5798 Federal Register / Vol. 80, No. 22 / Tuesday, February 3, 2015 / Notices 

Dated at Rockville, Maryland, this 26th day 

of January 2015. 

For the Nuclear Regulatory Commission. 

Michele G. Evans, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2015–02067 Filed 2–2–15; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2015–0015] 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses and Combined Licenses 

Involving No Significant Hazards 

Considerations 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Biweekly notice. 

SUMMARY: Pursuant to Section 189a. (2) 

of the Atomic Energy Act of 1954, as 

amended (the Act), the U.S. Nuclear 

Regulatory Commission (NRC) is 

publishing this regular biweekly notice. 

The Act requires the Commission to 

publish notice of any amendments 

issued, or proposed to be issued, and 

grants the Commission the authority to 

issue and make immediately effective, 

any amendment to an operating license 

or combined license, as applicable, 

upon a determination by the 

Commission that such amendment 

involves no significant hazards 

consideration, notwithstanding the 

pendency before the Commission of a 

request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued, from January 8, 

2015, to January 21, 2015. The last 

biweekly notice was published on 

January 20, 2015. 

DATES: Comments must be filed by 

March 5, 2015. A request for a hearing 

must be filed by April 6, 2015. 

ADDRESSES: You may submit comments 

by any of the following methods (unless 

this document describes a different 

method for submitting comments on a 

specific subject): 

• Federal Rulemaking Web site: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2015–0015. Address 

questions about NRC dockets to Carol 

Gallagher; telephone: 301–287–3422; 

email: Carol.Gallagher@nrc.gov. 

• Mail comments to: Cindy Bladey, 

Office of Administration, Mail Stop: 

3WFN–06–A44M, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001. 

For additional direction on obtaining 

information and submitting comments, 

see ‘‘Obtaining Information and 

Submitting Comments’’ in the 

SUPPLEMENTARY INFORMATION section of 

this document. 

FOR FURTHER INFORMATION CONTACT: 

Beverly A. Clayton, Office of Nuclear 

Reactor Regulation, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001; telephone: 301–415– 

3475, email: Beverly.Clayton@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and 

Submitting Comments 

A. Obtaining Information 

Please refer to Docket ID NRC–2015– 

0015 when contacting the NRC about 

the availability of information for this 

action. You may obtain 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–2015–0015. 

• 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– 

0015 in the subject line of your 

comment submission, in order to ensure 

that the NRC is able to make your 

comment submission available to the 

public in this docket. 

The NRC cautions you not to include 

identifying or contact information that 

you do not want to be publicly 

disclosed in your comment submission. 

The NRC posts all comment 

submissions at 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. 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 


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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 

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 

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 WebVerDate Sep<11>2014 14:46 Feb 02, 2015 Jkt 235001 PO 00000 Frm 00072 Fmt 4703 Sfmt 4703 E:\FR\FM\03FEN1.SGM 03FEN1

rljohnson on DSK3VPTVN1PROD with NOTICES
5800 Federal Register / Vol. 80, No. 22 / Tuesday, February 3, 2015 / Notices 

based submission form, including the 

installation of the Web browser plug-in, 

is available on the NRC’s public Web 

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

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. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station 

(Columbia), Benton County, Washington 

Date of amendment request: 

November 17, 2014. A publiclyavailable version is in ADAMS under 

Accession No. ML14336A100. 

Description of amendment request: 

The proposed amendment would 

modify the Technical Specifications to 

revise values for the safety limit 

minimum critical power ratio (SLMCPR) 

due to core loading fuel management 

changes for the upcoming Columbia 

operating cycle. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The basis of the Safety Limit Minimum 

Critical Power Ratio (SLMCPR) is to ensure 

no mechanistic fuel damage is calculated to 

occur if the limit is not violated. The new 

SLMCPR values preserve the existing margin 

to transition boiling. The derivation of the 

revised SLMCPR for Columbia, for 

incorporation into the Technical 

Specifications and its use to determine plant 

and cycle-specific thermal limits, has been 

performed using NRC approved methods. 

The revised SLMCPR values do not change 

the method of operating the plant and have 

no effect on the probability of an accident 

initiating event or transient. 

Based on the above, Energy Northwest has 

concluded that the proposed change will not 

result in a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

analyzed? 

Response: No. 

The proposed changes result only from a 

specific analysis for the Columbia core reload 

design. These changes do not involve any 

new or different methods for operating the 

facility. No new initiating events or 

transients result from these changes. 

Based on the above, Energy Northwest has 

concluded that the proposed change will not 

create the possibility of a new or different 

kind of accident from those previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The new SLMCPR is calculated using NRC 

approved methods with plant and cycle 

specific parameters for the current core 

design. The SLMCPR value remains 

conservative enough to ensure that at least 

99.9% of all fuel rods in the core will avoid 

transition boiling if the limit is not violated, 


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thereby preserving the fuel cladding 

integrity. The operating limit minimum 

critical power ratio (MCPR) is established to 

ensure that no fuel damage results during 

anticipated operational occurrences (AOOs). 

Accordingly, the margin of safety is 

maintained with the revised values. 

As a result, Energy Northwest has 

determined that the proposed change will not 

result in a significant reduction in a margin 

of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: William A. 

Horin, Esq., Winston & Strawn, 1700 K 

Street NW., Washington, DC 20006– 

3817. 

Acting NRC Branch Chief: Eric R. 

Oesterle. 

Florida Power & Light Company, et al., 

Docket Nos. 50–335 and 50–389, St. 

Lucie Plant, Units 1 and 2, St. Lucie 

County, Florida 

Date of amendment request: 

December 5, 2014. A publicly-available 

version is in ADAMS under Accession 

No. ML14351A074. 

Description of amendment request: 

The amendment would revise Technical 

Specifications (TSs) Section 3.6.2.1, 

regarding containment spray and 

cooling systems, by eliminating second 

completion times limiting time from 

discovery of failure to meet a limiting 

condition for operation (LCO). The 

proposed revision is consistent with 

NRC-approved Technical Specifications 

Task Force (TSTF) Traveler TSTF–439, 

Revision 2, ‘‘Eliminate Second 

Completion Times Limiting Time from 

Discovery of Failure to Meet an LCO’’ 

(Adams Accession No. ML051860296). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of any accident previously 

evaluated? 

Response: No. 

The proposed change that incorporated 

TSTF–439, Revision 2, [will eliminate] 

certain Completion Times from the TS. 

Completion Times are not an initiator to any 

accident previously evaluated. As a result, 

the probability of an accident previously 

evaluated is not affected. The consequences 

of an accident during the revised Completion 

Times are no different [from] the 

consequences of the same accident during 

the existing Completion Times. As a result, 

the consequences of an accident previously 

evaluated are not affected by this change. The 

proposed change does not alter or prevent the 

ability of structures, systems, or components 

(SSCs) from performing their intended 

function to mitigate the consequences of an 

initiating event within the assumed 

acceptance limits. 

The proposed change does not affect the 

source term, containment isolation, or 

radiological release assumptions used in 

evaluating the radiological consequences of 

an accident previously evaluated. Further, 

the proposed change does not increase the 

types or amounts of radioactive effluent that 

may be released offsite, nor significantly 

increase individual or cumulative 

occupational/public radiation exposures. The 

proposed change is consistent with the 

[previous] safety analysis assumptions and 

resultant consequences. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

The proposed change does not [involve] a 

physical alteration of the plant (i.e., no new 

or different type of equipment will be 

installed) or a change in the methods 

governing normal plant operation. The 

proposed change does not alter any 

assumptions made in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in the margin of safety? 

Response: No. 

The proposed change to delete the second 

Completion Times does not alter the manner 

in which safety limits, limiting safety system 

settings, or limiting conditions for operation 

are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed change will not result 

in plant operation in a configuration outside 

the design basis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and determined that 

the three standards of 10 CFR 50.92(c) 

are satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: William S. 

Blair, Managing Attorney—Nuclear, 

Florida Power & Light Company, 700 

Universe Blvd., MS LAW/JB, Juno 

Beach, FL 33408–0420. 

NRC Branch Chief: Shana R. Helton. 

Omaha Public Power District (OPPD), 

Docket No. 50–285, Fort Calhoun 

Station, Unit 1, Washington County, 

Nebraska 

Date of amendment request: 

December 26, 2014. A publicly-available 

version is in ADAMS under Accession 

No. ML14365A123. 

Description of amendment request: 

The proposed amendment upgrades the 

Emergency Action Level (EAL) scheme 

by adopting NRC-endorsed Nuclear 

Energy Institute (NEI) 99–01, Revision 6, 

‘‘Methodology for the Development of 

Emergency Action Levels for NonPassive Reactors,’’ issued January 2011 

(ADAMS Accession No. ML110240324). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes to OPPD’s EAL 

scheme to adopt the NRC-endorsed guidance 

in NEI 99–01, Revision 6, ‘‘Development of 

Emergency Action Levels for Non-Passive 

Reactors,’’ do not reduce the capability to 

meet the emergency planning requirements 

established in 10 CFR 50.47 and 10 CFR 50, 

Appendix E. The proposed changes do not 

reduce the functionality, performance, or 

capability of OPPD’s ERO [emergency 

response organization] to respond in 

mitigating the consequences of any design 

basis accident. 

