident analysis is met should a previously evaluated accident 
occur.
    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 the Required Actions in the 
Specifications for Rod Group Alignment Limits and Quadrant Power 
Tilt Ratio to require measurement of both the steady state and 
transient portions of the Heat Flux Hot Channel Factor, 
FQ(Z). This change is a correction that ensures that the 
plant conditions are as assumed in the accident analysis.
    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.

Request No. 15: TSTF-315-A, Revision 0, ``Reduce Plant Trips Due to 
Spurious Signals to the NIS During Physics Testing''

    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 3.1.8, ``PHYSICS TESTS 
Exceptions--MODE 2,'' to allow the number of channels required by 
LCO 3.3.1, ``RTS Instrumentation,'' to be reduced from ``4'' to 
``3'' to allow one nuclear instrumentation channel to be used as an 
input to the reactivity computer for physics testing without placing 
the nuclear instrumentation channel in a tripped condition. A 
reduction in the number of required nuclear instrumentation channels 
is not an initiator to any accident previously evaluated. With the 
nuclear instrumentation channel placed in bypass instead of in trip, 
reactor protection is provided by the intermediate range neutron 
flux detectors and the nuclear instrumentation system operating in a 
two-out-of-three channel logic. As a result, the ability to mitigate 
any accident previously evaluated is not significantly affected.
    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 reduces the probability of a spurious 
reactor trip during physics testing. The reactor trip system 
continues to be capable of protecting the reactor utilizing the 
intermediate range neutron flux reactor trip and the power range 
neutron flux trips operating in a two-out-of-three trip logic. As a 
result, the reactor is protected and the probability of a spurious 
reactor trip is significantly reduced.
    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

[[Page 5810]]

amendment request involves no significant hazards consideration.

Request No. 16: TSTF-325, Revision 0, ``ECCS Conditions and Required 
Actions with Less Than 100% Equivalent ECCS Flow''

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change corrects the structure of Technical 
Specification 3.5.2 to assure its proper application. There is no 
change in intent or in the way the Technical Specification is 
applied. The literal (and unintended) interpretation of the existing 
LCO structure could, under some circumstances, provide longer than 
intended Completion Times for restoration of operability. The 
proposed change only clarifies the requirements of the Required 
Actions. Since the proposed change affects neither the Technical 
Specification intent, nor its application, the proposed change will 
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 corrects the structure of the Technical 
Specification to assure its correct application. There is no change 
in intent or in the way the Technical Specification is applied. The 
proposed changes would not result in any physical alterations to the 
plant configuration, no new equipment is added, no equipment 
interfaces are modified, and no changes to any equipment's function 
or the method of operating the equipment are being made. As the 
proposed changes would not change the design, configuration or 
operation of the plant, no new or different kinds of accident modes 
are created.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change corrects the structure of the Technical 
Specification to assure its correct application. There is no change 
in intent or in the way the Technical Specification is applied.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

Request No. 17: TSTF-340-A, Revision 3, ``Allow 7 Day Completion Time 
for a Turbine-Driven AFW Pump Inoperable''

    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 3.7.5, ``Auxiliary 
Feedwater (AFW) System,'' to allow a 7 day Completion Time to 
restore an inoperable turbine-driven pump in Mode 3 immediately 
following a refueling outage, if Mode 2 has not been entered. An 
inoperable AFW turbine-driven pump is not an initiator of any 
accident previously evaluated. The ability of the plant to mitigate 
an accident is no different while in the extended Completion Time 
than 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 revises Specification 3.7.5, ``Auxiliary 
Feedwater (AFW) System,'' to allow a 7 day Completion Time to 
restore an inoperable turbine-driven AFW pump in Mode 3 immediately 
following a refueling outage if Mode 2 has not been entered. In Mode 
3 immediately following a refueling outage, core decay heat is low 
and the need for AFW is also diminished. The two operable motor 
driven AFW pumps are available and there are alternate means of 
decay heat removal if needed. As a result, the risk presented by the 
extended Completion Time is minimal.
    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.

