ML15138A275

2015/04/28 NRR E-mail Capture - NRC Notification of State of Florida Regarding St. Lucie License Amendment - Technical Specifications Surveillance Frequencies

ML15138A275
Person / Time
Site:	Saint Lucie
Issue date:	04/28/2015
From:	Audrey Klett Plant Licensing Branch II
To:	Becker C State of FL, Dept of Health, Bureau of Radiation Control
References
Download: ML15138A275 (23)

Text

1 NRR-PMDAPEm Resource From: Klett, Audrey Sent: Tuesday, April 28, 2015 2:37 PM To: Becker, Cindy L (Cindy.Becker@flhealth.gov)

Cc: Saba, Farideh

Subject:

NRC Notification of State of Florida Regardi ng St. Lucie License Amendment - Technical Specifications Surveillance Frequencies Attachments:

2015-04298 second notice.pdfGood Afternoon Ms. Becker, The U.S. Nuclear Regulatory Commission (NRC) is about to issue license amendments for the St. Lucie Nuclear Plant, Units 1 and 2. In accordance with Title 10 of the Code of Federal Regulations, Section 50.91(b), I am notifying you of the proposed issuance of these amendments.

Please reply if the State of Florida has comments on the following licensing action submitted to the NRC by Florida Power & Light Company, et al. (the licensee): "Application for Technical Specification Change Regarding Risk-Informed Justifications for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program," dated February 20, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14070A087), as supplemented (ADAMS Accession Nos. ML14349A333, ML15029A497, and ML15042A122). There is an additional supplement dated April 18, 2015, which is not yet available in ADAMS. I will email you the ADAMS Accession number when it becomes available.

Attached is the associated proposed no significant hazards consideration for this amendment request that was published in the Federal Register on March 3, 2015 (80 FR 11477, see attached).

Your response is requested by May 8, 2015. If you have any questions, please call me at (301) 415-0489.

Thank you, Audrey Klett Project Manager U.S. Nuclear Regulatory Commission

301-415-0489

Hearing Identifier: NRR_PMDA Email Number: 2080 Mail Envelope Properties (Audrey.Klett@nrc.gov20150428143600)

Subject:

NRC Notification of State of Florida Regarding St. Lucie License Amendment - Technical Specifications Surveillance Frequencies Sent Date: 4/28/2015 2:36:36 PM Received Date: 4/28/2015 2:36:00 PM From: Klett, Audrey Created By: Audrey.Klett@nrc.gov Recipients: "Saba, Farideh" <Farideh.Saba@nrc.gov> Tracking Status: None "Becker, Cindy L (Cindy.Becker@flhealth.gov)" <Cindy.Becker@flhealth.gov> Tracking Status: None Post Office: Files Size Date & Time MESSAGE 1459 4/28/2015 2:36:00 PM 2015-04298 second notice.pdf 324180 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date: Recipients Received:

11472 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices www.regulations.gov as well as entering the comment submissions into ADAMS.

The NRC does not routinely edit

comment submissions to remove

identifying or contact information.

If you are requesting or aggregating comments from other persons for

submission to the NRC, then you should

inform those persons not to include

identifying or contact information that

they do not want to be publicly

disclosed in their comment submission.

Your request should state that the NRC

does not routinely edit comment

submissions to remove such information

before making the comment

submissions available to the public or

entering the comment submissions into

ADAMS. II. Background In accordance with the Paperwork Reduction Act of 1995 (44 U.S.C.

Chapter 35), the NRC is requesting

public comment on its intention to

request the OMBs approval for the

information collection summarized

below. 1. The title of the information collection:

NRC Request for Information Concerning Patient Release Practices.

2. OMB approval number:

OMB control number has not yet been

assigned to this proposed information

collection.

3. Type of submission:

New. 4. The form number, if applicable: N/

A. 5. How often the collection is required or requested:

Once. 6. Who will be required or asked to respond: Medical professional organizations, physicians, patients, patient advocacy groups, NRC and

Agreement State medical use licensees, Agreement States, and other interested

individuals who use, receive, license or

have interest in the use of I-131 sodium

iodine (hereafter referred to as I-131)

for the treatment of thyroid conditions.

7. The estimated number of annual responses: A one-time collection

estimated to have 1,180 responses (620

medical community + 560 patients).

8. The estimated number of annual respondents: 1,180 respondents (620

medical community

+ 560 patients). 9. The estimated number of hours needed annually to comply with the

information collection requirement or

request: 457.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (255 medical community + 202.5 patients).

10. Abstract:

The NRC is requesting a one-time information collection that

will be solicited in a Federal Register notice (FRN). The FRN will have

specific I-131 patient release questions

associated with: (1) Existing Web sites

that the responders believe provide access to clear and consistent patient information about I-131 treatment

processes and procedures; (2)

information the responders believe

represent best practices used in making

informed decisions on releasing I-131

patients and stand alone or

supplemental voluntary patient/licensee guidance acknowledgment forms, if

available; (3) an existing set of

guidelines that the responder developed

or received that provides instructions to

released patients; and (4) an existing

guidance brochure that the responder

believes would be acceptable for

nationwide distribution. The responses

will form the basis for patient release

guidance products developed in

response to the NRCs April 28, 2014, Staff RequirementsCOMAMM

0001/COMWDM-14-0001

Background and Proposed Direction to

NRC Staff to Verify Assumptions Made

Concerning Patient Release Guidance.

The Commission, based on information

from patients and patient advocacy

groups, questioned the availability of

clear, consistent, patient friendly and

timely patient release information and

directed the staff to work with a wide

variety of stakeholders when developing

new guidance products. This

information collection effort was

developed to gain input from as many

stakeholders as possible. The NRC

solicitation in the Federal Register is to obtain existing information from a

variety of stakeholders.

III. Specific Requests for Comments The NRC is seeking comments that address the following questions:

1. Is the proposed collection of information necessary for the NRC to

properly perform its functions? Does the

information have practical utility?

2. Is the estimate of the burden of the information collection accurate?

3. Is there a way to enhance the quality, utility, and clarity of the

information to be collected?

4. How can the burden of the information collection on respondents

be minimized, including the use of

automated collection techniques or

other forms of information technology?

Dated at Rockville, Maryland, this 25th day of February, 2015.

For the Nuclear Regulatory Commission.

Tremaine Donnell, NRC Clearance Officer, Office of Information Services.

[FR Doc. 2015-04318 Filed 3-2-15; 8:45 am]

BILLING CODE 7590-01-P NUCLEAR REGULATORY COMMISSION

[NRC-2015-0041]

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses and Combined Licenses

Involving No Significant Hazards

Considerations AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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 February 5, 2015 to February 18, 2015. The last

biweekly notice was published on

February 17, 2015.

DATES: Comments must be filed by April 2, 2015. A request for a hearing must be

filed by May 4, 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 http://www.regulations.gov and search for Docket ID NRC-2015-0041. Address

questions about NRC dockets to Carol

Gallagher; telephone: 301-415-3463;

email: Carol.Gallagher@nrc.gov.

• Mail comments to:

Cindy Bladey, Office of Administration, Mail Stop:

OWFN-12-H08, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001.

For additional direction on obtaining information and submitting comments, see Obtaining Information and

Submitting Comments in the SUPPLEMENTARYINFORMATION section of this document. FORFURTHERINFORMATIONCONTACT

Shirley Rohrer, Office of Nuclear

Reactor Regulation, U.S. Nuclear

Regulatory Commission, Washington DC

20555-0001; telephone: 301-415-5411, email: Shirley.Rohrer@nrc.gov. VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00086Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11473 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices SUPPLEMENTARYINFORMATION

I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2015-0041 when contacting the NRC about

the availability of information for this

action. You may obtain publicly-

available information related to this

action by any of the following methods:

• Federal rulemaking Web site:

Go to http://www.regulations.gov and search for Docket ID NRC-2015-0041.

• NRCs Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the

ADAMS Public Documents collection at

http://www.nrc.gov/reading-rm/

adams.html.

To begin the search, select ADAMS Public Documents and then select Begin Web-based ADAMS Search.

For problems with ADAMS, please contact the NRCs 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. *NRCs PDR:

You may examine and purchase copies of public documents at

the NRCs PDR, Room O1-F21, One

White Flint North, 11555 Rockville

Pike, Rockville, Maryland 20852.

B. Submitting Comments Please include Docket ID NRC-2015-0041, facility name, unit number(s),

application date, and subject in your

comment submission.

The NRC cautions you not to include identifying or contact information that

you do not want to be publicly

disclosed in your comment submission.

The NRC will post all comment

submissions at http://www.regulations.gov as well as enter 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 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 Commissions regulations in

§50.92 of Title 10 of the Code of Federal Regulations (10 CFR), this means that operation of the facility in accordance

with the proposed amendment would

not (1) involve a significant increase in

the probability or consequences of an

accident previously evaluated, or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of 60 days after the date of

publication of this notice. The

Commission may issue the license

amendment before expiration of the 60-

day period provided that its final

determination is that the amendment

involves no significant hazards

consideration. In addition, the

Commission may issue the amendment

prior to the expiration of the 30-day

comment period should circumstances

change during the 30-day comment

period such that failure to act in a

timely way would result, for example in

derating or shutdown of the facility.