The probability of a reactor accident 

requiring implementation of Emergency Plan 

EALs has no relevance in determining 

whether the proposed changes to the EALs 

reduce the effectiveness of the Emergency 

Plans. As discussed in Section D, ‘‘Planning 

Basis,’’ of NUREG–0654, Revision 1, ‘‘Criteria 

for Preparation and Evaluation of 

Radiological Emergency Response Plans and 

Preparedness in Support of Nuclear Power 

Plants’’ [issued November 1980; ADAMS 

Accession No. ML040420012]: 

. . . The overall objective of emergency 

response plans is to provide dose savings 

(and in some cases immediate life saving) for 

a spectrum of accidents that could produce 

offsite doses in excess of Protective Action 

Guides (PAGs). No single specific accident 

sequence should be isolated as the one for 

which to plan because each accident could 

have different consequences, both in nature 

and degree. Further, the range of possible 

selection for a planning basis is very large, 

starting with a zero point of requiring no 

planning at all because significant offsite 

radiological accident consequences are 

unlikely to occur, to planning for the worst 

possible accident, regardless of its extremely 

low likelihood . . . 

Therefore, OPPD did not consider the risk 

insights regarding any specific accident 


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initiation or progression in evaluating the 

proposed changes. 

The proposed changes do not involve any 

physical changes to plant equipment or 

systems, nor do they alter the assumptions of 

any accident analyses. The proposed changes 

do not adversely affect accident initiators or 

precursors nor do they alter the design 

assumptions, conditions, and configuration 

or the manner in which the plant is operated 

and maintained. The proposed changes do 

not adversely affect the ability of Structures, 

Systems, or Components (SSCs) to perform 

their intended safety functions in mitigating 

the consequences of an initiating event 

within the assumed acceptance limits. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes to OPPD’s EAL 

scheme to adopt the NRC-endorsed guidance 

in NEI 99–01, Revision 6, do not involve any 

physical changes to plant systems or 

equipment. The proposed changes do not 

involve the addition of any new plant 

equipment. The proposed changes will not 

alter the design configuration, or method of 

operation of plant to be performed as 

required. The proposed changes do not create 

any new credible failure mechanisms, 

malfunctions, or accident initiators. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from those that have been 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes to OPPD’s EAL 

scheme to adopt the NRC-endorsed guidance 

in NEI 99–01, Revision 6, do not alter or 

exceed a design basis or safety limit. There 

is no change being made to safety analysis 

assumptions, safety limits, or limiting safety 

system settings that would adversely affect 

plant safety as a result of the proposed 

change. There are no changes to setpoints or 

environmental conditions of any SSC or the 

manner in which any SSC is operated. 

Margins of safety are unaffected by the 

proposed changes to adopt the NEI 99–01, 

Revision 6, EAL scheme guidance. The 

applicable requirements of 10 CFR 50.47 and 

10 CFR 50, Appendix E will continue to be 

met. 

Therefore, the proposed changes do not 

involve any reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David A. Repka, 

Esq., Winston & Strawn, 1700 K Street 

NW., Washington, DC 20006–3817. 

Acting NRC Branch Chief: Eric R. 

Oesterle. 

South Carolina Electric and Gas 

Company Docket Nos.: 52–027 and 52– 

028, Virgil C. Summer Nuclear Station, 

Units 2 and 3, Fairfield County, South 

Carolina 

Date of amendment request: July 17, 

2014. A publicly-available version is in 

ADAMS under Accession No. 

ML14202A088. 

Description of amendment request: 

The proposed changes would revise the 

Combined Licenses (COLs) by (1) 

providing additional detail to describe 

the mechanical connection between the 

internal containment structural module 

steel faceplates and the base concrete, 

(2) allowing for increases in the 

thickness of the structural wall module 

faceplates, (3) identifying changes to the 

wall thicknesses for portions of some 

internal containment structural wall 

modules, and (4) identifying the use of 

steel plates, structural shapes, 

reinforcement bars, or tie bars between 

the faceplates of the structural wall 

modules, where needed to meet 

applicable code requirements. 

Because this proposed change 

requires a departure from Tier 1 

information in the Westinghouse 

Advanced Passive 1000 Design Control 

Document (DCD), the licensee also 

requested an exemption from the 

requirements of the Generic DCD Tier 1 

in accordance with 10 CFR 52.63(b)(1). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The design function of the internal 

containment structures is to provide support, 

protection, and separation for the seismic 

Category I mechanical and electrical 

equipment located in those structures. These 

structures are structurally designed to meet 

seismic Category I requirements as defined in 

Regulatory Guide 1.29. 

The changes to the design details for the 

structural modules do not have an adverse 

impact on the response of the nuclear island 

structures to safe shutdown earthquake 

ground motions or loads due to anticipated 

transients or postulated accident conditions, 

nor do they change the seismic Category I 

classification. Evaluations have been 

performed which determined that the 

proposed changes do not have a significant 

impact on the calculated loads for the 

affected structural modules, or critical 

locations, and no significant impact on the 

global seismic model. The changes to the 

design details for the structural modules do 

not impact the support, design, or operation 

of mechanical and fluid systems. There is no 

change to plant systems or the response of 

systems to postulated accident conditions. 

There is no change to the predicted 

radioactive releases due to postulated 

accident conditions. The plant response to 

previously evaluated accidents or external 

events is not adversely affected, nor does the 

change described create any new accident 

precursors. 

Therefore, the proposed amendment does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes are to revise design 

details for the internal containment structural 

modules. The changes do not change the 

design requirements of the nuclear island 

structures, nor do they change the seismic 

Category I classification. The changes to the 

design details for the internal containment 

structural modules do not change the design 

function, support, design, or operation of 

mechanical and fluid systems. The changes 

to the design details for the internal 

containment structural modules do not result 

in a new failure mechanism for the nuclear 

island structures or introduce any new 

accident precursors. As a result, the design 

function of the nuclear island structures is 

not adversely affected by the proposed 

change. 

Therefore, the proposed amendment does 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The requested amendment proposes 

changes to the structural details associated 

with the in-containment structural modules. 

The purpose of these changes is to ensure 

that the requirements contained in the 

applicable construction codes are met. As 

discussed in UFSAR [Updated Final Analysis 

Report], Section 3.8.3.5, ‘‘Design Procedures 

and Acceptance Criteria,’’ the in-containment 

structural modules are designed in 

accordance with ACI [American Concrete 

Institute] 349 and AISC [American Institute 

of Steel Construction] N690. Thus, the 

identification of additional structural module 

connection details, the increase in structural 

module faceplate and wall thicknesses, and 

the addition of additional reinforcement in 

specific areas are proposed to ensure that the 

codes of record, and the associated margins 

contained therein, continue to be met as 

specified in the design basis. Structural and 

seismic analysis of the modified sections in 

accordance with the methodologies 

identified in the UFSAR has confirmed that 

the applicable requirements of ACI 349 and 

AISC N690 continue to be met for affected incontainment structural modules. 

As a result, the proposed changes do not 

adversely affect any safety related equipment 


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or other design functions, design code 

compliance, design analysis, safety analysis 

input or result, or design/safety margin. No 

safety analysis or design basis acceptance 

limit/criterion is challenged or exceeded by 

the proposed changes. 

Therefore, the requested amendment does 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ms. Kathryn M. 

Sutton, Morgan, Lewis & Bockius LLC, 

1111 Pennsylvania Avenue NW., 

Washington, DC 20004–2514. 

NRC Branch Chief: Lawrence J. 

Burkhart. 

South Carolina Electric & Gas Company, 

South Carolina Public Service 

Authority, Docket No. 50–395, Virgil C. 

Summer Nuclear Station, Unit 1, 

Fairfield County, South Carolina 

Date of amendment request: 

December 19, 2014. A publicly-available 

version is in ADAMS Package Accession 

No. ML14363A422. 

Description of amendment request: 

The licensee proposes to expand the 

emergency planning zone (EPZ) 

boundary, to revise the evacuation time 

estimates (ETA) analysis, and revise the 

alert and notification system (ANS) 

design reports to encompass the 

expanded EPZ boundary. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes, which include 

expansion of the EPZ boundary and revision 

of the ETE analysis and ANS design reports 

to encompass the expanded EPZ boundary, 

do not impact the physical function of plant 

structures, systems, or components (SSC) or 

the manner in which SSCs perform their 

design function. The proposed changes 

neither adversely affect accident initiators or 

precursors, nor alter design assumptions. The 

proposed changes do not alter or prevent the 

ability of SSCs to perform their intended 

function to mitigate the consequences of an 

initiating event within assumed acceptance 

limits. No operating procedures or 

administrative controls that function to 

prevent or mitigate accidents are affected by 

the proposed changes. Therefore, the 

proposed changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not involve a 

physical alteration of the plant (i.e., no new 

or different type of equipment will be 

installed or removed) or a change in the 

method of plant operation. The proposed 

changes will not introduce failure modes that 

could result in a new accident, and the 

change does not alter assumptions made in 

the safety analysis. The proposed changes, 

which include expansion of the EPZ 

boundary and revision of the ETE analysis 

and ANS design reports to encompass the 

expanded EPZ boundary, are not initiators of 

any accidents. Therefore, the proposed 

changes do not create the possibility of a new 

or different kind of accident from any 

accident previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Margin of safety is associated with the 

ability of the fission product barriers (i.e., 

fuel cladding, reactor coolant system 

pressure boundary, and containment 

structure) to limit the level of radiation dose 

to the public. The proposed changes, which 

include expansion of the EPZ boundary and 

revision of the ETE analysis and ANS design 

reports to encompass the expanded EPZ 

boundary, do not impact operation of the 

plant or its response to transients or 

accidents. The proposed changes do not alter 

requirements of the Technical Specifications 

or the Unit 1 Operating License. The 

proposed changes do not involve a change in 

the method of plant operation and no 

accident analyses will be affected by the 

proposed changes. 