Request No. 18: TSTF-343, Revision 1, ``Containment Structural 
Integrity''

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the Technical Specifications (TS) 
Administrative Controls programs for consistency with the 
requirements of 10 CFR 50, paragraph 55a(g)(4) for components 
classified as Code Class CC. The proposed changes affect the 
frequency of visual examinations that will be performed for the 
concrete surfaces of the containment for the purpose of the 
Containment Leakage Rate Testing Program, and allows those 
examinations to be performed during power operation in addition to 
during a refueling outage.
    The frequency of visual examinations of the containment and the 
mode of operation during which those examinations are performed does 
not affect the initiation of any accident previously evaluated. The 
use of NRC approved methods and frequencies for performing the 
inspections will ensure the containment continues to perform the 
mitigating function assumed for accidents previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes revise the TS Administrative Controls 
programs for consistency with the requirements of 10 CFR 50, 
paragraph 55a(g)(4) for components classified as Code Class CC. The 
proposed changes affect the frequency of visual examinations that 
will be performed for the concrete surfaces of the containment for 
the purpose of the Containment Leakage Rate Testing Program, and 
allows those examinations to be performed during power operation in 
addition to during a refueling outage.
    The proposed changes do not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed changes 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 off-site and there is no increase in individual or 
cumulative occupational exposure.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes revise the Technical Specifications (TS) 
Administrative Controls programs for consistency with the 
requirements of 10 CFR 50, paragraph 55a(g)(4) for components 
classified as Code Class CC. The proposed changes affect the 
frequency of visual examinations that will be performed for the 
concrete surfaces of the containment for the purpose of the 
Containment Leakage Rate Testing Program,

[[Page 5811]]

and allows those examinations to be performed during power operation 
in addition to during a refueling outage. The safety function of the 
containment as a fission product barrier 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.

Request No. 19: TSTF-349-A, Revision 1, ``Add Note to LCO 3.9.5 
Allowing Shutdown Cooling Loops Removal from Operation''

    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 an LCO Note to LCO 3.9.5, ``RHR and 
Coolant Circulation--Low Water Level,'' to allow securing the 
operating train of Residual Heat Removal (RHR) for up to 15 minutes 
to support switching operating trains. The allowance is restricted 
to conditions in which core outlet temperature is maintained at 
least 10 degrees F below the saturation temperature, when there are 
no draining operations, and when operations that could reduce the 
reactor coolant system (RCS) boron concentration are prohibited. 
Securing an RHR train to facilitate the changing of the operating 
train is not an initiator to any accident previously evaluated. The 
restrictions on the use of the allowance ensure that an RHR train 
will not be needed during the 15 minute period to mitigate 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 adds an LCO Note to LCO 3.9.5, ``RHR and 
Coolant Circulation--Low Water Level,'' to allow securing the 
operating train of RHR to support switching operating trains. The 
allowance is restricted to conditions in which core outlet 
temperature is maintained at least 10 degrees F below the saturation 
temperature, when there are no draining operations, and when 
operations that could reduce the reactor coolant system (RCS) boron 
concentration are prohibited. With these restrictions, combined with 
the short time frame allowed to swap operating RHR trains and the 
ability to start an operating RHR train if needed, the occurrence of 
an event that would require immediate operation of an RHR train is 
extremely remote.
    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.

Request No. 20: TSTF-355-A, Revision 0, ``Changes to RTS and ESF 
Tables''