Should the Commission take action

prior to the expiration of either the

comment period or the notice period, it

will publish in the Federal Register a notice of issuance. Should the

Commission make a final No Significant

Hazards Consideration Determination, any hearing will take place after

issuance. The Commission expects that

the need to take this action will occur

very infrequently.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene Within 60 days after the date of publication of this notice, any person(s)

whose interest may be affected by this

action may file a request for a hearing

and a petition to intervene with respect

to issuance of the amendment to the subject facility operating license or combined license. Requests for a

hearing and a petition for leave to

intervene shall be filed in accordance

with the Commissions 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 NRCs PDR, located at One White Flint North, Room

O1-F21, 11555 Rockville Pike (first

floor), Rockville, Maryland 20852. The

NRCs regulations are accessible

electronically from the NRC Library on

the NRCs Web site at http://www.nrc.gov/reading-rm/doc-

collections/cfr/. If a request for a hearing or petition for leave to intervene is filed

by the above date, the Commission or a

presiding officer designated by the

Commission or by the Chief

Administrative Judge of the Atomic

Safety and Licensing Board Panel, will

rule on the request and/or petition; and

the Secretary or the Chief

Administrative Judge of the Atomic

Safety and Licensing Board will issue a

notice of a hearing or an appropriate

order. As required by 10 CFR 2.309, a petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following general requirements: (1) The

name, address, and telephone number of

the requestor or petitioner; (2) the

nature of the requestors/petitioners

right under the Act to be made a party

to the proceeding; (3) the nature and

extent of the requestors/petitioners

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

requestors/petitioners 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 VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00087Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11474 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices 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 NRCs E-Filing rule

(72 FR 49139; August 28, 2007). The E-

Filing process requires participants to

submit and serve all adjudicatory

documents over the internet, or in some

cases to mail copies on electronic

storage media. Participants may not

submit paper copies of their filings

unless they seek an exemption in

accordance with the procedures

described below.

To comply with the procedural requirements of E-Filing, at least ten 10 days prior to the filing deadline, the participant should contact the Office of

the Secretary by email at

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

NRCs public Web site at http://www.nrc.gov/site-help/e-submittals/

getting-started.html.

System requirements for accessing the E-

Submittal server are detailed in the

NRCs Guidance for Electronic

Submission, which is available on the

agencys public Web site at http://www.nrc.gov/site-help/e-

submittals.html.

Participants may attempt to use other software not listed

on the Web site, but should note that the

NRCs 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 NRCs 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 NRCs Web

site. Further information on the Web-

based submission form, including the

installation of the Web browser plug-in, is available on the NRCs public Web

site at http://www.nrc.gov/site-help/e-submittals.html.

Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in Portable Document Format (PDF) in accordance with NRC guidance available on the NRCs public Web site at http://www.nrc.gov/site-help/e-submittals.html.

A filing is considered complete at the time the documents are submitted through the NRCs 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 NRCs 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 NRCs adjudicatory E-Filing system

may seek assistance by contacting the

NRC Meta System Help Desk through

the Contact Us link located on the

NRCs public Web site at http://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 VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00088Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11475 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices 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 NRCs

electronic hearing docket which is

available to the public at http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are

requested not to include personal

privacy information, such as social

security numbers, home addresses, or

home phone numbers in their filings, unless an NRC regulation or other law

requires submission of such

information. However, a request to

intervene will require including

information on local residence in order

to demonstrate a proximity assertion of

interest in the proceeding. With respect to copyrighted works, except for limited

excerpts that serve the purpose of the

adjudicatory filings and would

constitute a Fair Use application, participants are requested not to include

copyrighted materials in their

submission.

Petitions for leave to intervene must be filed no later than 60 days from the

date of publication of this notice.

Requests for hearing, petitions for leave

to intervene, and motions for leave to

file new or amended contentions that

are filed after the 60-day deadline will

not be entertained absent a

determination by the presiding officer

that the filing demonstrates good cause

by satisfying the three factors in 10 CFR

2.309(c)(1)(i)-(iii).

For further details with respect to these license amendment applications, see the application for amendment

which is available for public inspection

in ADAMS and at the NRCs PDR. For

additional direction on accessing

information related to this document, see the Obtaining Information and

Submitting Comments section of this

document.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2 (ANO-2), Pope County, Arkansas Date of amendment request:

February 6, 2015. A publicly-available version is

in ADAMS under Accession No.

ML15041A068.

Description of amendment request:

The amendment would revise a Note to

Technical Specification (TS)

Surveillance Requirement (SR) 4.1.3.1.2

to exclude Control Element Assembly (CEA) 18 from being exercised per the

SR for the remainder of Cycle 24 due to

a degrading upper gripper coil.

Basis for proposed no significant hazards consideration determination:

As required by 10 CFR 50.91(a), the licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below: 1. Does the proposed change involve a significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

One function of the CEAs is to provide a means of rapid negative reactivity addition

into the core. This occurs upon receipt of a

signal from the Reactor Protection System.

This function will continue to be

accomplished with the approval of the

proposed change. Typically, once per 92 days

each CEA is moved at least five inches to

ensure the CEA is free to move. CEA 18

remains trippable (free to move) as illustrated

by the last performance of SR 4.1.3.1.2 in

January 2015. However, due to abnormally

high coil voltage and current measured on

the CEA 18 Upper Gripper Coil (UGC), future

exercising of the CEA could result in the CEA inadvertently inserting into the core, if the

UGC were to fail during the exercise test. The

mis-operation of a CEA, which includes a

CEA drop event, is an abnormal occurrence

and has been previously evaluated as part of

the ANO-2 accident analysis. Inadvertent

CEA insertion will result in a reactivity

transient and power reduction, and could

lead to a reactor shutdown if the CEA is

deemed to be unrecoverable. The proposed

change would minimize the potential for

inadvertent insertion of CEA 18 into the core

by maintaining the CEA in place using the

Lower Gripper Coil (LGC), which is operating

normally. The proposed change will not

affect the CEAs ability to insert fully into the

core upon receipt of a reactor trip signal.

No modifications are proposed to the Reactor Protection System or associated

Control Element Drive Mechanism Control

System logic with regard to the ability of CEA

18 to remain available for immediate

insertion. The accident mitigation features of

the plant are not affected by the proposed

amendment. Because CEA 18 remains

trippable, no additional reactivity

considerations need to be taken into

consideration. Nevertheless, Entergy has

evaluated the reactivity consequences

associated with failure of CEA 18 to insert

upon a reactor trip in accordance with TS

requirements for Shutdown Margin (SDM)

and has determined that SDM requirements

would be met should such an event occur at

any time during the remainder of Cycle 24

operation.

Therefore, this change does not involve a significant increase in the probability or

consequences of an accident previously

evaluated.

2. Does the proposed change create the possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

CEA 18 remains trippable. The proposed change will not introduce any new design

changes or systems that can prevent the CEA

from [performing] its specified safety

function. As discussed previously, CEA mis-operation has been previously evaluated in the ANO-2 accident analysis. Furthermore, SDM has been shown to remain within limits

should an event occur at any time during the

remainder of operating Cycle 24 such that

CEA 18 fails to insert into the core upon

receipt of a reactor trip signal.

Therefore, this change does not create the possibility of a new or different kind of

accident from an accident previously

evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

SR 4.1.3.1.2 is intended to verify CEAs are free to move (i.e., not mechanically bound).

The physical and electrical design of the

CEAs, and past operating experience, provides high confidence that CEAs remain

trippable whether or not exercised during

each SR interval. Eliminating further

exercising of CEA 18 for the remainder of

Cycle 24 operation does not directly relate to

the potential for CEA binding to occur. No

mechanical binding has been previously

experienced at ANO-2. CEA 18 is contained

within a Shutdown CEA Group and is not

used for reactivity control during power

maneuvers (the CEA must remain fully

withdrawn at all times when the reactor is

critical). In addition, Entergy has concluded

that required SDM will be maintained should

CEA 18 fail to insert following a reactor trip

at any point during the remainder of Cycle

24 operation.

Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensees analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee:

Joseph A.

Aluise, Associate General Counsel

Nuclear, Entergy Services, Inc., 639

Loyola Avenue, New Orleans, Louisiana

70113. NRC Acting Branch Chief:

Eric R. Oesterle.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of amendment request:

October 1, 2014, as supplemented by letter dated

February 2, 2015. Publicly-available

versions are in ADAMS under

Accession Nos. ML14275A374 and

ML15033A482.

Description of amendment request:

The amendment would relocate

Technical Specifications 3.9.6, Refuel

Machine, and 3.9.7, Crane Travel, to

the Technical Requirements Manual.

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 VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00089Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11476 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability

or consequences of an accident

previously evaluated?

Response: No.

This proposed change relocates Technical Specifications (TS) 3.9.6 (Refuel Machine) and TS 3.9.7 (Crane

Travel) to the Waterford 3 Technical

Requirements Manual (TRM). This is consistent with the requirements of [10

CFR 50.36(c)(2)(ii)] and aligns with

NUREG-1432 (Combustion Engineering

Standard Technical Specifications).

The applicable TS 3.9.6 and TS 3.9.7 design basis accident is the Fuel

Handling Accident (FHA) described in

the Updated Final Safety Analysis

Report (UFSAR) Section 15.7.3.4. The

limiting FHA results in all the fuel pins

in the dropped and impacted fuel

assemblies failing (472 pins or 236 per

assembly). The analysis assumes that a

fuel assembly is dropped as an initial

condition and no equipment or

intervention can prevent the initiating

condition. The proposed change was

evaluated against [10 CFR

50.36(c)(2)(ii)] criteria and shows no

impact to the lowest functional

capability or performance levels of

equipment required for safe operation of

the facility because the TS 3.9.6 and TS 3.9.7 requirements do not prevent the

accident conditions from occurring and

do not limit the severity of the accident.

Since, the dropped fuel assembly and

the impacted fuel assembly are both

already failed in the design basis

accident scenario, this change could not

result in a significant increase in the

accident consequences. The TS 3.9.6

and TS 3.9.7 equipment are not required

to respond, mitigate, or terminate any

design basis accident, thus this change

will not adversely impact the likelihood

or probability of a design basis accident.

The TS 3.9.6 and TS 3.9.7 requirements do not prevent the

accident conditions from occurring and

do not limit the severity of the accident.

Therefore the TS 3.9.6 and TS 3.9.7 relocation to the TRM would not cause

a significant increase in the accident

probability or accident consequences.