Additionally, the proposed changes will 

not relax any criteria used to establish safety 

limits and will not relax any safety system 

settings. The safety analysis acceptance 

criteria are not affected by these proposed 

changes. The proposed changes will not 

result in plant operation in a configuration 

outside the design basis. The proposed 

changes do not adversely affect systems that 

respond to safely shut down the plant and to 

maintain the plant in a safe shutdown 

condition. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: J. Hagood 

Hamilton, Jr., South Carolina Electric & 

Gas Company, Post Office Box 764, 

Columbia, SC 29218. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Southern Nuclear Operating Company, 

Inc. Docket Nos. 52–025 and 52–026, 

Vogtle Electric Generating Plant (VEGP), 

Units 3 and 4, Burke County, Georgia 

Date of amendment request: January 

8, 2015. A publicly-available version is 

in ADAMS under Accession No. 

ML15008A466. 

Description of amendment request: 

The proposed change would amend 

Combined License Nos. NPF–91 and 

NPF–92 for the VEGP, Units 3 and 4 by 

departing from the plant-specific Design 

Control Document (DCD) Tier 1 (and 

corresponding Combined License 

Appendix C information) and Tier 2 

material by making changes to specify 

the use of latching control relays in lieu 

of breakers to de-energize the control 

rod drive mechanism (CRDM) motor 

generator (MG) set generator field on a 

diverse actuation system (DAS) signal. 

Because this proposed change 

requires a departure from Tier 1 

information in the Westinghouse 

Advanced Passive 1000 DCD, the 

licensee also requested an exemption 

from the requirements of the Generic 

DCD Tier 1 in accordance with 10 CFR 

52.63(b)(1). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change to use field control 

relays in lieu of field circuit breakers to deenergize the CRDM MG Set excitation field 

does not result in a change to the basic MG 

Set design function, which is to supply 

reliable electrical power to the CRDMs while 

providing a trip function on a DAS signal, 

allowing the control rods to drop. The 

Probabilistic Risk Assessment (PRA) is not 

adversely affected. No safety-related 

structure, system, or component (SSC) or 

function is adversely affected. The change 

does not involve nor interface with any SSC 

accident initiator or initiating sequence of 

events, and thus, the probabilities of the 

accidents evaluated in the UFSAR are not 

affected. Because the change maintains the 

CRDM MG set trip function used to mitigate 

an accident, the consequences of the 

accidents evaluated in the UFSAR are not 

affected. 

Therefore, there is no significant increase 

in the probability or consequences of an 

accident previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 


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There is no safety-related SSC or function 

adversely affected by this proposed change to 

use control relays instead of breakers to deenergize the CRDM MG set generator field on 

demand. This proposed change does not 

change any equipment qualification or 

fission product barrier. The change does not 

result in a new failure mode, malfunction or 

sequence of events that could affect safety or 

safety-related equipment. This activity will 

not allow for a new fission product release 

path, result in a new fission product barrier 

failure mode, or create a new sequence of 

events that would result in significant fuel 

cladding failures. 

Therefore, this activity does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

There is no safety-related SSC or function 

adversely affected by this proposed change to 

use relays instead of breakers to control the 

CRDM MG set generator field. The function 

to trip the MG set generator field on a DAS 

signal, allowing the control rods to drop, is 

not adversely affected by the use of relays as 

the device to de-energize the generator field. 

The proposed change does not affect any 

safety-related design code, function, design 

analysis, safety analysis input or result, or 

design/safety margin. No safety analysis or 

design basis acceptance limit/criterion is 

challenged or exceeded by the requested 

change, thus, no margin of safety is reduced. 

Therefore, the proposed amendment does 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. M. Stanford 

Blanton, Balch & Bingham LLP, 1710 

Sixth Avenue North, Birmingham, AL 

35203–2015. 

NRC Branch Chief: Lawrence 

Burkhart. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–348 and 50–364, 

Joseph M. Farley Nuclear Plant, Units 1 

and 2, Houston County, Alabama 

Date of amendment request: 

November 24, 2014. A publiclyavailable version is in ADAMS under 

Accession Package No. ML14335A689. 

Description of amendment request: 

The licensee requested 24 revisions to 

the Technical Specifications. Twenty 

two revisions adopt various previously 

NRC approved Technical Specifications 

Task Force Travelers and two revisions 

are not associated with Travelers. A list 

of the requested revisions is included in 

Enclosure 1 of the application. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration for each of the 24 changes 

requested, which is presented below: 

Request No. 1: TSTF–27–A, Revision 3, 

‘‘Revise SR Frequency for Minimumn 

Temperature for Criticality’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the 

Surveillance Frequency for monitoring RCS 

temperature to ensure the minimum 

temperature for criticality is met. The 

Frequency is changed from a 30 minute 

Frequency when certain conditions are met 

to a periodic Frequency that it is controlled 

in accordance with the Surveillance 

Frequency Control Program. The 

measurement of RCS [reactor coolant system] 

temperature is not an initiator of any 

accident previously evaluated. The minimum 

RCS temperature for criticality is not 

changed. As a result, the mitigation of any 

accident previously evaluated is not affected. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change revises the 

Surveillance Frequency for monitoring RCS 

temperature to ensure the minimum 

temperature for criticality is met. The 

current, condition based Frequency 

represents a distraction to the control room 

operator during the critical period of plant 

startup. RCS temperature is closely 

monitored by the operator during the 

approach to criticality and temperature is 

recorded on charts and computer logs. 

Allowing the operator to monitor 

temperature as needed by the situation and 

logging RCS temperature at a periodic 

Frequency that it is controlled in accordance 

with the Surveillance Frequency Control 

Program is sufficient to ensure that the LCO 

[limiting condition for operation] is met 

while eliminating a diversion of the 

operator’s attention. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 2: TSTF–46–A, Revision 1, 

‘‘Clarify the CIV Surveillance to Apply Only 

to Automatic Isolation Valves’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the 

requirements in Technical Specification SR 

3.6.3.4, and the associated Bases, to delete 

the reference to verifying the isolation time 

of ‘‘each power operated’’ containment 

isolation valve (CIV) and only require 

verification of each ‘‘automatic power 

operated containment isolation valve.’’ The 

closure times for CIVs that do not receive an 

automatic closure signal are not an initiator 

of any design basis accident or event, and 

therefore the proposed change does not 

increase the probability of any accident 

previously evaluated. The CIVs are used to 

respond to accidents previously evaluated. 

Power operated CIVs that do not receive an 

automatic closure signal are not assumed to 

close in a specified time. The proposed 

change does not change how the plant would 

mitigate an accident previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not result in a 

change in the manner in which the CIVs 

provide plant protection or introduce any 

new or different operational conditions. 

Periodic verification that the closure times 

for CIVs that receive an automatic closure 

signal are within the limits established by the 

accident analysis will continue to be 

performed under SR 3.6.3.4. The change does 

not alter assumptions made in the safety 

analysis, and is consistent with the safety 

analysis assumptions and current plant 

operating practice. There are also no design 

changes associated with the proposed 

changes, and the change does not involve a 

physical alteration of the plant (i.e., no new 

or different type of equipment will be 

installed). 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change provides clarification 

that only CIVs that receive an automatic 


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isolation signal are within the scope of the 

SR 3.6.3.4. The proposed change does not 

result in a change in the manner in which the 

CIVs provide plant protection. Periodic 

verification that closure times for CIVs that 

receive an automatic isolation signal are 

within the limits established by the accident 

analysis will continue to be performed. The 

proposed change does not affect the safety 

analysis acceptance criteria for any analyzed 

event, nor is there a change to any Safety 

Analysis Limit. The proposed change does 

not alter the manner in which safety limits, 

limiting safety system settings or limiting 

conditions for operation are determined, nor 

is there any adverse effect on those plant 

systems necessary to assure the 

accomplishment of protection functions. The 

proposed change will not result in plant 

operation in a configuration outside the 

design basis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 3: TSTF–87–A, Revision 2, 

‘‘Revise ‘‘RTBs Open’’ and ‘‘CRDM 

Deenergized’’ Actions to ‘‘Incapable of Rod 

Withdrawal’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This change revises the Required Actions 

for LCO 3.4.5, ‘‘RCS Loops—Mode 3,’’ 

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

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

the Rod Control System in a condition 

incapable of rod withdrawal.’’ It also revises 

LCO 3.4.9, ‘‘Pressurizer,’’ Required Action A. 

1, from requiring the Reactor Trip Breakers 

to be open after reaching MODE 3 to ‘‘Place 

the Rod Control System in a condition 

incapable of rod withdrawal,’’ and to require 

full insertion of all rods. Inadvertent rod 

withdrawal can be an initiator for design 

basis accidents or events during certain plant 

conditions, and therefore must be prevented 

under those conditions. The proposed 

Required Actions for LCO 3.4.5 and LCO 

3.4.9 satisfy the same intent as the current 

Required Actions, which is to prevent 

inadvertent rod withdrawal when an 

applicable Condition is not met, and is 

consistent with the assumptions of the 

accident analysis. As a result, the proposed 

change does not increase the probability of 

any accident previously evaluated. The 

proposed change does not change how the 

plant would mitigate an accident previously 

evaluated as in both the current and 

proposed requirements, rod withdrawal is 

prohibited. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change provides less 

specific, but equivalent, direction on the 

manner in which inadvertent control rod 

withdrawal is to be prevented when the 

Conditions of LCO 3.4.5 and LCO 3.4.9 are 

not met. Rod withdrawal will continue to be 

prevented when the applicable Conditions of 

LCO 3.4.5 and LCO 3.4.9 are met. There are 

no design changes associated with the 

proposed changes, and the change does not 

involve a physical alteration of the plant (i.e., 

no new or different type of equipment will 

be installed). The change does not alter 

assumptions made in the safety analysis, and 

is consistent with the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change provides the 

operational flexibility of allowing alternate, 

but equivalent, methods of preventing rod 

withdrawal when LCO 3.4.5 and LCO 3.4.9 

are not met. The proposed change does not 

affect the safety analysis acceptance criteria 

for any analyzed event, nor is there a change 

to any safety analysis limit. The proposed 

change does not alter the manner in which 

safety limits, limiting safety system settings 

or limiting conditions for operation are 

determined, nor is there any adverse effect on 

those plant systems necessary to assure the 

accomplishment of protection functions. The 

proposed change will not result in plant 

operation in a configuration outside the 

design basis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 4: TSTF–245–A, Revision 1, 

‘‘AFW Train Operable When in Service’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the 

requirements in Technical Specification 

3.7.5, ‘‘Auxiliary Feedwater (AFW) System,’’ 

to clarify the operability of an AFW train 

when it is aligned for manual steam generator 

level control. The AFW System is not an 

initiator of any design basis accident or 

event, and therefore the proposed change 

does not increase the probability of any 

accident previously evaluated. The AFW 

System is used to respond to accidents 

previously evaluated. The proposed change 

does not affect the design of the AFW 

System, and no physical changes are made to 

the plant. The proposed change does not 

significantly change how the plant would 

mitigate an accident previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not result in a 

change in the manner in which the AFW 

System provides plant protection. The AFW 

System will continue to supply water to the 

steam generators to remove decay heat and 

other residual heat by delivering at least the 

minimum required flow rate to the steam 

generators. There are no design changes 

associated with the proposed changes, and 

the change does not involve a physical 

alteration of the plant (i.e., no new or 

different type of equipment will be installed). 