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The RTS [Reactor Trip System] and ESFAS [Engineered Safety 
Feature Actuations System] instrument functions are part of the 
accident mitigation response and are not themselves an initiator of 
any accident previously evaluated. Therefore, the probability of an 
accident previously evaluated is not significantly affected by the 
proposed changes. The changes ensure that automatic protective 
actions will be initiated at or before the condition assumed in the 
safety analysis, and are in accordance with the intent of the 
Technical Specifications. The proposed changes will not cause any 
design or analysis acceptance criteria to be exceeded. Since there 
will be no adverse effect on the trip setpoints or the 
instrumentation associated with the trip setpoints, there will be no 
significant increase in the 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 include modifications to the format of the 
nominal trip setpoints that preserve safety analysis assumptions 
related to accident mitigation. The protection system will continue 
to initiate the protective actions as assumed in the safety 
analysis. The proposed changes will continue to ensure that the trip 
setpoints are maintained consistent with the setpoint methodology 
and the plant safety analysis. As the proposed changes do not change 
the design, configuration or operation of the plant, no new or 
different kinds of accident modes are created.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes do not alter any nominal trip setpoints, 
allowable values, or limiting safety system settings, and will 
continue to ensure that the trip setpoints are maintained consistent 
with the setpoint methodology and the plant safety analysis. The 
response of protection systems to accident transients reported in 
the Final Safety Analysis Report is unaffected by this change, and 
accident analysis acceptance criteria are consequently not affected.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

Request No. 21: TSTF-371-A, Revision 1, ``NIS Power Range Channel Daily 
SR TS Change to Address Low Power Decalibration''

    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 3.3.1, ``RTS 
Instrumentation,'' Surveillances 3.3.1.2 and 3.3.1.3 to move 
requirements currently in a Note to the Surveillance itself. The 
change in presentation is editorial and does not affect the 
application of the Surveillances. The proposed change does not 
affect 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, 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 Specification 3.3.1, ``RTS 
Instrumentation,'' Surveillances 3.3.1.2 and 3.3.1.3 to move 
requirements currently in a Note to the Surveillance itself. The 
proposed change represents an editorial preference and does not 
affect the

[[Page 5812]]

performance of the Surveillance or plant operation. The safety 
function tested by the Surveillance is unaffected.
    Therefore, this 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.

Request No. 22: TSTF-439-A, Revision 2, ``Eliminate Second Completion 
Times Limiting Time From Discovery of Failure To Meet an LCO''

    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 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.
    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.

    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.

Request No. 23: ISTS Adoption #1--Revise LCO 3.3.2 ESFAS Interlock P-4 
Required Action Completion Time

    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 Condition to be entered when the 
ESFAS Interlock P-4 is inoperable. Current Technical Specifications 
require restoring the channel to Operable status within 24 hours or 
be in Mode 3 within the next 12 hours and Mode 5 within the 
following 52 hours. The proposed change provides 48 hours to restore 
the inoperable channel, or be in Mode 3 in 54 hours and Mode 4 in 60 
hours. The ESFAS P-4 interlock is not an initiator to any accident 
previously evaluated. The consequences of any accident previously 
evaluated during the proposed Completion Time are no different from 
the consequences during the existing Completion Time. As a result, 
the proposed change does not result in a significant increase in the 
consequences of any accident previously evaluated.
    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 provides an additional 24 hours to restore 
an inoperable ESFAS P-4 Interlock. During the proposed Completion 
Time, manual actions can perform the functions provided by the 
inoperable P-4 interlock. Also, the proposed Completion Time is 
reasonable given the available redundant channel, and the low 
probability of an event occurring during this interval.
    Therefore, this 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.

Request No. 24: Revise LCO 3.5.5 to 8-hour Completion Time and Note 
allowance

    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 the LCO 3.5.5, ``Seal Injection 
Flow,'' Action A, ``Seal injection flow not within limit,'' 
Completion Time from 4 hours to 8 hours and the Note to SR 3.5.5.1 
to allow 8 hours instead of 4 hours to stabilize reactor coolant 
system (RCS) pressure prior to verifying the seal injection throttle 
valves are properly adjusted. The proposed change does not involve 
the addition or removal of any equipment, or any design changes to 
the facility. Seal injection flow is not an initiator of any 
accident previously evaluated. The consequences of any accident 
previously evaluated during the extended Completion Time or Note 
allowance are the same as during the existing Completion Time and 
Note allowance.
    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 provides additional time to verify seal 
injection flow is within limit or to restore seal injection flow to 
within limit if it is discovered that it is not within limit. The 
additional time is acceptable on the basis that there is little 
likelihood of an event that would challenge the ECCS occurring 
during the 8-hour window, and it reduces the pressure on the 
operations staff should iterations in the adjustment procedure be 
necessary to balance seal injection flow.
    Therefore, this 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: Leigh D. Perry, SVP & General Counsel of 
Operations and Nuclear, Southern Nuclear