2. Does the proposed change create the possibility of a new or different kind

of accident from any accident

previously evaluated?

Response: No.

This proposed change relocates TS 3.9.6 (Refuel Machine) and TS 3.9.7 (Crane Travel) to the Waterford 3 TRM.

In general, Technical Specifications are

based upon the accident analyses. The

accident analyses assumptions and

initial conditions must be protected by the Technical Specifications. This is a requirement as outlined in [10 CFR

50.36]. [10 CFR 50.36(b)] states the technical specifications will be derived from the

analyses and evaluation included in the

safety analysis report.

[10 CFR 50.36(c)(2)(i)] states that

[]the limiting conditions for operation

are the lowest functional capability or

performance levels of equipment

required for safe operation of the facility[... .] [10 CFR 50.36(c)(2)(ii)]

provides the four criteria in which any

one met requires a limiting condition for operation. The proposed change demonstrated that the [10 CFR

50.36(c)(2)(ii)] criteria were not met and

the relocation to the TRM is allowable.

By not meeting the [10 CFR

50.36(c)(2)(ii)] criteria for inclusion into

the TS means that TS 3.9.6 and TS 3.9.7

do not impact the accident analyses

previously evaluated and would not

create the possibility of a new or

different kind of accident.

Specifically, TS 3.9.6 and TS 3.9.7 equipment are not instrumentation used

to detect, and indicate in the control

room, a significant abnormal

degradation of the reactor coolant

pressure boundary (Criterion 1). TS 3.9.6 and TS 3.9.7 do not contain a

process variable, design feature, or

operating restriction that is an initial

condition of a Design Basis Accident or

Transient analysis that either assumes

the failure of or presents a challenge to

the integrity of a fission product barrier (Criterion 2). TS 3.9.6 and TS 3.9.7 does

not contain a structure, system, or

component that is part of the primary success path and which functions or actuates to mitigate a Design Basis

Accident or Transient that either

assumes the failure of or presents a

challenge to the integrity of a fission

product barrier (Criterion 3). Lastly, TS 3.9.6 and TS 3.9.7 do not contain a

structure, system, or component which

operating experience or probabilistic

safety assessment has shown to be

significant to public health and safety (Criterion 4).

TS 3.9.6 and 3.9.7 are not required to meet the lowest functional capability or

performance levels of equipment

required for safe operation of the

facility.

Therefore, the accident analyses are not impacted and the possibility of a

new or different kind of accident from

any accident previously evaluated has

not changed.

3. Does the proposed change involve a significant reduction in a margin of

safety? Response: No.

The proposed TS 3.9.6 (Refuel Machine) and TS 3.9.7 (Crane Travel)

relocation to the Waterford 3 TRM is

administrative in nature because all

requirements will be relocated. Any

changes after being relocated to the Waterford 3 TRM will require that the

[10 CFR 50.59] process be entered

ensuring the public health and safety is

maintained. By using the [10 CFR 50.59]

process for future changes, the

regulatory requirements ensure that no

significant reduction in the margin of

safety occurs.

In addition, the TS 3.9.6 and TS 3.9.7 requirements do not prevent the design

basis accident conditions from

occurring and do not limit the severity

of the accident. Thus, TS 3.9.6 and TS 3.9.7 relocation will not adversely

impact the accident analyses and will

not cause a significant reduction in the

margin of safety.

The NRC staff has reviewed the licensees analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Attorney for licensee:

Joseph A.

Aluise, Associate General Council

Nuclear, Entergy Services, Inc., 639

Loyola Avenue, New Orleans, Louisiana

70113. NRC Branch Chief:

Meena K. Khanna.

Exelon Generation Company, LLC (EGC), Docket No. 50-410, Nine Mile

Point Nuclear Station, Unit 2 (NMP2),

Oswego County, New York Date of amendment request:

November 17, 2014. A publicly

available version is in ADAMS under

Accession No. ML14321A744.

Description of amendment request:

The proposed amendment would revise

the NMP2 Technical Specification (TS)

Allowable Value for the Main Steam

Line Tunnel Lead Enclosure

Temperature-High instrumentation from

an ambient temperature dependent (variable setpoint) to ambient

temperature independent (constant

Allowable Value). The changes would

delete Surveillance Requirement (SR)

3.3.6.1.2 and revise the Allowable Value

for Function 1.g on Table 3.3.6.1-1, Primary Containment Isolation

Instrumentation.

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: VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00090Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11477 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices

1. Does the proposed amendment involve a significant increase in the

probability or consequences of an

accident previously evaluated?

Response: No.

The proposed changes do not involve a significant increase in the probability

or consequences of an accident

previously evaluated because the

performance of any equipment credited

in the radiological consequences of an

accident is not affected by the change in

the leak detection capability.

The Main Steam Line Tunnel Lead Enclosure TemperatureHigh is

provided to detect a steam leak in the

lead enclosure and provides diversity to

the high flow instrumentation. This

function provides a mitigating action for

a steam leak in the Main Steam Line

Tunnel Lead Enclosure, which could

lead to a pipe break. This function does

not affect any accident precursors, and

the proposed changes do not affect the

leak detection capability. Additionally, the proposed changes do not degrade

the performance of or increase the

challenges to any safety systems

assumed to function in the accident

analysis.

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 create the possibility of a new or different kind

of accident from any accident

previously evaluated because the

proposed changes do not add or remove

equipment and do not physically alter

the isolation instrumentation. In

addition, the Main Steam Line Tunnel

Lead Enclosure LDS [Leak Detection

System] is not utilized in a different

manner. The proposed changes do not

introduce any new accident initiators

and new failure modes, nor do they

reduce or adversely affect the

capabilities of any plant structure, system, or component to perform their

safety function. The Main Steam Line

Tunnel Lead Enclosure LDS will

continue to be operated in the same

manner. 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 do not involve a significant reduction in a margin of safety because the changes eliminate the temperature setpoint dependency on

lead enclosure temperature while

maintaining the existing upper AV

[Allowable Value] = 175.6

°F, that was previously evaluated and approved.

There is no adverse impact on the

existing equipment capability as well as

associated structures. The increase in

the steam leak rate and associated crack

size continues to be well below the leak

rate associated with critical crack size

that leads to pipe break. The proposed

changes continue to provide the same

level of protection against a main steam line break as the existing setpoint values. Therefore, the proposed changes do not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the licensees 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. Bradley Fewell, Senior Vice President, Regulatory Affairs, Nuclear, and General

Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Branch Chief:

Benjamin G.

Beasley. Florida Power and Light Company, et al. (FPL), Docket Nos. 50-335 and 50-389, St. Lucie Plant, Unit Nos. 1 and 2, St.

Lucie County, Florida Date of amendment request:

February 20, 2014, as supplemented by letters

dated December 11, 2014, January 13

and January 28, 2015. Publicly-available

in ADAMS under Accession Nos.

ML14070A087, ML14349A333, ML15029A497 and ML15042A122.

Description of amendment request:

The NRC staff has previously made a

proposed determination that the

amendment request dated February 20, 2014, involves no significant hazards

consideration (see 79 FR 42550, July 22, 2014). Subsequently, by letter dated

January 28, 2015, the licensee provided

additional information that expanded

the scope of the amendment request as

originally noticed. Accordingly, this

notice supersedes the previous notice in

its entirety.

The amendment would revise the Technical Specifications (TSs) by

relocating specific surveillance

frequency requirements to a licensee-

controlled program with

implementation of Nuclear Energy

Institute (NEI) 04-10 (Revision 1), Risk-Informed Technical Specifications Initiative 5b, Risk-

Informed Method for Control of

Surveillance Frequencies (ADAMS

Accession No. ML071360456). The

licensee stated that the NEI 04-10

methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed

changes to surveillance frequencies, consistent with Regulatory Guide 1.177, An Approach for Plant-Specific, Risk-

Informed Decisionmaking: Technical

Specifications (ADAMS Accession No.

ML003740176). The licensee stated that

the changes are consistent with NRC-

approved Technical Specification Task

Force (TSTF) Improved Standard

Technical Specifications Change

Traveler TSTF-425, Relocate

Surveillance Frequencies to Licensee

ControlRITSTF [Risk-Informed

Technical Specifications Task Force]

Initiative 5b, Revision 3 (ADAMS

Accession No. ML090850642). The

Federal Register notice published on July 6, 2009 (74 FR 31996), announced

the availability of TSTF-425, Revision

3. In the supplement dated January 28, 2015, the licensee requested (1)

additional surveillance frequencies be

relocated to the licensee-controlled

program, (2) editorial changes, (3)

administrative deviations from TSTF-

425, and (4) other changes resulting

from differences between the St. Lucie

Plant TSs and the TSs on which TSTF-

425 was based.

Basis for proposed no significant hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee provided its analysis of the

issue of no significant hazards

consideration, which is presented

below: 1. Does the proposed change involve a significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

The proposed change relocates the specified frequencies for periodic

surveillance requirements to licensee control

under a new Surveillance Frequency Control

Program. Surveillance frequencies are not an

initiator to any accident previously

evaluated. As a result, the probability of any

accident previously evaluated is not

significantly increased. The systems and

components required by the technical

specifications for which the surveillance

frequencies are relocated are still required to

be operable, meet the acceptance criteria for

the surveillance requirements, and be

capable of performing any mitigation

function assumed in the accident analysis.

As a result, the consequences of any accident

previously evaluated are not significantly

increased.

Therefore, the proposed change does not involve a significant increase in the VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00091Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11478 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices 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 previously evaluated?

Response: No.

No new or different accidents result from utilizing the proposed change. The changes

do not involve a physical alteration of the

plant (i.e., no new or different type of equipment will be installed) or a change in

the methods governing normal plant

operation. In addition, the changes do not

impose any new or different requirements.