The change does not alter assumptions made 

in the safety analysis, and is consistent with 

the safety analysis assumptions and current 

plant operating practice. Manual control of 

AFW level control valves is not an accident 

initiator. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change provides the 

operational flexibility of allowing an AFW 

train(s) to be considered operable when it is 

not in the normal standby alignment and is 

temporarily incapable of automatic initiation, 

such as during alignment and operation for 

manual steam generator level control, 

provided it is capable of being manually 

realigned to the AFW heat removal mode of 

operation. The proposed change does not 

result in a change in the manner in which the 

AFW System provides plant protection. The 

AFW System will continue to supply water 

to the steam generators to remove decay heat 

and other residual heat by delivering at least 

the minimum required flow rate to the steam 

generators. The proposed change does not 

affect the safety analysis acceptance criteria 

for any analyzed event, nor is there a change 

to any Safety Analysis Limit. The proposed 

change does not alter the manner in which 

safety limits, limiting safety system settings 

or limiting conditions for operation are 

determined, nor is there any adverse effect on 

those plant systems necessary to assure the 

accomplishment of protection functions. The 

proposed change will not result in plant 

operation in a configuration outside the 

design basis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 


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satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 5: TSTF–247–A, Revision 0, 

‘‘Provide Separate Condition Entry for Each 

PORV and Block Valve’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the 

requirements in Technical Specification 

3.4.11, ‘‘Pressurizer PORVs [power operated 

relief valves],’’ to clarify that separate 

Condition entry is allowed for each block 

valve. Additionally, the Actions are modified 

to no longer require that the PORVs be placed 

in manual operation when both block valves 

are inoperable and cannot be restored to 

operable status within the specified 

Completion Time. This preserves the 

overpressure protection capabilities of the 

PORVs. The pressurizer block valves are used 

to isolate their respective PORV in the event 

it is experiencing excessive leakage, and are 

not an initiator of any design basis accident 

or event. Therefore the proposed change does 

not increase the probability of any accident 

previously evaluated. The PORV and block 

valves are used to respond to accidents 

previously evaluated. The proposed change 

does not affect the design of the PORV and 

block valves, and no physical changes are 

made to the plant. The proposed change does 

not change how the plant would mitigate an 

accident previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not result in a 

change in the manner in which the PORV 

and block valves provide plant protection. 

The PORVs will continue to provide 

overpressure protection, and the block valves 

will continue to provide isolation capability 

in the event a PORV is experiencing 

excessive leakage. There are no design 

changes associated with the proposed 

changes, and the change does not involve a 

physical alteration of the plant (i.e., no new 

or different type of equipment will be 

installed). The change does not alter 

assumptions made in the safety analysis, and 

is consistent with the safety analysis 

assumptions and current plant operating 

practice. Operation of the PORV block valves 

is not an accident initiator. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes provide clarification 

that separate Condition entry is allowed for 

each block valve. Additionally, the Actions 

are modified to no longer require that the 

PORVs be placed in manual operation when 

both block valves are inoperable and cannot 

be restored to operable status within the 

specified Completion Time. This preserves 

the overpressure protection capabilities of 

the PORVs. The proposed change does not 

result in a change in the manner in which the 

PORV and block valves provide plant 

protection. The PORVs will continue to 

provide overpressure protection, and the 

block valves will continue to provide 

isolation capability in the event a PORV is 

experiencing excessive leakage. The 

proposed change does not affect the safety 

analysis acceptance criteria for any analyzed 

event, nor is there a change to any safety 

analysis limit. The proposed change does not 

alter the manner in which safety limits, 

limiting safety system settings or limiting 

conditions for operation are determined, nor 

is there any adverse effect on those plant 

systems necessary to assure the 

accomplishment of protection functions. The 

proposed change will not result in plant 

operation in a configuration outside the 

design basis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 6: TSTF–248–A, Revision 0, 

‘‘Revise Shutdown Margin Definition for 

Stuck Rod Exception’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change modifies the 

definition of Shutdown Margin to eliminate 

the requirement to assume the highest worth 

control rod is fully withdrawn when 

calculating Shutdown Margin if it can be 

verified by two independent means that all 

control rods are inserted. The method for 

calculating shutdown margin is not an 

initiator of any accident previously 

evaluated. If it can be verified by two 

independent means that all control rods are 

inserted, the calculated Shutdown Margin, 

without the conservatism of assuming the 

highest worth control rod is withdrawn, is 

accurate and consistent with the assumptions 

in the accident analysis. As a result, the 

mitigation of any accident previously 

evaluated is not affected. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change modifies the 

definition of Shutdown Margin to eliminate 

the requirement to assume the highest worth 

control rod is fully withdrawn when 

calculating Shutdown Margin if it can be 

verified by two independent means that all 

control rods are inserted. The additional 

margin of safety provided by the assumption 

that the highest worth control rod is fully 

withdrawn is unnecessary if it can be 

independently verified that all controls rods 

are inserted. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 7: TSTF–266–A, Revision 3, 

‘‘Eliminate the Remote Shutdown System 

Table of Instrumentation and Controls’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change removes the list of 

Remote Shutdown System instrumentation 

and controls from the Technical 

Specifications and places them in the Bases. 

The Technical Specifications continue to 

require that the instrumentation and controls 

be operable. The location of the list of 

Remote Shutdown System instrumentation 

and controls is not an initiator to any 

accident previously evaluated. The proposed 

change will have no effect on the mitigation 

of any accident previously evaluated because 

the instrumentation and controls continue to 

be required to be operable. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 


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kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change removes the list of 

Remote Shutdown System instrumentation 

and controls from the Technical 

Specifications and places it in the Bases. The 

review performed by the NRC when the list 

of Remote Shutdown System instrumentation 

and controls is revised will no longer be 

needed unless the criteria in 10 CFR 50.59 

are not met such that prior NRC review is 

required. The Technical Specification 

requirement that the Remote Shutdown 

System be operable, the definition of 

operability, the requirements of 10 CFR 

50.59, and the Technical Specifications Bases 

Control Program are sufficient to ensure that 

revision of the list without prior NRC review 

and approval does not introduce a significant 

safety risk. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 8: TSTF–272–A, Revision 1, 

‘‘Refueling Boron Concentration 

Clarification’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change modifies the 

Applicability of Specification 3.9.1, ‘‘Boron 

Concentration,’’ to clarify that the boron 

concentration limits are only applicable to 

the refueling canal and the refueling cavity 

when those volumes are attached to the 

Reactor Coolant System (RCS). The boron 

concentration of water volumes not 

connected to the RCS are not an initiator of 

an accident previously evaluated. The ability 

to mitigate any accident previously evaluated 

is not affected by the boron concentration of 

water volumes not connected to the RCS. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change modifies the 

Applicability of Specification 3.9.1, ‘‘Boron 

Concentration,’’ to clarify that the boron 

concentration limits are only applicable to 

the refueling canal and the refueling cavity 

when those volumes are attached to the RCS. 

Technical Specification SR 3.0.4 requires that 

Surveillances be met prior to entering the 

Applicability of a Specification. As a result, 

the boron concentration of the refueling 

cavity or the refueling canal must be verified 

to satisfy the LCO prior to connecting those 

volumes to the RCS. The margin of safety 

provided by the refueling boron 

concentration is not affected by this change 

as the RCS boron concentration will continue 

to satisfy the LCO. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 9: TSTF–273–A, Revision 2, 

‘‘Safety Function Determination Program 

Clarifications’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed TS changes add explanatory 

text to the programmatic description of the 

Safety Function Determination Program 

(SFDP) in Specification 5.5.15 to clarify in 

the requirements that consideration does not 

have to be made for a loss of power in 

determining loss of function. The Bases for 

LCO 3.0.6 is revised to provide clarification 

of the ‘‘appropriate LCO for loss of function,’’ 

and that consideration does not have to be 

made for a loss of power in determining loss 

of function. The changes are editorial and 

administrative in nature, and therefore do not 

increase the probability of any accident 

previously evaluated. No physical or 

operational changes are made to the plant. 

