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 OMB’s 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 hours (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 NRC’s April 28, 2014, 

Staff Requirements—COMAMM–14– 

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 

SUPPLEMENTARY INFORMATION section of 

this document. 

FOR FURTHER INFORMATION CONTACT: 

Shirley Rohrer, Office of Nuclear 

Reactor Regulation, U.S. Nuclear 

Regulatory Commission, Washington DC 

20555–0001; telephone: 301–415–5411, 

email: Shirley.Rohrer@nrc.gov. 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11473 

SUPPLEMENTARY INFORMATION: 

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 publiclyavailable 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. 

• NRC’s Agencywide Documents 

Access and Management System 

(ADAMS): You may obtain publiclyavailable documents online in the 

ADAMS Public Documents collection at 

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

adams.html. To begin the search, select 

‘‘ADAMS Public Documents’’ and then 

select ‘‘Begin Web-based ADAMS 

Search.’’ For problems with ADAMS, 

please contact the NRC’s Public 

Document Room (PDR) reference staff at 

1–800–397–4209, 301–415–4737, or by 

email to pdr.resource@nrc.gov. The 

ADAMS accession number for each 

document referenced (if it is available in 

ADAMS) is provided the first time that 

it is mentioned in the SUPPLEMENTARY

INFORMATION section. 

• NRC’s PDR: You may examine and 

purchase copies of public documents at 

the NRC’s PDR, Room O1–F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852. 

B. Submitting Comments 

Please include Docket ID NRC–2015– 

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 Commission’s 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 Commission’s ‘‘Agency Rules 

of Practice and Procedure’’ in 10 CFR 

part 2. Interested person(s) should 

consult a current copy of 10 CFR 2.309, 

which is available at the NRC’s PDR, 

located at One White Flint North, Room 

O1–F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland 20852. The 

NRC’s regulations are accessible 

electronically from the NRC Library on 

the NRC’s Web site at http://

www.nrc.gov/reading-rm/doccollections/cfr/. If a request for a hearing 

or petition for leave to intervene is filed 

by the above date, the Commission or a 

presiding officer designated by the 

Commission or by the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board Panel, will 

rule on the request and/or petition; and 

the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the requestor/

petitioner seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the requestor/petitioner shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the requestor/petitioner 

intends to rely in proving the contention 

at the hearing. The requestor/petitioner 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the requestor/petitioner intends 

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 NRC’s E-Filing rule 

(72 FR 49139; August 28, 2007). The EFiling 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 NRCissued digital ID certificate). Based upon 

this information, the Secretary will 

establish an electronic docket for the 

hearing in this proceeding if the 

Secretary has not already established an 

electronic docket. 

Information about applying for a 

digital ID certificate is available on the 

NRC’s public Web site at http://

www.nrc.gov/site-help/e-submittals/

getting-started.html. System 

requirements for accessing the ESubmittal server are detailed in the 

NRC’s ‘‘Guidance for Electronic 

Submission,’’ which is available on the 

agency’s public Web site at http://

www.nrc.gov/site-help/esubmittals.html. Participants may 

attempt to use other software not listed 

on the Web site, but should note that the 

NRC’s E-Filing system does not support 

unlisted software, and the NRC Meta 

System Help Desk will not be able to 

offer assistance in using unlisted 

software. 

If a participant is electronically 

submitting a document to the NRC in 

accordance with the E-Filing rule, the 

participant must file the document 

using the NRC’s online, Web-based 

submission form. In order to serve 

documents through the Electronic 

Information Exchange System, users 

will be required to install a Web 

browser plug-in from the NRC’s Web 

site. Further information on the Webbased submission form, including the 

installation of the Web browser plug-in, 

is available on the NRC’s public Web 

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

Once a participant has obtained a 

digital ID certificate and a docket has 

been created, the participant can then 

submit a request for hearing or petition 

for leave to intervene. Submissions 

should be in Portable Document Format 

(PDF) in accordance with NRC guidance 

available on the NRC’s public Web site 

at http://www.nrc.gov/site-help/esubmittals.html. A filing is considered 

complete at the time the documents are 

submitted through the NRC’s E-Filing 

system. To be timely, an electronic 

filing must be submitted to the E-Filing 

system no later than 11:59 p.m. Eastern 

Time on the due date. Upon receipt of 

a transmission, the E-Filing system 

time-stamps the document and sends 

the submitter an email notice 

confirming receipt of the document. The 

E-Filing system also distributes an email 

notice that provides access to the 

document to the NRC’s Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically using 