The changes do not alter assumptions made

in the safety analysis assumptions and

current plant operating practice.

Therefore, the proposed changes do not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the proposed change involve a significant reduction in the margin of safety?

Response: No.

The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified

in applicable codes and standards (or

alternatives approved for use by the NRC)

will continue to be met as described in the

plant licensing basis (including the final

safety analysis report and bases to TS), since

these are not affected by changes to the

surveillance frequencies. Similarly, there is

no impact to safety analysis acceptance

criteria as described in the plant licensing

basis. To evaluate a change in the relocated

surveillance frequency, FPL will perform a

probabilistic risk evaluation using the

guidance contained in NRC-approved NEI

04-10, Revision 1 in accordance with the TS

Surveillance Frequency Control Program. NEI

04-10, Revision 1, methodology provides

reasonable acceptance guidelines and

methods for evaluating the risk increase of

proposed changes to surveillance frequencies

consistent with Regulatory Guide (RG) 1.177.

Therefore, the proposed changes do not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the licensees 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 S.

Blair, Managing AttorneyNuclear, Florida Power & Light Company, 700

Universe Boulevard, MS LAW/JB, Juno

Beach, FL 33408-0420.

NRC Branch Chief:

Shana R. Helton.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey

Point Nuclear Generating, Unit Nos. 3

and 4, Miami-Dade County, Florida Date of amendment request:

November 13, 2014. A publicly-

available version is in ADAMS under

Accession No. ML14337A013.

Description of amendment request:

The amendment would revise Technical

Specification (TS) 3/4.5.2, ECCS

[Emergency Core Cooling System]

SubsystemsT avg[average temperature]

Greater Than or Equal to 350

°F [degrees Fahrenheit], to correct non-

conservative TS requirements. The

licensee also requested editorial changes

to the TS.

Basis for proposed no significant hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee provided its analysis of the

issue of no significant hazards

consideration, which is presented as

follows: 1. Does the proposed change involve a significant increase in the probability

or consequences of an accident

previously evaluated?

No. The proposed TS changes involve TS 3.5.2 Action a, new TS 3.5.2 Action

h, and the provision in SR

[Surveillance Requirement] 4.5.2.a to

address non-conservative TS

requirements. Editorial changes are also

proposed for consistency and clarity.

These changes do not affect any

precursors to any accident previously

evaluated and subsequently, will not

impact the probability or consequences

of an accident previously evaluated.

Furthermore, these changes do not

adversely affect mitigation equipment or

strategies.

Therefore, the proposed changes do not involve a significant increase in the

probability or consequences of an

accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind

of accident from any previously

evaluated?

No. The proposed TS changes involve TS 3.5.2 Action a, new TS 3.5.2 Action

h, and the provision in SR 4.5.2.a to

address non-conservative TS

requirements. Editorial changes are also

proposed for consistency and clarity.

The proposed changes provide better

assurance that the ECCS systems, subsystems, and components are

properly aligned to support safe reactor

operation consistent with the licensing

basis requirements. The proposed

changes do not introduce new modes of

plant operation and do not involve

physical modifications to the plant (no

new or different type of equipment will

be installed). There are no changes in

the method by which any safety related

plant structure, system, or component (SSC) performs its specified safety

function. As such, the plant conditions

for which the design basis accident

analyses were performed remain valid.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of the proposed

change. There will be no adverse effect

or challenges imposed on any SSC as a

result of the proposed change.

Therefore, the proposed changes do not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. Does the proposed change involve a significant reduction in the margin of

safety? No. Margin of safety is related to confidence in the ability of the fission

product barriers to perform their

accident mitigation functions. The

proposed TS changes involve TS 3.5.2

Action a, new TS 3.5.2 Action h, and

the provision in SR 4.5.2.a to address

non-conservative TS requirements.

Editorial changes are also proposed for

consistency and clarity. The proposed

changes provide better assurance that

the ECCS systems, subsystems, and

components are properly aligned to

support safe reactor operation consistent

with the licensing basis requirements.

The proposed changes do not physically

alter any SSC. There will be no effect on

those SSCs necessary to assure the

accomplishment of specified functions.

There will be no impact on the

overpower limit, departure from

nucleate boiling ratio (DNBR) limits, loss of cooling accident peak cladding

temperature (LOCA PCT), or any other

margin of safety. The applicable

radiological dose consequence

acceptance criteria will continue to be

met. Therefore, the proposed changes do

not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the licensees 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 S.

Blair, Managing AttorneyNuclear, Florida Power & Light Company, P.O.

Box 14000, Juno Beach, FL 33408-0420.

NRC Branch Chief:

Shana R. Helton.

Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316, Donald C. Cook Nuclear Plant, Units 1

and 2, Berrien County, Michigan Date of amendment request:

February 6, 2015. A publicly-available version is

in ADAMS under Accession No.

ML15041A069.

Description of amendment request:

The proposed amendments would

modify the technical specifications

requirements for unavailable barriers by

adding limiting condition for operation VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00092Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11479 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices (LCO) 3.0.8. The changes are consistent with the NRC approved Technical

Specification Task Force (TSTF)

Standard Technical Specification

change TSTF-427, Allowance for Non-

Technical Specification Barrier

Degradation on Supported System

OPERABILITY, Revision 2.

Basis for proposed no significant hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has affirmed the applicability

of the model proposed no significant

hazards consideration published on

October 3, 2006 (71 FR 58444), Notice

of Availability of the Model Safety

Evaluation. The findings presented in

that evaluation are presented below:

Criterion 1The Proposed Change Does Not Involve a Significant Increase in the

Probability or Consequences of an

Accident Previously Evaluated The proposed change allows a delay time for entering a supported system

technical specification (TS) when the

inoperability is due solely to an

unavailable barrier if risk is assessed

and managed. The postulated initiating

events which may require a functional

barrier are limited to those with low

frequencies of occurrence, and the

overall TS system safety function would

still be available for the majority of

anticipated challenges. Therefore, the

probability of an accident previously

evaluated is not significantly increased, if at all. The consequences of an

accident while relying on the allowance

provided by proposed LCO 3.0.8 are no

different than the consequences of an

accident while relying on the TS

required actions in effect without the

allowance provided by proposed LCO 3.0.8. Therefore, the consequences of an

accident previously evaluated are not

significantly affected by this change.

The addition of a requirement to assess

and manage the risk introduced by this

change will further minimize possible

concerns. Therefore, this change does

not involve a significant increase in the

probability or consequences of an

accident previously evaluated.

Criterion 2The Proposed Change Does Not Create the Possibility of a New or

Different Kind of Accident From any

Previously Evaluated The proposed change does not involve a physical alteration of the plant (no new or different type of equipment

will be installed). Allowing delay times

for entering supported system TS when

inoperability is due solely to an

unavailable barrier, if risk is assessed

and managed, will not introduce new

failure modes or effects and will not, in

the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents

previously evaluated. The addition of a

requirement to assess and manage the

risk introduced by this change will

further minimize possible concerns.

Thus, this change does not create the

possibility of a new or different kind of

accident from an accident previously evaluated.

Criterion 3The Proposed Change Does Not Involve a Significant Reduction in

a Margin of Safety The proposed change allows a delay time for entering a supported system TS

when the inoperability is due solely to

an unavailable barrier, if risk is assessed

and managed. The postulated initiating

events which may require a functional

barrier are limited to those with low

frequencies of occurrence, and the

overall TS system safety function would

still be available for the majority of

anticipated challenges. The risk impact

of the proposed TS changes was

assessed following the three-tiered

approach recommended in RG

[Regulatory Guide] 1.177. A bounding

risk assessment was performed to justify

the proposed TS changes. This

application of LCO 3.0.8 is predicated

upon the licensees performance of a

risk assessment and the management of

plant risk. The net change to the margin

of safety is insignificant as indicated by

the anticipated low levels of associated

risk (ICCDP [incremental conditional

core damage probability] and ICLERP

[incremental large early release

probability]) as shown in Table 1 of

Section 3.1.1 in the Safety Evaluation.

Therefore, this change does not involve

a significant reduction in a margin of

safety. The NRC staff has reviewed the analysis and, based on this review, it appears that the

three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards consideration.

Attorney for licensee:

Robert B. Haemer, Senior Nuclear Counsel, One Cook Place, Bridgman, Michigan 49106.

NRC Branch Chief:

David L. Pelton.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County, Pennsylvania Date of amendment requests:

October 27, 2014. A publicly-available version is

available in ADAMS under Accession No.

ML14317A052.

Description of amendment requests:

The proposed amendments will modify the

Susquehanna technical specifications (TS).

Specifically, the proposed amendments will

modify the TS by relocating specific

surveillance frequencies to a licensee-

controlled program, the Surveillance Frequency Control Program (SFCP), with implementation of Nuclear Energy Institute (NEI) 04-10, Risk-Informed Technical

Specifications Initiative 5b, Risk-Informed

Method for Control of Surveillance Frequencies (ADAMS Accession No.

ML071360456). The changes are consistent

with NRC-approved TS Task Force (TSTF)

Standard TS change TSTF-425, Relocate

Surveillance Frequencies to Licensee

Control-Risk Informed Technical

Specifications Task Force (RITSTF) Initiative

5b, Revision 3 (ADAMS Accession No.

ML090850642). The Federal Register notice published on July 6, 2009 (74 FR 31996),

announced the availability of this TSTF improvement, and included a model no significant hazards consideration and safety

evaluation.

Basis for proposed no significant hazards consideration determination:

An analysis of the no significant hazards consideration was

presented in the TSTF-425. The licensee has

affirmed its applicability of the model 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 relocates the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control

Program. Surveillance frequencies are not an

initiator to any accident previously

evaluated. As a result, the probability of any

accident previously evaluated is not

significantly increased. The systems and

components required by the technical

specifications for which the surveillance

frequencies are relocated are still required to

be operable, meet the acceptance criteria for

the surveillance requirements, and be

capable of performing any mitigation

function assumed in the accident analysis.