The proposed change does not change how 

the plant would mitigate an accident 

previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes are editorial and 

administrative in nature and do not result in 

a change in the manner in which the plant 

operates. The loss of function of any specific 

component will continue to be addressed in 

its specific TS LCO and plant configuration 

will be governed by the required actions of 

those LCOs. The proposed changes are 

clarifications that do not degrade the 

availability or capability of safety related 

equipment, and therefore do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. There are no design changes 

associated with the proposed changes, and 

the changes do not involve a physical 

alteration of the plant (i.e., no new or 

different type of equipment will be installed). 

The changes do not alter assumptions made 

in the safety analysis, and are consistent with 

the safety analysis assumptions and current 

plant operating practice. Due to the 

administrative nature of the changes, they 

cannot be an accident initiator. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes to TS 5.5.15 are 

clarifications and are editorial and 

administrative in nature. No changes are 

made the LCOs for plant equipment, the time 

required for the TS Required Actions to be 

completed, or the out of service time for the 

components involved. The proposed changes 

do not affect the safety analysis acceptance 

criteria for any analyzed event, nor is there 

a change to any safety analysis limit. The 

proposed changes do not alter the manner in 

which safety limits, limiting safety system 

settings or limiting conditions for operation 

are determined, nor is there any adverse 

effect on those plant systems necessary to 

assure the accomplishment of protection 

functions. The proposed changes will not 

result in plant operation in a configuration 

outside the design basis. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 10: TSTF–283–A, Revision 3, 

‘‘Modify Section 3.8 Mode Restriction Notes’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change modifies Mode 

restriction Notes on four diesel generator 

(DG) Surveillances to allow performance of 

the Surveillance in whole or in part to 

reestablish DG Operability. The emergency 

diesel generators and their associated 

emergency loads are accident mitigating 

features, and are not an initiator of any 

accident previously evaluated. As a result the 

probability of any accident previously 

evaluated is not increased. The proposed 

change allows Surveillance testing to be 

performed in whole or in part to reestablish 


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Operability of a DG. The consequences of an 

accident previously evaluated during the 

period that the DG is being tested to 

reestablish Operability are no different from 

the consequences of an accident previously 

evaluated while the DG is inoperable. As a 

result, the consequences of any accident 

previously evaluated are not increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The purpose of Surveillances is to verify 

that equipment is capable of performing it’s 

assumed safety function. The proposed 

change will only allow the performance of 

the Surveillances to reestablish Operability 

and the proposed changes may not be used 

to remove a DG from service. In addition, the 

proposed change will potentially shorten the 

time that a DG is unavailable because testing 

to reestablish Operability can be performed 

without a plant shutdown. The proposed 

changes also require an assessment to verify 

that plant safety will be maintained or 

enhanced by performance of the Surveillance 

in the normally prohibited Modes. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 11: TSTF–284–A, Revision 3, 

‘‘Add ‘Met vs. Perform’ to Technical 

Specification 14, Frequency’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes insert a discussion 

paragraph into Specification 1.4, and several 

new examples are added to facilitate the use 

and application of SR Notes that utilize the 

terms ‘‘met’’ and ‘‘perform’’. The changes 

also modify SRs in multiple Specifications to 

appropriately use ‘‘met’’ and ‘‘perform’’ 

exceptions. The changes are administrative 

in nature because they provide clarification 

and correction of existing expectations, and 

therefore the proposed change does not 

increase the probability of any accident 

previously evaluated. No physical or 

operational changes are made to the plant. 

The proposed change does not significantly 

change how the plant would mitigate an 

accident previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes are administrative 

in nature and do not result in a change in the 

manner in which the plant operates. The 

proposed changes provide clarification and 

correction of existing expectations that do 

not degrade the availability or capability of 

safety related equipment, and therefore do 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. There are no design 

changes associated with the proposed 

changes, and the changes do not involve a 

physical alteration of the plant (i.e., no new 

or different type of equipment will be 

installed). The changes do not alter 

assumptions made in the safety analysis, and 

are consistent with the safety analysis 

assumptions and current plant operating 

practice. Due to the administrative nature of 

the changes, they cannot be an accident 

initiator. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes are administrative 

in nature and do not result in a change in the 

manner in which the plant operates. The 

proposed changes provide clarification and 

correction of existing expectations that do 

not degrade the availability or capability of 

safety related equipment, or alter their 

operation. The proposed changes do not 

affect the safety analysis acceptance criteria 

for any analyzed event, nor is there a change 

to any safety analysis limit. The proposed 

changes do not alter the manner in which 

safety limits, limiting safety system settings 

or limiting conditions for operation are 

determined, nor is there any adverse effect on 

those plant systems necessary to assure the 

accomplishment of protection functions. The 

proposed changes will not result in plant 

operation in a configuration outside the 

design basis. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 12: TSTF–308–A, Revision 1, 

‘‘Determination of Cumulative and Projected 

Dose Contributions in RECP’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises Specification 

5.5.4, ‘‘Radioactive Effluent Controls 

Program,’’ paragraph e, to describe the 

original intent of the dose projections. The 

cumulative and projection of doses due to 

liquid releases are not an assumption in any 

accident previously evaluated and have no 

effect on the mitigation of any accident 

previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change revises Specification 

5.5.4, ‘‘Radioactive Effluent Controls 

Program,’’ paragraph e, to describe the 

original intent of the dose projections. The 

cumulative and projection of doses due to 

liquid releases are administrative tools to 

assure compliance with regulatory limits. 

The proposed change revises the requirement 

to clarify the intent, thereby improving the 

administrative control over this process. As 

a result, any effect on the margin of safety 

should be minimal. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 13: TSTF–312–A, Revision 1, 

‘‘Administrative Control of Containment 

Penetrations’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change would allow 

containment penetrations to be unisolated 

under administrative controls during core 


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alterations or movement of irradiated fuel 

assemblies within containment. The status of 

containment penetration flow paths (i.e., 

open or closed) is not an initiator for any 

design basis accident or event, and therefore 

the proposed change does not increase the 

probability of any accident previously 

evaluated. The proposed change does not 

affect the design of the primary containment, 

or alter plant operating practices such that 

the probability of an accident previously 

evaluated would be significantly increased. 

The proposed change does not significantly 

change how the plant would mitigate an 

accident previously evaluated, and is 

bounded by the fuel handling accident (FHA) 

analysis. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Allowing penetration flow paths to be open 

is not an initiator for any accident. The 

proposed change to allow open penetration 

flow paths will not affect plant safety 

functions or plant operating practices such 

that a new or different accident could be 

created. There are no design changes 

associated with the proposed changes, and 

the change does not involve a physical 

alteration of the plant (i.e., no new or 

different type of equipment will be installed). 

The change does not alter assumptions made 

in the safety analysis, and is consistent with 

the safety analysis assumptions and current 

plant operating practice. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

TS 3.9.3 provides measures to ensure that 

the dose consequences of a postulated FHA 

inside containment are minimized. The 

proposed change to LCO 3.9.3 will allow 

penetration flow path(s) to be open during 

refueling operations under administrative 

control. These administrative controls will 

provide assurance that prompt closure of 

open penetrations flow paths can and will be 

achieved in the event of an FHA inside 

containment, and will minimize dose 

consequences. The proposed change is 

bounded by the existing FHA analysis. The 

proposed change does not affect the safety 

analysis acceptance criteria for any analyzed 

event, nor is there a change to any safety 

analysis limit. The proposed change does not 

alter the manner in which safety limits, 

limiting safety system settings or limiting 

conditions for operation are determined, nor 

is there any adverse effect on those plant 

systems necessary to assure the 

accomplishment of protection functions. The 

proposed change will not result in plant 

operation in a configuration outside the 

design basis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 14: TSTF–314–A, Revision 0, 

‘‘Require Static and Transient FQ

Measurement’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the Required 

Actions of Specification 3.1.4, ‘‘Rod Group 

Alignment Limits,’’ and Specification 3.2.4, 

‘‘Quadrant Power Tilt Ratio,’’ to require 

measurement of both the steady state and 

transient portions of the Heat Flux Hot 

Channel Factor, FQ(Z). This change will 

ensure that the hot channel factors are within 

their limits when the rod alignment limits or 

quadrant power tilt ratio are not within their 

limits. The verification of hot channel factors 

is not an initiator of any accident previously 

evaluated. The verification that both the 

steady state and transient portion of FQ(Z) are 

within their limits will ensure this initial 

assumption of the accident analysis is met 

should a previously evaluated accident 

occur. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change revises the Required 

Actions in the Specifications for Rod Group 

Alignment Limits and Quadrant Power Tilt 

Ratio to require measurement of both the 

steady state and transient portions of the 

Heat Flux Hot Channel Factor, FQ(Z). This 

change is a correction that ensures that the 

plant conditions are as assumed in the 

accident analysis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 15: TSTF–315–A, Revision 0, 

‘‘Reduce Plant Trips Due to Spurious Signals 

to the NIS During Physics Testing’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises Specification 

3.1.8, ‘‘PHYSICS TESTS Exceptions—MODE 

2,’’ to allow the number of channels required 

by LCO 3.3.1, ‘‘RTS Instrumentation,’’ to be 

reduced from ‘‘4’’ to ‘‘3’’ to allow one nuclear 

instrumentation channel to be used as an 

input to the reactivity computer for physics 

testing without placing the nuclear 

instrumentation channel in a tripped 

condition. A reduction in the number of 

required nuclear instrumentation channels is 

not an initiator to any accident previously 

evaluated. With the nuclear instrumentation 

channel placed in bypass instead of in trip, 

reactor protection is provided by the 

intermediate range neutron flux detectors 

and the nuclear instrumentation system 

operating in a two-out-of-three channel logic. 