the NRC’s adjudicatory E-Filing system 

may seek assistance by contacting the 

NRC Meta System Help Desk through 

the ‘‘Contact Us’’ link located on the 

NRC’s public Web site at http://

www.nrc.gov/site-help/esubmittals.html, by email to 

MSHD.Resource@nrc.gov, or by a tollfree 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 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11475 

or party to use E-Filing if the presiding 

officer subsequently determines that the 

reason for granting the exemption from 

use of E-Filing no longer exists. 

Documents submitted in adjudicatory 

proceedings will appear in the NRC’s 

electronic hearing docket which is 

available to the public at http://

ehd1.nrc.gov/ehd/, unless excluded 

pursuant to an order of the Commission, 

or the presiding officer. Participants are 

requested not to include personal 

privacy information, such as social 

security numbers, home addresses, or 

home phone numbers in their filings, 

unless an NRC regulation or other law 

requires submission of such 

information. However, a request to 

intervene will require including 

information on local residence in order 

to demonstrate a proximity assertion of 

interest in the proceeding. With respect 

to copyrighted works, except for limited 

excerpts that serve the purpose of the 

adjudicatory filings and would 

constitute a Fair Use application, 

participants are requested not to include 

copyrighted materials in their 

submission. 

Petitions for leave to intervene must 

be filed no later than 60 days from the 

date of publication of this notice. 

Requests for hearing, petitions for leave 

to intervene, and motions for leave to 

file new or amended contentions that 

are filed after the 60-day deadline will 

not be entertained absent a 

determination by the presiding officer 

that the filing demonstrates good cause 

by satisfying the three factors in 10 CFR 

2.309(c)(1)(i)–(iii). 

For further details with respect to 

these license amendment applications, 

see the application for amendment 

which is available for public inspection 

in ADAMS and at the NRC’s PDR. For 

additional direction on accessing 

information related to this document, 

see the ‘‘Obtaining Information and 

Submitting Comments’’ section of this 

document. 

Entergy 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 misoperation 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 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Joseph A. 

Aluise, Associate General Counsel— 

Nuclear, Entergy Services, Inc., 639 

Loyola Avenue, New Orleans, Louisiana 

70113. 

NRC 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 

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 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Joseph A. 

Aluise, Associate General Council— 

Nuclear, Entergy Services, Inc., 639 

Loyola Avenue, New Orleans, Louisiana 

70113. 

NRC Branch Chief: 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: 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11477 

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 Temperature—High 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 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: J. 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 licenseecontrolled program with 

implementation of Nuclear Energy 

Institute (NEI) 04–10 (Revision 1), 

‘‘Risk-Informed Technical 

Specifications Initiative 5b, RiskInformed 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, RiskInformed Decisionmaking: Technical 

Specifications’’ (ADAMS Accession No. 

ML003740176). The licensee stated that 

the changes are consistent with NRCapproved Technical Specification Task 

Force (TSTF) Improved Standard 

Technical Specifications Change 

Traveler TSTF–425, ‘‘Relocate 

Surveillance Frequencies to Licensee 

Control—RITSTF [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 

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 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: William S. 

Blair, Managing Attorney—Nuclear, 

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 publiclyavailable 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] 

Subsystems—Tavg [average temperature] 

Greater Than or Equal to 350 °F [degrees 

Fahrenheit],’’ to correct nonconservative 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 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: William S. 

Blair, Managing Attorney—Nuclear, 

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 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11479 

(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 NonTechnical 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 1—The 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 2—The 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 3—The 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 licensee’s 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 licenseecontrolled 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. 