[[Page 5813]]

Operating Company, 40 Iverness Center Parkway, Birmingham, AL 35201.
    NRC Branch Chief: Robert J. Pascarelli.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri
    Date of amendment request: October 2, 2014. A publicly-available 
version is in ADAMS under Accession No. ML14275A441.
    Description of amendment request: The proposed amendment upgrades 
the Emergency Action Level scheme by adopting NRC-endorsed Nuclear 
Energy Institute 99-01, Revision 6, ``Methodology for the Development 
of Emergency Action Levels for Non-Passive Reactors,'' issued January 
2011 (ADAMS Accession No. ML110240324).
    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 changes to the Callaway Plant emergency action 
levels do not impact the physical function of plant structures, 
systems, or components (SSC) or the manner in which SSCs perform 
their design function. The proposed changes neither adversely affect 
accident initiators or precursors, nor alter design assumptions. The 
proposed changes do not alter or prevent the ability of SSCs to 
perform their intended function to mitigate the consequences of an 
initiating event within assumed acceptance limits. No operating 
procedures or administrative controls that function to prevent or 
mitigate accidents are affected by the proposed changes.
    Therefore, the proposed changes 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 changes do not involve a physical alteration of the 
plant (i.e., no new or different type of equipment will be installed 
or removed) or a change in the method of plant operation. The 
proposed changes will not introduce failure modes that could result 
in a new accident, and the change does not alter assumptions made in 
the safety analysis. The proposed changes to the Callaway Plant 
emergency action levels are not initiators of any accidents.
    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.
    Margin of safety is associated with the ability of the fission 
product barriers (i.e., fuel cladding, reactor coolant system 
pressure boundary, and containment structure) to limit the level of 
radiation dose to the public. The proposed changes do not impact 
operation of the plant or its response to transients or accidents. 
The changes do not affect the Technical Specifications or the 
operating license. The proposed changes do not involve a change in 
the method of plant operation, and no accident analyses will be 
affected by the proposed changes. Additionally, the proposed changes 
will 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 these changes. The proposed changes 
will not result in plant operation in a configuration outside the 
design basis. The proposed changes do not adversely affect systems 
that respond to safely shut down the plant and to maintain the plant 
in a safe shutdown condition. The emergency plan will continue to 
activate an emergency response commensurate with the extent of 
degradation of plant 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: John O'Neill, Esq., Pillsbury Winthrop Shaw 
Pittman LLP, 2300 N Street NW., Washington, DC 20037.
    Acting NRC Branch Chief: Eric R. Oesterle.

III. 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.

Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear 
Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

    Date of amendment request: April 24, 2014.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3/4.4.5, ``Steam Generator Tube Integrity,'' TS 
6.8.4.I, ``Steam Generator Program,'' and TS 6.9.1.7, ``Steam Generator 
Tube Inspection Report'' to address implementation associated with the 
inspections and reporting requirements as described in Technical 
Specifications Task Force (TSTF) TSTF-510, Revision 2, ``Revision to 
Steam Generator Program Inspection Frequencies and Tube Sample 
Selection.''
    Date of issuance: January 9, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 145. A publicly-available version is in ADAMS under 
Accession No. ML14307A800; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-63 The amendment revised the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: July 22, 2014 (79 FR 
42543).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 9, 2015.
    No significant hazards consideration comments received: No.