As a result, the ability to mitigate any 

accident previously evaluated is not 

significantly affected. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change reduces the 

probability of a spurious reactor trip during 

physics testing. The reactor trip system 

continues to be capable of protecting the 

reactor utilizing the intermediate range 

neutron flux reactor trip and the power range 

neutron flux trips operating in a two-out-ofthree trip logic. As a result, the reactor is 

protected and the probability of a spurious 

reactor trip is significantly reduced. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 


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5810 Federal Register / Vol. 80, No. 22 / Tuesday, February 3, 2015 / Notices 

amendment request involves no 

significant hazards consideration. 

Request No. 16: TSTF–325, Revision 0, 

‘‘ECCS Conditions and Required Actions 

with Less Than 100% Equivalent ECCS 

Flow’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change corrects the structure 

of Technical Specification 3.5.2 to assure its 

proper application. There is no change in 

intent or in the way the Technical 

Specification is applied. The literal (and 

unintended) interpretation of the existing 

LCO structure could, under some 

circumstances, provide longer than intended 

Completion Times for restoration of 

operability. The proposed change only 

clarifies the requirements of the Required 

Actions. Since the proposed change affects 

neither the Technical Specification intent, 

nor its application, the proposed change will 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change corrects the structure 

of the Technical Specification to assure its 

correct application. There is no change in 

intent or in the way the Technical 

Specification is applied. The proposed 

changes would not result in any physical 

alterations to the plant configuration, no new 

equipment is added, no equipment interfaces 

are modified, and no changes to any 

equipment’s function or the method of 

operating the equipment are being made. As 

the proposed changes would not change the 

design, configuration or operation of the 

plant, no new or different kinds of accident 

modes are created. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change corrects the structure 

of the Technical Specification to assure its 

correct application. There is no change in 

intent or in the way the Technical 

Specification is applied. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 17: TSTF–340–A, Revision 3, 

‘‘Allow 7 Day Completion Time for a 

Turbine-Driven AFW Pump Inoperable’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises Specification 

3.7.5, ‘‘Auxiliary Feedwater (AFW) System,’’ 

to allow a 7 day Completion Time to restore 

an inoperable turbine-driven pump in Mode 

3 immediately following a refueling outage, 

if Mode 2 has not been entered. An 

inoperable AFW turbine-driven pump is not 

an initiator of any accident previously 

evaluated. The ability of the plant to mitigate 

an accident is no different while in the 

extended Completion Time than during the 

existing Completion Time. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change revises Specification 

3.7.5, ‘‘Auxiliary Feedwater (AFW) System,’’ 

to allow a 7 day Completion Time to restore 

an inoperable turbine-driven AFW pump in 

Mode 3 immediately following a refueling 

outage if Mode 2 has not been entered. In 

Mode 3 immediately following a refueling 

outage, core decay heat is low and the need 

for AFW is also diminished. The two 

operable motor driven AFW pumps are 

available and there are alternate means of 

decay heat removal if needed. As a result, the 

risk presented by the extended Completion 

Time is minimal. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 18: TSTF–343, Revision 1, 

‘‘Containment Structural Integrity’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes revise the Technical 

Specifications (TS) Administrative Controls 

programs for consistency with the 

requirements of 10 CFR 50, paragraph 

55a(g)(4) for components classified as Code 

Class CC. The proposed changes affect the 

frequency of visual examinations that will be 

performed for the concrete surfaces of the 

containment for the purpose of the 

Containment Leakage Rate Testing Program, 

and allows those examinations to be 

performed during power operation in 

addition to during a refueling outage. 

The frequency of visual examinations of 

the containment and the mode of operation 

during which those examinations are 

performed does not affect the initiation of 

any accident previously evaluated. The use 

of NRC approved methods and frequencies 

for performing the inspections will ensure 

the containment continues to perform the 

mitigating function assumed for accidents 

previously evaluated. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes revise the TS 

Administrative Controls programs for 

consistency with the requirements of 10 CFR 

50, paragraph 55a(g)(4) for components 

classified as Code Class CC. The proposed 

changes affect the frequency of visual 

examinations that will be performed for the 

concrete surfaces of the containment for the 

purpose of the Containment Leakage Rate 

Testing Program, and allows those 

examinations to be performed during power 

operation in addition to during a refueling 

outage. 

The proposed changes do not involve a 

modification to the physical configuration of 

the plant (i.e., no new equipment will be 

installed) or change in the methods 

governing normal plant operation. The 

proposed changes will not impose any new 

or different requirements or introduce a new 

accident initiator, accident precursor, or 

malfunction mechanism. Additionally, there 

is no change in the types or increases in the 

amounts of any effluent that may be released 

off-site and there is no increase in individual 

or cumulative occupational exposure. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes revise the Technical 

Specifications (TS) Administrative Controls 

programs for consistency with the 

requirements of 10 CFR 50, paragraph 

55a(g)(4) for components classified as Code 

Class CC. The proposed changes affect the 

frequency of visual examinations that will be 

performed for the concrete surfaces of the 

containment for the purpose of the 

Containment Leakage Rate Testing Program, 


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Federal Register / Vol. 80, No. 22 / Tuesday, February 3, 2015 / Notices 5811 

and allows those examinations to be 

performed during power operation in 

addition to during a refueling outage. The 

safety function of the containment as a 

fission product barrier will be maintained. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 19: TSTF–349–A, Revision 1, 

‘‘Add Note to LCO 3.9.5 Allowing Shutdown 

Cooling Loops Removal from Operation’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change adds an LCO Note to 

LCO 3.9.5, ‘‘RHR and Coolant Circulation— 

Low Water Level,’’ to allow securing the 

operating train of Residual Heat Removal 

(RHR) for up to 15 minutes to support 

switching operating trains. The allowance is 

restricted to conditions in which core outlet 

temperature is maintained at least 10 degrees 

F below the saturation temperature, when 

there are no draining operations, and when 

operations that could reduce the reactor 

coolant system (RCS) boron concentration are 

prohibited. Securing an RHR train to 

facilitate the changing of the operating train 

is not an initiator to any accident previously 

evaluated. The restrictions on the use of the 

allowance ensure that an RHR train will not 

be needed during the 15 minute period to 

mitigate any accident previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change adds an LCO Note to 

LCO 3.9.5, ‘‘RHR and Coolant Circulation— 

Low Water Level,’’ to allow securing the 

operating train of RHR to support switching 

operating trains. The allowance is restricted 

to conditions in which core outlet 

temperature is maintained at least 10 degrees 

F below the saturation temperature, when 

there are no draining operations, and when 

operations that could reduce the reactor 

coolant system (RCS) boron concentration are 

prohibited. With these restrictions, combined 

with the short time frame allowed to swap 

operating RHR trains and the ability to start 

an operating RHR train if needed, the 

occurrence of an event that would require 

immediate operation of an RHR train is 

extremely remote. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 20: TSTF–355–A, Revision 0, 

‘‘Changes to RTS and ESF Tables’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The RTS [Reactor Trip System] and ESFAS 

[Engineered Safety Feature Actuations 

System] instrument functions are part of the 

accident mitigation response and are not 

themselves an initiator of any accident 

previously evaluated. Therefore, the 

probability of an accident previously 

evaluated is not significantly affected by the 

proposed changes. The changes ensure that 

automatic protective actions will be initiated 

at or before the condition assumed in the 

safety analysis, and are in accordance with 

the intent of the Technical Specifications. 

The proposed changes will not cause any 

design or analysis acceptance criteria to be 

exceeded. Since there will be no adverse 

effect on the trip setpoints or the 

instrumentation associated with the trip 

setpoints, there will be no significant 

increase in the consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes include 

modifications to the format of the nominal 

trip setpoints that preserve safety analysis 

assumptions related to accident mitigation. 

The protection system will continue to 

initiate the protective actions as assumed in 

the safety analysis. The proposed changes 

will continue to ensure that the trip setpoints 

are maintained consistent with the setpoint 

methodology and the plant safety analysis. 

As the proposed changes do not change the 

design, configuration or operation of the 

plant, no new or different kinds of accident 

modes are created. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed changes do not alter any 

nominal trip setpoints, allowable values, or 

limiting safety system settings, and will 

continue to ensure that the trip setpoints are 

maintained consistent with the setpoint 

methodology and the plant safety analysis. 

The response of protection systems to 

accident transients reported in the Final 

Safety Analysis Report is unaffected by this 

change, and accident analysis acceptance 

criteria are consequently not affected. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 21: TSTF–371–A, Revision 1, 

‘‘NIS Power Range Channel Daily SR TS 

Change to Address Low Power 

Decalibration’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises Specification 

3.3.1, ‘‘RTS Instrumentation,’’ Surveillances 

3.3.1.2 and 3.3.1.3 to move requirements 

currently in a Note to the Surveillance itself. 

The change in presentation is editorial and 

does not affect the application of the 

Surveillances. The proposed change does not 

affect any accident initiators or analyzed 

events or assumed mitigation of accident or 

transient events. The proposed change does 

not involve the addition or removal of any 

equipment, or any design changes to the 

facility. 

Therefore, this proposed change does not 

represent a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change revises Specification 

3.3.1, ‘‘RTS Instrumentation,’’ Surveillances 

3.3.1.2 and 3.3.1.3 to move requirements 

currently in a Note to the Surveillance itself. 

The proposed change represents an editorial 

preference and does not affect the 


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5812 Federal Register / Vol. 80, No. 22 / Tuesday, February 3, 2015 / Notices 

performance of the Surveillance or plant 

operation. The safety function tested by the 

Surveillance is unaffected. 

Therefore, this proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 22: TSTF–439–A, Revision 2, 

‘‘Eliminate Second Completion Times 

Limiting Time From Discovery of Failure To 

Meet an LCO’’ 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change eliminates certain 

Completion Times from the Technical 

Specifications. Completion Times are not an 

initiator to any accident previously 

evaluated. As a result, the probability of an 

accident previously evaluated is not affected. 