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 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: 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 hours. This will ensure that Tavg

[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 

operator’s 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. 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11481 

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] Deenergized’ 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. 

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 Loops—Mode 3,’’ 

Conditions C.2 and D.1, from ‘‘De-energize 

all control rod drive mechanisms,’’ to ‘‘Place 

the Rod Control System in a condition 

incapable of rod withdrawal.’’ It also revises 

LCO 3.4.9, ‘‘Pressurizer,’’ Required Action 

A.1, from requiring 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 

Flux—High 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 Flux—High 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 Flux—High trip setpoints ensures 

that reactor power is not inadvertently 

increased. Reducing the Power Range 

Neutron Flux—High trip setpoints is not an 

initiator to any accident previously 

evaluated. The consequences of any accident 

previously evaluated with the Power Range 

Neutron Flux—High 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 Flux—High 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 Flux—High 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. 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11483 

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 

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. 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11485 

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 

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. 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11487 

20: TSTF–314–A, Revision 0, ‘‘Require Static 

and Transient FQ 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, FQ(Z). This 

change is a correction that ensures that the 

plant conditions are as assumed in the 

accident analysis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

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 Turbine—Driven 

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 

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 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Leigh D. Perry, 

SVP & General Counsel of Operations 

and Nuclear, Southern Nuclear 

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 hours. 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 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: 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 

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

A notice of consideration of issuance 

of amendment to facility operating 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11489 

license or combined license, as 

applicable, proposed no significant 

hazards consideration determination, 

and opportunity for a hearing in 

connection with these actions, was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

amendment, (2) the amendment, and (3) 

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items can be accessed as described in 

the ‘‘Obtaining Information and 

Submitting Comments’’ section of this 

document. 

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 publiclyavailable 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 staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 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 Commission’s 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 riskinformed, 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 publiclyavailable 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 staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 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 

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 publiclyavailable 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 staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 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 licensee’s submittal of the 

certifications, as required by 10 CFR 

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

Amendment No.: 262. A publiclyavailable 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 staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s 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 publiclyavailable 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 Commission’s 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 publiclyavailable 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 staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 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 

mstockstill on DSK4VPTVN1PROD with NOTICES

Federal Register / Vol. 80, No. 41 / Tuesday, March 3, 2015 / Notices 11491 

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 publiclyavailable 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 staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 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 publiclyavailable 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 Commission’s 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 poweroperated 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 1—203; Unit 

2—191. 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 staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated 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 PressureTemperature 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 staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in the SE 

dated February 2, 2015. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 23rd day 

of February 2015. 

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 

SUPPLEMENTARY INFORMATION section of 

this document. 

FOR FURTHER INFORMATION CONTACT: 

Shirley Rohrer, Office of Nuclear 

Reactor Regulation, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001; telephone: 301–415– 

5411; email: Shirley.Rohrer@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

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 publiclyavailable 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. 

• NRC’s Agencywide Documents 

Access and Management System 

(ADAMS): You may obtain publiclyavailable documents online in the 

ADAMS Public Documents collection at 

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

adams.html. To begin the search, select 

‘‘ADAMS Public Documents’’ and then 

select ‘‘Begin Web-based ADAMS 

Search.’’ For problems with ADAMS, 

please contact the NRC’s Public 

Document Room (PDR) reference staff at 

1–800–397–4209, 301–415–4737, or by 

email to pdr.resource@nrc.gov. The 

ADAMS accession number for each 

document referenced (if it is available in 

ADAMS) is provided the first time that 

it is mentioned in the SUPPLEMENTARY

INFORMATION section. 

• NRC’s PDR: You may examine and 

purchase copies of public documents at 

the NRC’s PDR, Room O1–F21, One 

White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852. 

B. Submitting Comments 

Please include Docket ID NRC–2015– 

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 Commission’s 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 

mstockstill on DSK4VPTVN1PROD with NOTICES