As a result, the consequences of any accident

previously evaluated are not significantly

increased.

Therefore, the proposed change does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

No new or different accidents result from utilizing the proposed change. The changes

do not involve a physical alteration of the

plant (i.e., no new or different type of equipment will be installed) or a change in

the methods governing normal plant

operation. In addition, the changes do not

impose any new or different requirements.

The changes do not alter assumptions made

in the safety analysis. The proposed changes

are consistent with the safety analysis

assumptions and current plant operating

practice.

Therefore, the proposed changes do not create the possibility of a new or different

kind of accident from any accident

previously evaluated. VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00093Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11480 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices

3. Does the proposed change involve a significant reduction in the margin of safety?

Response: No.

The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified

in applicable codes and standards (or

alternatives approved for use by the NRC)

will continue to be met as described in the

plant licensing basis (including the final

safety analysis report and bases to TS), since these are not affected by changes to the

surveillance frequencies. Similarly, there is

no impact to safety analysis acceptance

criteria as described in the plant licensing

basis. To evaluate a change in the relocated

surveillance frequency, PPL will perform a

risk evaluation using the guidance contained

in NRC approved NEI 04-10, Rev. 1 in

accordance with the TS SFCP. NEI 04-10, Rev. 1, methodology provides reasonable

acceptance guidelines and methods for

evaluating the risk increase of proposed

changes to surveillance frequencies

consistent with Regulatory Guide 1.177.

Therefore, the proposed changes do not involve a significant reduction in a margin of

safety. The NRC staff has reviewed the licensees 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:

Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL

Services Corporation, 2 North Ninth St.,

GENTW3, Allentown, PA 18101-1179.

NRC Branch Chief:

Douglas A.

Broaddus.

Southern Nuclear Operating Company, Inc. (SNC), Docket Nos. 50-424 and 50-

425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia Date of amendment request:

July 18, 2014. A publicly-available version is in

ADAMS under Accession Package No.

ML14203A124.

Description of amendment request:

The licensee requested 23 revisions to

the Technical Specifications (TSs).

These revisions adopt various

previously NRC-approved Technical

Specifications Task Force (TSTF)

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:

1: TSTF-2-A, Revision 1, Relocate the 10 Year Sediment Cleaning of the Fuel Oil

Storage Tank to Licensee Control for TS

pages 3.8.3-3 and 3.8.3-4

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 Surveillance Requirement for performing

sediment cleaning of diesel fuel oil storage

tanks every 10 years from the Technical

Specifications and places it under licensee

control. Diesel fuel oil storage tank cleaning

is not an initiator of any accident previously

evaluated. This change will have no effect on

diesel generator fuel oil quality, which is

tested in accordance with other Technical

Specifications requirements. Removing the

diesel fuel oil storage tank sediment cleaning

requirements from the Technical

Specifications will have no effect on the

ability to mitigate an accident.

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 removes the requirement to clean sediment from the

diesel fuel oil storage tank from the

Technical Specifications and places it under

licensee control. The margin of safety

provided by the fuel oil storage tank

sediment cleaning is unaffected by this

relocation because the quality of diesel fuel

oil is tested in accordance with other

Technical Specifications requirements.

Therefore, the proposed change does not involve a significant reduction in a margin of

safety. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

2: TSTF-27-A, Revision 3, Revise SR

[Surveillance Requirement] Frequency for

Minimum Temperature for Criticality for TS 3.4.2, TS Page 3.4.2-1

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

[reactor coolant system] 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 initial Frequency for this Surveillance

will be 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This will ensure that T avg [average temperature] is logged at appropriate

intervals (in addition to strip chart recorders

and computer logging of temperature). The

measurement of RCS 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

operators attention.

Therefore, the proposed change does not involve a significant reduction in a margin of

safety. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified. VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00094Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11481 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices 3: TSTF-28-A, Revision 0, Delete Unnecessary Action to Measure Gross

Specific Activity, TS 3.4.16, TS page 3.4-16

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 Required Action B.1 of Specification 3.4.16, RCS

Specific Activity, which requires verifying

that Dose Equivalent I-131 specific activity is

within limits. Determination of Dose

Equivalent I-131 is not an initiator of any

accident previously evaluated. Determination

of Dose Equivalent I-131 has 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 eliminates a Required Action. The activities performed

under the Required Action will still be

performed to determine if the LCO is met or

the plant will exit the Applicability of the

Specification. In either case, the presence of

the Required Action does not provide any

significant margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of

safety. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

4: TSTF-45-A, Revision 2, Exempt Verification of CIVs that are Locked, Sealed

or Otherwise Secured, TS 3.6.3, TS pages

3.6.3-4, 3.6.3-5

1. Does the proposed amendment involve a significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

The proposed change exempts containment isolation valves (CIVs) located inside and

outside of containment that are locked, sealed, or otherwise secured in position from

the periodic verification of valve position

required by Surveillance Requirements 3.6.3.3 and 3.6.2.4. The exempted valves are

verified to be in the correct position upon being locked, sealed, or secured. Because the valves are in the condition assumed in the

accident analysis, the proposed change will

not affect the initiators or mitigation of any

accident previously evaluated.

Therefore, the proposed change does not involve a significant increase in the

probability or consequences of any accident previously evaluated.

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 replaces the periodic verification of valve position with

verification of valve position followed by

locking, sealing, or otherwise securing the

valve in position. Periodic verification is also

effective in detecting valve mispositioning.

However, verification followed by securing

the valve in position is effective in

preventing valve mispositioning. Therefore, the proposed change does not involve a

significant reduction in a margin of safety.

Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

5: TSTF-46-A, Revision 1, Clarify the CIV Surveillance to Apply Only to Automatic

Isolation Valves, TS 3.6.3, TS page 3.6.3.5

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 3.6.3.5, and the associated Bases, to delete

the requirement to verify the isolation time

of each power operated containment

isolation valve (CIV) and only require

verification of closure time for each

automatic power operated 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 3.6.3.5. 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

isolation signal are within the scope of the

SR 3.6.3.5. 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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

6: TSTF-87-A, Revision 2, Revise RTBs

[Reactor Trip Breaker] Open and CRDM

[Control Rod Drive Mechanism] De-

energized Actions to Incapable of Rod

Withdrawal, TS 3.4.5, TS Pages 3.4.5-2, 3.4.9-1 1. Does the proposed amendment involve a significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No. VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00095Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11482 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices This change revises the Required Actions for LCO 3.4.5, RCS LoopsMode 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 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 the applicable Conditions

of LCO 3.4.5 and LCO 3.4.9 are 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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

7: TSTF-95-A, Revision 0, Revise Completion Time for Reducing Power Range

High trip Setpoint from 8 to 72 Hours, TS 3.2.1, TS Pages 3.2.1-1 and 3.2.2-1

1. Does the proposed amendment involve a significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

The proposed change extends the time allowed to reduce the Power Range Neutron

FluxHigh trip setpoint when Specification

3.2.1, Heat Flux Hot Channel Factor, or

Specification 3.2.2, Nuclear Enthalpy Rise

Hot Channel Factor, are not within their

limits. Both specifications require a power

reduction followed by a reduction in the

Power Range Neutron FluxHigh trip

setpoint. Because reactor power has been

reduced, the reactor core power distribution

limits are within the assumptions of the

accident analysis. Reducing the Power Range

Neutron FluxHigh trip setpoints ensures

that reactor power is not inadvertently

increased. Reducing the Power Range

Neutron FluxHigh trip setpoints is not an

initiator to any accident previously

evaluated. The consequences of any accident

previously evaluated with the Power Range

Neutron FluxHigh trip setpoints not

reduced are no different under the proposed

Completion Time than under 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 provides additional time before requiring the Power Range

Neutron FluxHigh trip setpoint be reduced

when the reactor core power distribution

limits are not met. The manual reduction in

reactor power required by the specifications

provides the necessary margin of safety for

this condition. Reducing the Power Range

Neutron FluxHigh trip setpoints carries an increased risk of a reactor trip. Delaying the trip setpoint reduction until the power

reduction has been completed and the

condition is verified will minimize overall

plant risk.

Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

8: TSTF-110-A, Revision 2, Delete SR Frequencies Based on Inoperable Alarms,

TS 3.1, TS pages 3.1.4-3, 3.1.6-3, 3.2.3-1, 3.2.4-4 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 surveillance Frequencies associated with inoperable

alarms (rod position deviation monitor, rod

insertion limit monitor, AFD [Axial Flux

Difference] monitor and QPTR [Quadrant

Power Tilt Ratio] alarm) from the Technical

Specifications and places the actions in plant

administrative procedures. The subject plant

alarms are not an initiator of any accident

previously evaluated. The subject plant

alarms are not used to mitigate any accident

previously evaluated, as the control room

indications of these parameters are sufficient

to alert the operator of an abnormal condition

without the alarms. The alarms are not

credited in the accident analysis.

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 removes surveillance Frequencies associated with inoperable

alarms (rod position deviation monitor, rod

insertion limit monitor, AFD monitor and

QPTR alarm) from the Technical

Specifications and places the actions in plant

administrative procedures. The alarms are

not being removed from the plant. The

actions to be taken when the alarms are not

available are proposed to be controlled under

licensee administrative procedures. As a

result, plant operation is unaffected by this

change and there is no effect on a margin of

safety. VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00096Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11483 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices Therefore, the proposed change does not involve a significant reduction in a margin of

safety. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

9: TSTF-142-A, Revision 0, Increase the Completion Time When the Core Reactivity

Balance is Not Within Limit, TS 3.1.2, TS

Page 3.1.2-1

1. Does the proposed amendment involve a significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

The proposed change extends the Completion Time to take the Required

Actions when measured core reactivity is not

within the specified limit of the predicted

values. The Completion Time to respond to

a difference between predicted and measured

core reactivity is not an initiator to any

accident previously evaluated. The

consequences of an accident during the

proposed Completion Time are no different

from the consequences of an accident 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 provides additional time to investigate and to implement

appropriate operating restrictions when

measured core reactivity is not within the

specified limit of the predicted values. The

additional time will not have a significant

effect on plant safety due to the

conservatisms used in designing the reactor

core and performing the safety analyses and

the low probability of an accident or

transient which would approach the core

design limits during the additional time.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

10: TSTF-234-A, Revision 1, Add Action for More Than One [D]RPI Inoperable, TS 3.1.7, TS Pages 3.1.7-1 and 3.1.7-2.