The consequences of an accident during the 

remaining Completion Time are no different 

than the consequences of the same accident 

during the removed Completion Times. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of any accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change to delete the second 

Completion Time does not alter the manner 

in which safety limits, limiting safety system 

settings or limiting conditions for operation 

are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed changes will not result 

in plant operation in a configuration outside 

of the design basis. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 23: ISTS Adoption #1—Revise 

LCO 3.3.2 ESFAS Interlock P–4 Required 

Action Completion Time 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the Condition 

to be entered when the ESFAS Interlock P– 

4 is inoperable. Current Technical 

Specifications require restoring the channel 

to Operable status within 24 hours or be in 

Mode 3 within the next 12 hours and Mode 

5 within the following 52 hours. The 

proposed change provides 48 hours to restore 

the inoperable channel, or be in Mode 3 in 

54 hours and Mode 4 in 60 hours. The 

ESFAS P–4 interlock is not an initiator to any 

accident previously evaluated. The 

consequences of any accident previously 

evaluated during the proposed Completion 

Time are no different from the consequences 

during the existing Completion Time. As a 

result, the proposed change does not result 

in a significant increase in the consequences 

of any accident previously evaluated. 

Therefore, this proposed change does not 

represent a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change provides an 

additional 24 hours to restore an inoperable 

ESFAS P–4 Interlock. During the proposed 

Completion Time, manual actions can 

perform the functions provided by the 

inoperable P–4 interlock. Also, the proposed 

Completion Time is reasonable given the 

available redundant channel, and the low 

probability of an event occurring during this 

interval. 

Therefore, this proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Request No. 24: Revise LCO 3.5.5 to 8-hour 

Completion Time and Note allowance 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change modifies the LCO 

3.5.5, ‘‘Seal Injection Flow,’’ Action A, ‘‘Seal 

injection flow not within limit,’’ Completion 

Time from 4 hours to 8 hours and the Note 

to SR 3.5.5.1 to allow 8 hours instead of 4 

hours to stabilize reactor coolant system 

(RCS) pressure prior to verifying the seal 

injection throttle valves are properly 

adjusted. The proposed change does not 

involve the addition or removal of any 

equipment, or any design changes to the 

facility. Seal injection flow is not an initiator 

of any accident previously evaluated. The 

consequences of any accident previously 

evaluated during the extended Completion 

Time or Note allowance are the same as 

during the existing Completion Time and 

Note allowance. 

Therefore, this proposed change does not 

represent a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical alteration to the plant (i.e., no new 

or different type of equipment will be 

installed) or a change to the methods 

governing normal plant operation. The 

changes do not alter the assumptions made 

in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change provides additional 

time to verify seal injection flow is within 

limit or to restore seal injection flow to 

within limit if it is discovered that it is not 

within limit. The additional time is 

acceptable on the basis that there is little 

likelihood of an event that would challenge 

the ECCS occurring during the 8-hour 

window, and it reduces the pressure on the 

operations staff should iterations in the 

adjustment procedure be necessary to 

balance seal injection flow. 

Therefore, this proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Leigh D. Perry, 

SVP & General Counsel of Operations 

and Nuclear, Southern Nuclear 


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Operating Company, 40 Iverness Center 

Parkway, Birmingham, AL 35201. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: October 

2, 2014. A publicly-available version is 

in ADAMS under Accession No. 

ML14275A441. 

Description of amendment request: 

The proposed amendment upgrades the 

Emergency Action Level scheme by 

adopting NRC-endorsed Nuclear Energy 

Institute 99–01, Revision 6, 

‘‘Methodology for the Development of 

Emergency Action Levels for NonPassive Reactors,’’ issued January 2011 

(ADAMS Accession No. ML110240324). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes to the Callaway 

Plant emergency action levels do not impact 

the physical function of plant structures, 

systems, or components (SSC) or the manner 

in which SSCs perform their design function. 

The proposed changes neither adversely 

affect accident initiators or precursors, nor 

alter design assumptions. The proposed 

changes do not alter or prevent the ability of 

SSCs to perform their intended function to 

mitigate the consequences of an initiating 

event within assumed acceptance limits. No 

operating procedures or administrative 

controls that function to prevent or mitigate 

accidents are affected by the proposed 

changes. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not involve a 

physical alteration of the plant (i.e., no new 

or different type of equipment will be 

installed or removed) or a change in the 

method of plant operation. The proposed 

changes will not introduce failure modes that 

could result in a new accident, and the 

change does not alter assumptions made in 

the safety analysis. The proposed changes to 

the Callaway Plant emergency action levels 

are not initiators of any accidents. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Margin of safety is associated with the 

ability of the fission product barriers (i.e., 

fuel cladding, reactor coolant system 

pressure boundary, and containment 

structure) to limit the level of radiation dose 

to the public. The proposed changes do not 

impact operation of the plant or its response 

to transients or accidents. The changes do not 

affect the Technical Specifications or the 

operating license. The proposed changes do 

not involve a change in the method of plant 

operation, and no accident analyses will be 

affected by the proposed changes. 

Additionally, the proposed changes will not 

relax any criteria used to establish safety 

limits and will not relax any safety system 

settings. The safety analysis acceptance 

criteria are not affected by these changes. The 

proposed changes will not result in plant 

operation in a configuration outside the 

design basis. The proposed changes do not 

adversely affect systems that respond to 

safely shut down the plant and to maintain 

the plant in a safe shutdown condition. The 

emergency plan will continue to activate an 

emergency response commensurate with the 

extent of degradation of plant safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: John O’Neill, 

Esq., Pillsbury Winthrop Shaw Pittman 

LLP, 2300 N Street NW., Washington, 

DC 20037. 

Acting NRC Branch Chief: Eric R. 

Oesterle. 

III. Notice of Issuance of Amendments 

to Facility Operating Licenses and 

Combined Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

of 1954, as amended (the Act), and the 

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

A notice of consideration of issuance 

of amendment to facility operating 

license or combined license, as 

applicable, proposed no significant 

hazards consideration determination, 

and opportunity for a hearing in 

connection with these actions, was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items can be accessed as described in 

the ‘‘Obtaining Information and 

Submitting Comments’’ section of this 

document. 

Duke Energy Progress, Inc., Docket No. 

50–400, Shearon Harris Nuclear Power 

Plant, Unit 1, Wake and Chatham 

Counties, North Carolina 

Date of amendment request: April 24, 

2014. 

Brief description of amendment: The 

amendment revises Technical 

Specification (TS) 3/4.4.5, ‘‘Steam 

Generator Tube Integrity,’’ TS 6.8.4.I, 

‘‘Steam Generator Program,’’ and TS 

6.9.1.7, ‘‘Steam Generator Tube 

Inspection Report’’ to address 

implementation associated with the 

inspections and reporting requirements 

as described in Technical Specifications 

Task Force (TSTF) TSTF–510, Revision 

2, ‘‘Revision to Steam Generator 

Program Inspection Frequencies and 

Tube Sample Selection.’’ 

Date of issuance: January 9, 2015. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days of issuance. 

Amendment No.: 145. A publiclyavailable version is in ADAMS under 

Accession No. ML14307A800; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Facility Operating License No. NPF– 

63 The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 22, 2014 (79 FR 42543). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 9, 2015. 

No significant hazards consideration 

comments received: No. 


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5814 Federal Register / Vol. 80, No. 22 / Tuesday, February 3, 2015 / Notices 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of application for amendment: 

October 31, 2013, as supplemented by 

letters dated May 29, 2014, and 

September 9, 2014. 

Brief description of amendment: The 

amendment revised Technical 

Specification Surveillance 

Requirements 3.5.1.4 and 3.5.2.5 for low 

pressure core spray and low pressure 

coolant injection pump flows. 

Date of issuance: January 7, 2015. 

Effective date: As of its date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment No.: 229. A publiclyavailable version is in ADAMS under 

Accession No. ML14335A189; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. NPF–21: The amendment revised 

the Facility Operating License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: April 8, 2014 (79 FR 19399). 

The supplemental letters dated May 29, 

2014, and September 9, 2014, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 7, 2015. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50– 

368, Arkansas Nuclear One, Unit 2, 

Pope County, Arkansas 

Date of application for amendment: 

January 21, 2014, as supplemented by 

letters dated March 17 and September 

24, 2014. 

Brief description of amendment: The 

amendment revised the Technical 

Specification 6.5.16 requirements for 

the local leak test required for the 

containment building emergency escape 

air lock doors, in that it would require 

a seal contact verification in lieu of the 

current seal pressure test to verify leak 

tightness. 

Date of issuance: January 22, 2015. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 299. A publiclyavailable version is in ADAMS under 

Accession No. ML14350B285; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. NPF–6: Amendment revised the 

Technical Specifications/license. 

Date of initial notice in Federal 

Register: April 15, 2014 (79 FR 21296). 

The supplemental letter dated 

September 24, 2014, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 22, 

2015. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–352 and 50–353, 

Limerick Generating Station, Units 1 

and 2, Montgomery County, 

Pennsylvania 

Date of amendment request: 

November 15, 2013, as supplemented by 

letters dated April 16, 2014; September 

11, 2014; and November 7, 2014. 

Brief description of amendments: The 

amendments revise the Technical 

Specification (TS) requirements related 

to the response time for the main steam 

line flow-high isolation function. 

Date of issuance: January 7, 2015. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 214 and 175. A 

publicly-available version is in ADAMS 

under Accession No. ML14344A681; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Renewed Facility Operating License 

Nos. NPF–39 and NPF–85: Amendments 

revised the Renewed Facility Operating 

License and TSs. 