[[Page 5814]]

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of application for amendment: October 31, 2013, as 
supplemented by letters dated May 29, 2014, and September 9, 2014.
    Brief description of amendment: The amendment revised Technical 
Specification Surveillance Requirements 3.5.1.4 and 3.5.2.5 for low 
pressure core spray and low pressure coolant injection pump flows.
    Date of issuance: January 7, 2015.
    Effective date: As of its date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 229. A publicly-available version is in ADAMS under 
Accession No. ML14335A189; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-21: The amendment 
revised the Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 8, 2014 (79 FR 
19399). The supplemental letters dated May 29, 2014, and September 9, 
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 January 7, 2015.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
2, Pope County, Arkansas

    Date of application for amendment: January 21, 2014, as 
supplemented by letters dated March 17 and September 24, 2014.
    Brief description of amendment: The amendment revised the Technical 
Specification 6.5.16 requirements for the local leak test required for 
the containment building emergency escape air lock doors, in that it 
would require a seal contact verification in lieu of the current seal 
pressure test to verify leak tightness.
    Date of issuance: January 22, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 299. A publicly-available version is in ADAMS under 
Accession No. ML14350B285; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications/license.
    Date of initial notice in Federal Register: April 15, 2014 (79 FR 
21296). The supplemental letter dated September 24, 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 January 22, 2015.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: November 15, 2013, as supplemented by 
letters dated April 16, 2014; September 11, 2014; and November 7, 2014.
    Brief description of amendments: The amendments revise the 
Technical Specification (TS) requirements related to the response time 
for the main steam line flow-high isolation function.
    Date of issuance: January 7, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 214 and 175. A publicly-available version is in 
ADAMS under Accession No. ML14344A681; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. NPF-39 and NPF-85: 
Amendments revised the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: February 4, 2014 (79 FR 
6642). The supplemental letters dated April 16, 2014; September 11, 
2014; and November 7, 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 January 7, 2015.
    No significant hazards consideration comments received: No.

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 requests: July 16, 2013, as supplemented by 
letters dated September 18, 2013, January 22, April 7, August 12, and 
November 11, 2014.
    Brief description of amendments: The amendments revises the 
Technical Specifications to include the use of neutron absorbing spent 
fuel pool rack inserts (i.e., NETCO-SNAP-IN[supreg] rack inserts) for 
the purpose of criticality control in the spent fuel pools.
    Date of issuance: December 31, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment Nos.: 253-Unit 1; 248-Unit 2. A publicly-available 
version is in ADAMS under Accession No. ML14346A306; documents related 
to these amendments are listed in the safety evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. DPR-29 and DPR-30: The 
amendments revised the Technical Specifications and Facility Operating 
License.
    Date of initial notice in Federal Register: July 8, 2014 (79 FR 
38577). The supplemental letters dated September 18, 2013, January 22, 
April 7, August 12, and November 11, 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 December 31, 2014.
    No significant hazards consideration comments received: No.

Northern States Power Company--Minnesota, Docket No. 50-263, Monticello 
Nuclear Generating Plant (MNGP), Wright County, Minnesota

    Date of amendment request: November 14, 2013.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.5.11, ``Primary Containment Leakage Rate Testing 
Program,'' by removing TS 5.5.11.d.2.b, the reduced pressure testing 
option for drywell airlock door leakage testing. This testing 
methodology is not required and does not reflect the current testing 
practice at MNGP. As such, the drywell

[[Page 5815]]

airlock door seals will be tested by performing an overall airlock 
leakage test as specified in current TS 5.5.11.d.2.a.
    Date of issuance: January 8, 2015.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 187. A publicly-available version is in ADAMS under 
Accession No. ML14323A033; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-22: This amendment 
revises the Renewed Facility Operating License and the Technical 
Specifications.
    Date of initial notice in Federal Register: August 5, 2014 (79 FR 
45478).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 8, 2015.
    No significant hazards consideration comments received: No.