1. Does the proposed amendment involve a significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

The proposed change provides a Condition and Required Actions for more than one

inoperable digital rod position indicator (DRPI) per rod group. The DRPIs are not an

initiator of any accident previously

evaluated. The DRPIs are one indication used

by operators to verify control rod insertion

following an accident, however other

indications are available. Therefore, allowing

a finite period to time to correct more than

one inoperable DRPI prior to requiring a

plant shutdown will not result in a

significant increase in the consequences 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 provides time to correct the condition of more than one DRPI

inoperable in a rod group. Compensatory

measures are required to verify that the rods

monitored by the inoperable DRPIs are not

moved to ensure that there is no effect on

core reactivity. Requiring a plant shutdown

with inoperable rod position indications

introduces plant risk and should not be

initiated unless the rod position indication

cannot be repaired in a reasonable period of

time. As a result, the safety benefit provided

by the proposed Condition offsets the small

decrease in safety resulting from continued

operation with more than one inoperable

DRPI. Therefore, the proposed change does not involve a significant reduction in a margin of

safety. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

11: TSTF-245-A, Revision 1, AFW Train Operable When in Service, TS 3.7.5, TS

Page 3.7.5-3

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?

Responses: 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 VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00097Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11484 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices or limiting conditions for operation are determined, nor is there any adverse effect on

those plant systems necessary to assure the

accomplishment of protection functions. The

proposed change will not result in plant

operation in a configuration outside the

design basis.

Therefore, the proposed change does not involve a significant reduction in a margin of

safety. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

12: TSTF-247-A, Revision 0, Provide Separate Condition Entry for Each [Power

Operated Relief Valve] PORV and Block

Valve, TS 3.4.11, TS Pages 3.4.11-1, 3.4.11-

2, 3.4.11-3

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, 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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

13: TSTF-248-A, Revision 0, Revise Shutdown Margin Definition for Stuck Rod

Exception, TS 1.1, TS Page 1.1-6

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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

14: TSTF-266-A, Revision 3, Eliminate the Remote Shutdown System Table of

Instrumentation and Controls, TS 3.3.4, TS

Pages 3.3.4-1, 3.3.4-3

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. VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00098Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11485 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices 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 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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

15: TSTF-272-A, Revision 1, Refueling Boron Concentration Clarification, TS 3.9.1, TS Page 3.9.1-1

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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

16: TSTF-273-A, Revision 2, Safety Function Determination Program

Clarifications, TS 5.5.15, TS Page 5.5-15

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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

17: TSTF-284-A, Revision 3, Add Met vs.

Perform to Technical Specification 1.4, Frequency, TS 1.4, TS 3.4, TS 3.9, TS Pages

1.4-1, 1.4-4, 3.4.11-3, 3.4.12-4 and 3.9.4-2

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 VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00099Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11486 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices 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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

18: TSTF-308-A, Revision 1, Determination of Cumulative and Projected Dose

Contributions in RECP [Radioactive Effluent

Controls Program], TS 5.5.4, TS Page 5.5-3

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises Specification 5.5.4, Radioactive Effluent Controls

Program, paragraph e, to describe the

original intent of the dose projections. The

cumulative and projection of doses due to liquid releases are not an assumption in any

accident previously evaluated and have no

effect on the mitigation of any accident

previously evaluated.

Therefore, the proposed change does not involve a significant increase in the

probability or consequences of any accident

previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

The proposed change does not involve a physical alteration to the plant (i.e., no new or different type of equipment will be

installed) or a change to the methods governing normal plant operation. The changes do not alter the assumptions made

in the safety analysis.

Therefore, the proposed change does not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change revises Specification 5.5.4, Radioactive Effluent Controls

Program, paragraph e, to describe the

original intent of the dose projections. The

cumulative and projection of doses due to

liquid releases are administrative tools to

assure compliance with regulatory limits.

The proposed change revises the requirement

to clarify the intent, thereby improving the

administrative control over this process. As

a result, any effect on the margin of safety

should be minimal.

Therefore, the proposed change does not involve a significant reduction in a margin of

safety. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

19: TSTF-312-A, Revision 1, Administrative Control of Containment

Penetrations, TS 3.9.4, TS Page 3.9.4-1

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

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)

accident analysis.

Therefore, the proposed change does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the proposed 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 3.9.4 provides measures to ensure that the dose consequences of a postulated FHA

inside containment are minimized. The

proposed change to LCO 3.9.4 will allow

penetration flow path(s) to be open during

refueling operations under administrative

control. These administrative controls will

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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified. VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00100Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11487 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices 20: TSTF-314-A, Revision 0, Require Static and Transient F Q Measurement, TS 3.1.4, 3.2.4, TS Pages 3.1.4-2, 3.2.4-1, 3.2.4-3

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, F Q (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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

21: TSTF-340-A, Revision 3, Allow 7 Day Completion Time for a TurbineDriven

AFW Pump Inoperable, TS 3.7.5, TS Page

3.7.5-1 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 AFW 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 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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

22: TSTF-343-A, Revision 1, Containment Structural Integrity, TS 5.5, TS Page 5.5-16

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 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 steel containment liner

plate for the purpose of the Containment

Leakage Rate Testing Program.

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 change revises 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

change affects the frequency of visual

examinations that will be performed for the

steel containment liner plate for the purpose

of the Containment Leakage Rate Testing

Program. 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 change affects the

frequency of visual examinations that will be

performed for the steel containment liner

plate for the purpose of the Containment

Leakage Rate Testing Program. 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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

23: TSTF-349-A, Revision 1, Add Note to LCO 3.9.5 Allowing Shutdown Cooling

Loops Removal From Operation, TS 3.9.6, TS Page 3.9.6-1

1. Does the proposed amendment involve a significant increase in the probability or VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00101Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11488 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices consequences of an accident previously evaluated?

Response: No.

The proposed change adds an LCO Note to LCO 3.9.6, 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 3.9.6, 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. Based on the above, SNC concludes that the proposed amendment does not involve a

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

The NRC staff has reviewed the licensees 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

Operating Company, 40 Iverness Center

Parkway, Birmingham, AL 35201.

NRC Branch Chief:

Robert J.

Pascarelli.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, North

Anna Power, Unit Nos. 1 and 2, Louisa

County, Virginia Date of amendment request:

February 4, 2015. A publicly-available version is

in ADAMS under Accession No.

ML15041A667.

Description of amendment request:

The proposed license amendment

requests the changes to the Technical

Specification (TS) TS 3.1.7, Rod

Position Indication, to provide an

additional monitoring option for an

inoperable control rod position

indicator. Specifically, the proposed

changes would allow monitoring of

control rod drive mechanism stationary

gripper coil voltage every eight hours as

an alternative to using the movable in

core detectors every eight hours to

verify control rod position.

Basis for proposed no significant hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below: 1. Does the proposed change involve a significant increase in the probability or

consequences of an accident previously

evaluated?

Response: No.

The proposed change provides an alternative method for verifying rod position

of one rod. The proposed change meets the

intent of the current specification in that it

ensures verification of position of the rod

once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The proposed change

provides only an alternative method of

monitoring rod position and does not change

the assumptions or results of any previously

evaluated accident.

Therefore, operation of the facility in accordance with the proposed amendment

would not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of

accident from any accident previously

evaluated?

Response: No.

The proposed change provides only an alternative method of determining the

position of one rod. No new accident

initiators are introduced by the proposed alternative manner of performing rod position verification. The proposed change

does not affect the reactor protection system.

Hence, no new failure modes are created that

would cause a new or different kind of

accidents from any accident previously

evaluated.

Therefore, operation of the facility in accordance with the proposed amendments

would not create the possibility of a new or

different kind of accident from any

previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The basis of TS 3.1.7 states that the operability of the rod position indicators is

required to determine control rod positions

and thereby ensure compliance with the

control rod alignment and insertion limits.

The proposed change does not alter the

requirement to determine rod position but

provides an alternative method for

determining the position of the affected rod.

As a result, the initial conditions of the

accident analysis are preserved and the

consequences of previously analyzed

accidents are unaffected.

Therefore, operation of the facility in accordance with the proposed amendment

would not involve a significant reduction in

a margin of safety.

Based on the above, Dominion concludes that the proposed amendment presents no

significant hazards consideration under the

standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant

hazards consideration is justified.

The NRC staff has reviewed the licensees analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Attorney for licensee:

Lillian M.

Cuoco, Senior Counsel, Dominion

Resources Services, Inc., 120 Tredegar

Street, RS-2, Richmond, VA 23219.

NRC Branch Chief:

Robert Pascarelli.

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

Commissions rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commissions 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 VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00102Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11489 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices 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 Commissions 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.

Dominion Energy Kewaunee, Inc.

Docket No. 50-305, Kewaunee Power

Station, Kewaunee County, Wisconsin Date of application for amendment:

May 29, 2013, as supplemented by

letters dated September 23, October 15, October 17, October 31, and November

7, 2013, and January 7, March 13, April

29, and October 6, 2014, and January 15, 2015. Brief description of amendment:

The amendment revised the Renewed

Facility Operating License and

associated Technical Specifications to

conform to the permanent shutdown

and defueled status of the facility. It also

denied a proposal to delete paragraphs

1.B, 1.I, and 1.J of the Kewaunee

Operating License.