Date of initial notice in Federal 

Register: February 4, 2014 (79 FR 

6642). The supplemental letters dated 

April 16, 2014; September 11, 2014; and 

November 7, 2014, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated January 7, 2015. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Units 1 

and 2, Rock Island County, Illinois 

Date of amendment requests: July 16, 

2013, as supplemented by letters dated 

September 18, 2013, January 22, April 7, 

August 12, and November 11, 2014. 

Brief description of amendments: The 

amendments revises the Technical 

Specifications to include the use of 

neutron absorbing spent fuel pool rack 

inserts (i.e., NETCO–SNAP–IN® rack 

inserts) for the purpose of criticality 

control in the spent fuel pools. 

Date of issuance: December 31, 2014. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days. 

Amendment Nos.: 253–Unit 1; 248– 

Unit 2. A publicly-available version is 

in ADAMS under Accession No. 

ML14346A306; documents related to 

these amendments are listed in the 

safety evaluation enclosed with the 

amendments. 

Renewed Facility Operating License 

Nos. DPR–29 and DPR–30: The 

amendments revised the Technical 

Specifications and Facility Operating 

License. 

Date of initial notice in Federal 

Register: July 8, 2014 (79 FR 38577). 

The supplemental letters dated 

September 18, 2013, January 22, April 7, 

August 12, and November 11, 2014, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated December 31, 

2014. 

No significant hazards consideration 

comments received: No. 

Northern States Power Company— 

Minnesota, Docket No. 50–263, 

Monticello Nuclear Generating Plant 

(MNGP), Wright County, Minnesota 

Date of amendment request: 

November 14, 2013. 

Brief description of amendment: The 

amendment revises Technical 

Specification (TS) 5.5.11, ‘‘Primary 

Containment Leakage Rate Testing 

Program,’’ by removing TS 5.5.11.d.2.b, 

the reduced pressure testing option for 

drywell airlock door leakage testing. 

This testing methodology is not required 

and does not reflect the current testing 

practice at MNGP. As such, the drywell 


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Federal Register / Vol. 80, No. 22 / Tuesday, February 3, 2015 / Notices 5815 

airlock door seals will be tested by 

performing an overall airlock leakage 

test as specified in current TS 

5.5.11.d.2.a. 

Date of issuance: January 8, 2015. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days of issuance. 

Amendment No.: 187. A publiclyavailable version is in ADAMS under 

Accession No. ML14323A033; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. DPR–22: This amendment revises 

the Renewed Facility Operating License 

and the Technical Specifications. 

Date of initial notice in Federal 

Register: August 5, 2014 (79 FR 45478). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 8, 2015. 

No significant hazards consideration 

comments received: No. 

South Carolina Electric and Gas 

Company Docket Nos. 52–027 and 52– 

028, Virgil C. Summer Nuclear Station 

(VCSNS), Units 2 and 3, Fairfield 

County, South Carolina 

Date of amendment request: April 3, 

2014, as supplemented by letter dated 

May 19, 2014. 

Brief description of amendment: The 

amendment revises Tier 2* information, 

incorporated into the VCSNS Units 2 

and 3 Updated Final Safety Analysis 

Report (UFSAR). Specifically, the 

amendment revises the details regarding 

the structural floor of the Auxiliary 

Building and its constructability. Notes 

are added to drawings in Subsection 

3H.5 of the UFSAR in order to clarify 

variations in detail design such as size 

and spacing or reinforcement and spans 

of the noncritical sections of floors. 

Date of issuance: July 18, 2014. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 14. A publiclyavailable version is in ADAMS under 

Accession No. ML14188B185; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Facility Combined Licenses No. NPF– 

93 and NPF–94: Amendment revised the 

Facility Combined Licenses. 

Date of initial notice in Federal 

Register: April 29, 2014 (79 FR 24024). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 18, 2014. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company 

Docket Nos. 52–025 and 52–026, Vogtle 

Electric Generating Plant (VEGP) Units 

3 and 4, Burke County, Georgia 

Date of amendment request: March 

17, 2014, and revised by letters dated 

May 8, September 2, and October 2, 

2014. 

Brief description of amendment: The 

amendment revises the VEGP Units 3 

and 4 Updated Final Safety Analysis 

Report (UFSAR) by clarifying how 

human diversity was applied during the 

design process for the Component 

Interface Module and Diverse Actuation 

System. The changes to the VEGP Units 

3 and 4 UFSAR include changes to 

Table 1.6, ‘‘Material Referenced,’’ 

Chapter 7, Sections 7.1.2.14.1, 7.1.7 and 

7.2.4 and the addition of Appendix 7A 

to Chapter 7. The changes to the VEGP 

Units 3 and 4 UFSAR modify 

information related to human diversity, 

as presented in a Tier 2* document, 

WCAP–17179–P and WCAP–17179–NP, 

‘‘AP1000 Component Interface Module 

Technical Report,’’ Revision 2, and two 

Tier 2 documents, WCAP–15775, 

‘‘AP1000 Instrumentation and Control 

Defense-in-Depth and Diversity Report,’’ 

Revision 4 and WCAP–17184–P, 

‘‘AP1000 Diverse Actuation System 

Planning and Functional Design 

Summary Technical Report,’’ that are 

incorporated by reference in the VEGP 

Units 3 and 4 UFSAR. 

Date of issuance: December 24, 2014. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 28. A publiclyavailable version is in ADAMS under 

Accession No. ML14329A298; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Facility Combined Licenses No. NPF– 

91 and NPF–92: Amendment revised the 

Facility Combined Licenses. 

Date of initial notice in Federal 

Register: April 29, 2014 (79 FR 24021). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated December 24, 

2014. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company 

Docket Nos. 52–025 and 52–026, Vogtle 

Electric Generating Plant (VEGP) Units 

3 and 4, Burke County, Georgia 

Date of amendment request: August 

22, 2014, and revised by letter dated 

September 23, 2014, and supplemented 

by letters dated October 30 and 

November 6, 2014. 

Brief description of amendment: The 

amendment revises the VEGP Units 3 

and 4 Updated Final Safety Analysis 

Report to reflect changes related to: 

(a) Installation of an additional nonsafety-related battery; 

(b) Revision to the annex building 

internal configuration by converting a 

shift turnover room to a battery room, 

adding an additional battery equipment 

room, and moving a fire area wall; 

(c) Increase in the height of a room in 

the annex building; and 

(d) Increase in thicknesses of certain 

annex building floor slabs. 

In addition, the proposed changes 

also include reconfiguring existing 

rooms and related rooms, wall, and 

access path changes and making 

changes to the corresponding Tier 1 

information in Appendix C to the 

Combined Licenses. 

Date of issuance: December 23, 2014. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 27. A publiclyavailable version is in ADAMS under 

Accession No. ML14323A609; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Facility Combined Licenses No. NPF– 

91 and NPF–92: Amendment revised the 

Facility Combined Licenses. 

Date of initial notice in Federal 

Register: October 14, 2014 (79 FR 

61662). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated December 23, 

2014. 

No significant hazards consideration 

comments received: No. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of application for amendment: 

January 23, 2014. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) 3.4.12, ‘‘Cold 

Overpressure Mitigation System 

(COMS),’’ to reflect the mass input 

transient analysis that assumes an 

Emergency Core Cooling System 

centrifugal charging pump and the 

normal charging pump capable of 

injecting into the reactor coolant system 

when TS 3.4.12 is applicable. The 

amendment also revised TS Table 3.3.1– 

1, ‘‘Reactor Trip System 

Instrumentation,’’ to remove 

unnecessary page number references. 

Date of issuance: January 20, 2015. 

Effective date: As of its date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 210. A publiclyavailable version is in ADAMS under 


rljohnson on DSK3VPTVN1PROD with NOTICES
5816 Federal Register / Vol. 80, No. 22 / Tuesday, February 3, 2015 / Notices 

Accession No. ML14350B239; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Facility Operating License No. NPF– 

30: The amendment revised the 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: April 1, 2014 (79 FR 18348). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 20, 

2015. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 26th day 

of January 2015. 

For the Nuclear Regulatory Commission. 

Michele G. Evans, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2015–01917 Filed 2–2–15; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2015–0007] 

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 three 

amendment requests. The amendment 

requests are for Catawba Nuclear 

Station, Units 1 and 2; McGuire Nuclear 

Station, Units 1 and 2; Oconee Nuclear 

Station, Units 1, 2, and 3; Perry Nuclear 

Plant, Unit 1; and Browns Ferry Nuclear 

Plant, Unit 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 

March 2, 2015. A request for a hearing 

must be filed by March 31, 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 February 9, 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–2015–0007. Address 

questions about NRC dockets to Carol 

Gallagher; telephone: 301–287–3422; 

email: Carol.Gallagher@nrc.gov. For 

technical questions, contact the 

individual listed in the FOR FURTHER

INFORMATION CONTACT section of this 

document. 

• Mail comments to: Cindy Bladey, 

Office of Administration, Mail Stop: 

3WFN–06–A44M, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001. 

For additional direction on obtaining 

information and submitting comments, 

see ‘‘Obtaining Information and 

Submitting Comments’’ in the 

SUPPLEMENTARY INFORMATION section of 

this document. 

FOR FURTHER INFORMATION CONTACT: 

Mable A. Henderson, Office of Nuclear 

Reactor Regulation, U.S. Nuclear 

Regulatory Commission, Washington DC 

20555–0001; telephone: 301–415–3760, 

email: Mable.Henderson@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Obtaining Information and 

Submitting Comments 

A. Obtaining Information 

Please refer to Docket ID NRC–2015– 

0007 when contacting the NRC about 

the availability of information regarding 

this document. You may obtain 

publicly-available information related to 

this action by the following methods: 

• Federal Rulemaking Web site: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2015–0007. 

• 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– 

0007, 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 


rljohnson on DSK3VPTVN1PROD with NOTICES