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: April 3, 2014, as supplemented by letter 
dated May 19, 2014.
    Brief description of amendment: The amendment revises Tier 2* 
information, incorporated into the VCSNS Units 2 and 3 Updated Final 
Safety Analysis Report (UFSAR). Specifically, the amendment revises the 
details regarding the structural floor of the Auxiliary Building and 
its constructability. Notes are added to drawings in Subsection 3H.5 of 
the UFSAR in order to clarify variations in detail design such as size 
and spacing or reinforcement and spans of the noncritical sections of 
floors.
    Date of issuance: July 18, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 14. A publicly-available version is in ADAMS under 
Accession No. ML14188B185; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses No. NPF-93 and NPF-94: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: April 29, 2014 (79 FR 
24024).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 18, 2014.
    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: March 17, 2014, and revised by letters 
dated May 8, September 2, and October 2, 2014.
    Brief description of amendment: The amendment revises the VEGP 
Units 3 and 4 Updated Final Safety Analysis Report (UFSAR) by 
clarifying how human diversity was applied during the design process 
for the Component Interface Module and Diverse Actuation System. The 
changes to the VEGP Units 3 and 4 UFSAR include changes to Table 1.6, 
``Material Referenced,'' Chapter 7, Sections 7.1.2.14.1, 7.1.7 and 
7.2.4 and the addition of Appendix 7A to Chapter 7. The changes to the 
VEGP Units 3 and 4 UFSAR modify information related to human diversity, 
as presented in a Tier 2* document, WCAP-17179-P and WCAP-17179-NP, 
``AP1000 Component Interface Module Technical Report,'' Revision 2, and 
two Tier 2 documents, WCAP-15775, ``AP1000 Instrumentation and Control 
Defense-in-Depth and Diversity Report,'' Revision 4 and WCAP-17184-P, 
``AP1000 Diverse Actuation System Planning and Functional Design 
Summary Technical Report,'' that are incorporated by reference in the 
VEGP Units 3 and 4 UFSAR.
    Date of issuance: December 24, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 28. A publicly-available version is in ADAMS under 
Accession No. ML14329A298; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: April 29, 2014 (79 FR 
24021).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 24, 2014.
    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: August 22, 2014, and revised by letter 
dated September 23, 2014, and supplemented by letters dated October 30 
and November 6, 2014.
    Brief description of amendment: The amendment revises the VEGP 
Units 3 and 4 Updated Final Safety Analysis Report to reflect changes 
related to:
    (a) Installation of an additional non-safety-related battery;
    (b) Revision to the annex building internal configuration by 
converting a shift turnover room to a battery room, adding an 
additional battery equipment room, and moving a fire area wall;
    (c) Increase in the height of a room in the annex building; and
    (d) Increase in thicknesses of certain annex building floor slabs.
    In addition, the proposed changes also include reconfiguring 
existing rooms and related rooms, wall, and access path changes and 
making changes to the corresponding Tier 1 information in Appendix C to 
the Combined Licenses.
    Date of issuance: December 23, 2014.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 27. A publicly-available version is in ADAMS under 
Accession No. ML14323A609; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses No. NPF-91 and NPF-92: Amendment revised 
the Facility Combined Licenses.
    Date of initial notice in Federal Register: October 14, 2014 (79 FR 
61662).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 23, 2014.
    No significant hazards consideration comments received: No.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: January 23, 2014.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.4.12, ``Cold Overpressure Mitigation System 
(COMS),'' to reflect the mass input transient analysis that assumes an 
Emergency Core Cooling System centrifugal charging pump and the normal 
charging pump capable of injecting into the reactor coolant system when 
TS 3.4.12 is applicable. The amendment also revised TS Table 3.3.1-1, 
``Reactor Trip System Instrumentation,'' to remove unnecessary page 
number references.
    Date of issuance: January 20, 2015.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 210. A publicly-available version is in ADAMS under

[[Page 5816]]

Accession No. ML14350B239; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-30: The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 1, 2014 (79 FR 
18348).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 20, 2015.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 26th day of January 2015.

    For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2015-01917 Filed 2-2-15; 8:45 am]
BILLING CODE 7590-01-P