Date of issuance:

February 13, 2015.

Effective date:

As of the date of issuance and shall be implemented

within 90 days from the date of

issuance.

Amendment No.:

215. A publicly-available version is in ADAMS under

Accession No. ML14237A045;

documents related to this amendment

are listed in the Safety Evaluation

enclosed with the amendment.

Renewed Facility Operating License No. DPR-43:

The amendment revised the renewed facility operating license

and Technical Specifications.

Date of initial notice in Federal Register: August 20, 2013 (78 FR 51224). The supplemental letters dated September 23, October 15, October 17, October 31, and November 7, 2013, and

January 7, March 13, April 29, and

October 6, 2014, and January 15, 2015, provided additional information that

clarified the application, did not expand

the scope of the application as originally

noticed, and did not change the staffs

original proposed no significant hazards

consideration determination as

published in the Federal Register. The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated February 13, 2015. No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370 McGuire

Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina Duke Energy Carolinas, LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and

3, Oconee County, South Carolina Date of application for amendments:

July 21, 2014.

Brief description of amendments:

The amendment revises the licensed

operator training requirements to be

consistent with the National Academy

for Nuclear Training (NANT) program.

Additionally, the amendment makes

administrative changes to Technical

Specification Sections 5.1, Responsibility; 5.2, Organization;

5.3, Unit Staff Qualifications; 5.5, Programs and Manuals; and for

Catawba and McGuire, Section 5.7, High Radiation Area.

Date of issuance:

February 12, 2015.

Effective date:

This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

Amendment Nos.:

273, 269, 276, 256, 389, 391, and 390. A publicly-available

version is available in ADAMS under

Accession No. ML15002A324.

Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, NPF-17, DPR-38, DPR-47, and DPR-55:

Amendments revised the licenses and

Technical Specifications.

Date of initial notice in Federal Register: November 12, 2014 (79 FR 67199). The Commissions related evaluation of the amendments is contained in a

Safety Evaluation dated February 12, 2015. No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County, Arkansas Date of application for amendment:

December 17, 2012, as supplemented by letters dated November 7, and December

4, 2013; January 6, May 22, June 30, August 7, September 24, and December

9, 2014. Brief description of amendment:

The amendment authorized the transition of

the Arkansas Nuclear One, Unit No. 2, fire protection program to a risk-

informed, performance-based program

based on National Fire Protection

Association (NFPA) 805, in accordance

with 10 CFR 50.48(c). NFPA 805 allows

the use of performance-based methods

such as fire modeling and risk-informed

methods such as fire probabilistic risk

assessment to demonstrate compliance

with the nuclear safety performance

criteria.

Date of issuance:

February 18, 2015.

Effective date:

As of its date of issuance and shall be implemented by 6

months from the date of issuance.

Amendment No.:

300. A publicly-available version is in ADAMS under

Accession No. ML14356A227;

documents related to this amendment

are listed in the Safety Evaluation

enclosed with the amendment.

Renewed Facility Operating License No. NPR-6:

Amendment revised the License and Technical Specifications.

Date of initial notice in Federal Register: July 23, 2013 (78 FR 44171).

The supplemental letters dated November 7 and December 4, 2013; and

January 6, May 22, June 30, August 7, September 24, and December 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 staffs

original proposed no significant hazards

consideration determination as

published in the Federal Register. The Commissions related evaluation of the amendment is contained in a Safety Evaluation dated February 18, 2015. No significant hazards consideration comments received: No.

Entergy Nuclear FitzPatrick, LLC and Entergy Nuclear Operations, Inc.,

Docket No. 50-333, James A. FitzPatrick

Nuclear Power Plant, Oswego County, New York Date of amendment request:

October 8, 2013, as supplemented by a letter

dated November 18, 2014.

Brief description of amendment:

The amendment modifies the Technical

Specifications (TSs) to reduce the

reactor steam dome pressure associated VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00103Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11490 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices with the Reactor Core Safety Limit from 785 psig to 685 psig in TS 2.1.1.1 and

TS 2.1.1.2. This change addresses the

potential to not meet the pressure/

thermal power/minimal critical power

ratio TS safety limit during a pressure

regulator failure-maximum demand (open) (PRFO) transient. The PRFO

transient was reported by General

Electric as a notification pursuant to

Title 10 of the Code of Federal Regulations, Part 21, Reporting of Defects and Noncompliance.

Date of issuance:

February 9, 2015.

Effective date:

As of the date of issuance and shall be implemented

within 30 days of issuance.

Amendment No.:

309. A publicly-available version is in ADAMS under

Accession No. ML15014A277;

documents related to this amendment

are listed in the Safety Evaluation

enclosed with the amendment.

Renewed Facility Operating License No. DPR-59:

Amendment revised the Renewed Facility Operating License and

Technical Specifications.

Date of initial notice in Federal Register: July 8, 2014 (79 FR 38589).

The supplemental letter dated

November 18, 2014, provided additional

information that clarified the

application, did not expand the scope of

the application as originally noticed, and did not change the staffs original

proposed no significant hazards

consideration determination as

published in the Federal Register. The Commissions related evaluation of the amendment is contained in a

Safety Evaluation dated February 9, 2015. No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc.,

Docket No. 50-271, Vermont Yankee

Nuclear Power Station, Vernon, Vermont Date of amendment request:

November 14, 2013, as supplemented by

letters dated June 9, 2014, August 6, 2014, and October 9, 2014.

Description of amendment request:

The amendment eliminates operability

requirements for secondary containment

when handling sufficiently decayed

irradiated fuel or a fuel cask following

a minimum of 13 days after the

permanent cessation of reactor

operation.

Date of Issuance:

February 12, 2015.

Effective date:

The license amendment becomes effective 13 days

after the licensees submittal of the

certifications, as required by 10 CFR

50.82(a)(1)(i) and (ii).

Amendment No.:

262. A publicly-available version is in ADAMS under

Accession No. ML14304A588;

documents related to this amendment

are listed in the Safety Evaluation

enclosed with the amendment.

Facility Operating License No. DPR-28: The amendment revised the Renewed Facility Operating License and

Technical Specifications.

Date of initial notice in Federal Register: September 16, 2014 (79 FR 55511). The supplemental letters dated June 9, 2014, August 6, 2014, and October 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 NRC staffs original proposed no

significant hazards consideration

determination as published in the

Federal Register. The Commissions related evaluation of this amendment is contained in a

Safety Evaluation dated February 12, 2015. No significant hazards consideration comments received: No.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold

Energy Center, Linn County, Iowa Date of amendment request:

June 23, 2014. Brief description of amendment:

The amendment revised the Technical

Specification (TS) requirements to

address NRC Generic Letter (GL) 2008-

01, Managing Gas Accumulation in

Emergency Core Cooling, Decay Heat

Removal, and Containment Spray

Systems, as described in TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas Accumulation.

Date of issuance:

February 10, 2015.

Effective date:

As of the date of issuance and shall be implemented

within 90 days.

Amendment No.:

290. A publicly-available version is in ADAMS under

Accession No. ML15014A200;

documents related to this amendment

are listed in the Safety Evaluation

enclosed with the amendment.

Renewed Facility Operating License No. DPR-49:

The amendment revised the Renewed Facility Operating License

and Technical Specifications.

Date of initial notice in Federal Register: September 30, 2014 (79 FR 58820). The Commissions related evaluation of the amendment is contained in a

Safety Evaluation dated February 10, 2015. No significant hazards consideration comments received: No NextEra Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.

1, Rockingham County, New Hampshire Date of amendment request:

June 24, 2014, as supplemented by letter dated

December 11, 2014.

Brief description of amendment:

The amendment revised the Seabrook

Technical Specifications (TSs).

Specifically, the amendment modifies

Seabrook TSs to address U.S. Nuclear

Regulatory Commission Generic Letter (GL) 2008-01, Managing Gas

Accumulation in Emergency Core

Cooling, Decay Heat Removal, and

Containment Spray Systems, as

described in TSTF-523, Revision 2, Generic Letter 2008-01, Managing Gas

Accumulation.

Date of issuance:

February 6, 2015.

Effective date:

As of its date of issuance and shall be implemented

within 60 days.

Amendment No.:

144. A publicly-available version is in ADAMS under

Accession No. ML14345A288;

documents related to this amendment

are listed in the Safety Evaluation

enclosed with the amendment.

Facility Operating License No. NPF-86: The amendment revised the License and TS. Date of initial notice in Federal Register: September 2, 2014 (79 FR 52066). The supplemental letter dated

December 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 staffs original

proposed no significant hazards

consideration determination as

published in the Federal Register. The Commissions related evaluation of the amendment is contained in a

Safety Evaluation dated February 6, 2015. No significant hazards consideration comments received:

No. South Carolina Electric and 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:

November 15, 2011, as supplemented by

letters dated November 22, 2011;

January 26 and October 10, 2012;

February 1, April 1, October 14, and

November 26, 2013; January 9, February

25, May 2, May 11, August 14, October

9, and December 11, 2014.

Brief description of amendment:

The amendment authorizes the transition of

the V.C. Summer fire protection

program to a risk-informed, performance-based program based on VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00104Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11491 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices National Fire Protection Association (NFPA) 805, Performance-Based

Standard for Fire Protection for Light

Water Reactor Electric Generating

Plants, 2001 Edition (NFPA 805), in

accordance with 10 CFR 50.48(c).

Date of issuance:

February 11, 2015.

Effective date:

This amendment is effective as of its date of issuance and

shall be implemented per the December

11, 2014, supplement, Attachment S, Table S-2 Implementation Items, requiring full implementation by March

31, 2016.

Amendment No.:

199. A publicly-available version is in ADAMS under

Accession No. ML14287A289;

documents related to this amendment

are listed in the Safety Evaluation

enclosed with the amendment.

Renewed Facility Operating License No. NPF-12:

Amendment revised the Facility Operating License.

Date of initial notice in Federal Register: August 14, 2012 (77 FR 48561). The supplemental letters dated

November 22, 2011; October 10, 2012;

February 1, April 1, October 14, and

November 26, 2013; January 9, February

25, May 2, May 11, August 14, October

9, and December 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 staffs original

proposed no significant hazards

consideration determination as

published in the Federal Register. The Commissions related evaluation of the amendment is contained in a

Safety Evaluation dated February 11, 2015. No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia, Docket No. 50-

366, Edwin I. Hatch Nuclear Plant (HNP), Unit No. 2, Appling County, Georgia Date of amendment request:

August 8, 2014, as supplemented by letters dated

September 8 and October 24, 2014.

Brief description of amendments:

The amendment revises the Technical

Specification value of the Safety Limit

Minimum Critical Power Ratio to

support operation in the next fuel cycle.

Date of issuance:

February 18, 2015.

Effective date:

As of the date of issuance and shall be implemented

prior to reactor startup following the

HNP, Unit 2, spring 2015 refueling

outage. Amendment No(s).:

218. A publicly-available version is in ADAMS under Accession No. ML15020A434; documents related to this amendment

are listed in the Safety Evaluation

enclosed with the amendment.

Renewed Facility Operating License Nos. DPR-57 and NPF-5:

Amendment revised the licenses and the Technical

Specifications.

Date of initial notice in Federal Register: January 6, 2015, (80 FR 536).

The Commissions related evaluation of the amendments is contained in a

Safety Evaluation dated February 18, 2015. No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project (STP), Units 1 and 2, Matagorda County, Texas Date of amendment request:

July 23, 2013, as supplemented by letters dated

May 12 (two letters), May 19, and

December 17, 2014.

Brief description of amendments:

The amendments revised the STP, Units 1

and 2, Fire Protection Program (FPP)

related to the alternate shutdown

capability. Specifically, it approves the following operator actions in the control room prior to evacuation due to a fire

for meeting the alternate shutdown

capability, in addition to manually

tripping the reactor that is currently

credited in the STP, Units 1 and 2, FPP

licensing basis:

• Initiate main steam line isolation

• Closing the pressurizer power-operated relief valves block valves

• Securing all reactor coolant pumps

• Closing feedwater isolation valves

• Securing the startup feedwater pump *Isolating reactor coolant system letdown *Securing the centrifugal charging pumps In addition, the licensee credits the automatic trip of the main turbine upon

the initiation of a manual reactor trip for

meeting the alternate shutdown

capability.

Date of issuance:

February 13, 2015.

Effective date:

As of the date of issuance and shall be implemented

within 45 days of issuance.

Amendment Nos.:

Unit 1203; Unit 2191. A publicly-available version is

in ADAMS under Accession No.

ML14339A170; documents related to

these amendments are listed in the

Safety Evaluation enclosed with the

amendments.

Facility Operating License Nos. NPF-76 and NPF-80:

The amendments revised the Facility Operating Licenses.

Date of initial notice in Federal Register: October 29, 2013 (78 FR 64546). The supplements dated May 12 (two letters), May 19, and December 17, 2014, provided additional information

that clarified the application, did not

expand the scope of the application as

originally noticed, and did not change

the staffs original proposed no

significant hazards consideration

determination as published in the

Federal Register. The Commissions related evaluation of the amendments is contained in a

Safety Evaluation dated February 13, 2015. No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No.

50-259, Browns Ferry Nuclear Plant, Unit 1, Limestone County, Alabama Date of amendment request:

December 18, 2013, as supplemented by

letter dated June 13, 2014.

Brief description of amendment:

The amendment revised the Technical

Specification (TS) 3.4.9, RCS [Reactor

Coolant System] Pressure and

Temperature (P/T) Limits, Figures

3.4.9-1 through 3.4.9-2. The P/T limits

are based on proprietary topical report

NEDC-33178P-A, Revision 1, GE

[General Electric] Hitachi Nuclear

Energy Methodology for Development of

Reactor Pressure Vessel Pressure-

Temperature Curves. NEDO-33178-A, Revision 1 is the non-proprietary

version of the NRC-approved topical

report. Date of issuance:

February 2, 2015.

Effective date:

As of the date of issuance and shall be implemented

within 60 days of issuance.

Amendment No.:

287. A publicly available version is in ADAMS under

Accession No. ML14325A501;

documents related to this amendment

are listed in the Safety Evaluation (SE)

enclosed with the amendment.

Renewed Facility Operating License No. DPR-33:

Amendment revised the TSs and the Operating License.

Date of initial notice in Federal Register: May 6, 2014 (79 FR 25902).

The supplemental letter dated June 13, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change

the staffs original proposed no

significant hazards consideration

determination as published in the

Federal Register. The Commissions related evaluation of the amendment is contained in the SE

dated February 2, 2015.

No significant hazards consideration comments received: No.

Dated at Rockville, Maryland, this 23rd day of February 2015. VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00105Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES 11492 Federal Register/Vol. 80, No. 41/Tuesday, March 3, 2015/Notices For the Nuclear Regulatory Commission.

Michele G. Evans, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor

Regulation.

[FR Doc. 2015-04298 Filed 3-2-15; 8:45 am]

BILLING CODE 7590-01-P NUCLEAR REGULATORY COMMISSION

[NRC-2015-0030]

Applications and Amendments to

Facility Operating Licenses and

Combined Licenses Involving

Proposed No Significant Hazards

Considerations and Containing

Sensitive Unclassified Non-Safeguards

Information and Order Imposing

Procedures for Access to Sensitive

Unclassified Non-Safeguards

Information AGENCY: Nuclear Regulatory Commission.

ACTION: License amendment request; opportunity to comment, request a

hearing, and petition for leave to

intervene; order.

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) received and is

considering approval of four

amendment requests. The amendment

requests are for Braidwood Station, Units 1 and 2, and Byron Station, Units

1 and 2; Peach Bottom Atomic Power

Station, Unit 2; Diablo Canyon Nuclear

Power Plant, Units 1 and 2; and Vogtle

Electric Generating Plant, Units 1 and 2, Joseph M. Farley Nuclear Plant, Units 1

and 2, and Edwin I. Hatch Nuclear

Plant, Units 1 and 2. The NRC proposes

to determine that each amendment

request involves no significant hazards

consideration. In addition, each

amendment request contains sensitive

unclassified non-safeguards information (SUNSI). DATES: Comments must be filed by April 2, 2015. A request for a hearing must be

filed by May 4, 2015. Any potential party as defined in §2.4 of Title 10 of

the Code of Federal Regulations (10 CFR), who believes access to SUNSI is

necessary to respond to this notice must

request document access by March 13, 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 http://www.regulations.gov and search for Docket ID NRC-2015-0030. Address

questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: Carol.Gallagher@nrc.gov. *Mail comments to:

Cindy Bladey, Office of Administration, Mail Stop:

O12-H08, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-

0001. For additional direction on obtaining information and submitting comments, see Obtaining Information and

Submitting Comments in the SUPPLEMENTARYINFORMATION section of this document. FORFURTHERINFORMATIONCONTACT

Shirley Rohrer, Office of Nuclear

Reactor Regulation, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-

5411; email:

Shirley.Rohrer@nrc.gov. SUPPLEMENTARYINFORMATION

I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2015-0030 when contacting the NRC about

the availability of information for this

action. You may obtain publicly-

available information related to this

action by any of the following methods:

• Federal Rulemaking Web site:

Go to http://www.regulations.gov and search for Docket ID NRC-2015-0030.

• NRCs Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the

ADAMS Public Documents collection at

http://www.nrc.gov/reading-rm/

adams.html.

To begin the search, select ADAMS Public Documents and then select Begin Web-based ADAMS Search. For problems with ADAMS, please contact the NRCs 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. *NRCs PDR:

You may examine and purchase copies of public documents at

the NRCs PDR, Room O1-F21, One

White Flint North, 11555 Rockville

Pike, Rockville, Maryland 20852.

B. Submitting Comments Please include Docket ID NRC-2015-0030, facility name, unit number(s),

application date, and subject in your

comment submission.

The NRC cautions you not to include identifying or contact information that

you do not want to be publicly

disclosed in your comment submission.

The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into ADAMS.

The NRC does not routinely edit

comment submissions to remove

identifying or contact information.

If you are requesting or aggregating comments from other persons for

submission to the NRC, then you should

inform those persons not to include

identifying or contact information that

they do not want to be publicly

disclosed in their comment submission.

Your request should state that the NRC

does not routinely edit comment

submissions to remove such information

before making the comment

submissions available to the public or entering the comment submissions into

ADAMS. II. Background Pursuant to Section 189a.(2) of the Atomic Energy Act of 1954, as amended (the Act), the NRC is publishing this

notice. The Act requires the

Commission to publish notice of any

amendments issued, or proposed to be

issued and grants the Commission the

authority to issue and make

immediately effective any amendment

to an operating license or combined

license, as applicable, upon a

determination by the Commission that

such amendment involves no significant

hazards consideration, notwithstanding

the pendency before the Commission of

a request for a hearing from any person.

This notice includes notices of amendments containing SUNSI.

III. Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses and Combined Licenses, Proposed No Significant Hazards

Consideration Determination, and

Opportunity for a Hearing The Commission has made a proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commissions regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated, or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated, or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public comments on this proposed

determination. Any comments received

within 30 days after the date of VerDate Sep<11>2014 19:24 Mar 02, 2015Jkt 235001PO 00000Frm 00106Fmt 4703Sfmt 4703E:\FR\FM\03MRN1.SGM03MRN1 mstockstill on DSK4VPTVN1PROD with NOTICES