Federal Register / Vol. 79, No. 208 / Tuesday, October 28, 2014 / Notices 64219 

NATIONAL ARCHIVES AND RECORDS 

ADMINISTRATION 

[NARA–2015–006] 

Agency Information Collection 

Activities: Proposed Collection; 

Comment Request 

AGENCY: National Archives and Records 

Administration (NARA). 

ACTION: Notice. 

SUMMARY: NARA is giving public notice 

that the agency proposes to request 

extension of a currently approved 

information collection used when 

veterans or other authorized individuals 

request information from or copies of 

documents in military service records. 

The public is invited to comment on the 

proposed information collection 

pursuant to the Paperwork Reduction 

Act of 1995. 

DATES: Written comments must be 

received on or before December 29, 2014 

to be assured of consideration. 

ADDRESSES: Comments should be sent 

to: Paperwork Reduction Act Comments 

(ISP), Room 4400, National Archives 

and Records Administration, 8601 

Adelphi Rd, College Park, MD 20740– 

6001; or faxed to 301–713–7409; or 

electronically mailed to 

tamee.fechhelm@nara.gov. 

FOR FURTHER INFORMATION CONTACT: 

Requests for additional information or 

copies of the proposed information 

collections and supporting statements 

should be directed to Tamee Fechhelm 

at telephone number 301–837–1694, or 

fax number 301–713–7409. 

SUPPLEMENTARY INFORMATION: Pursuant 

to the Paperwork Reduction Act of 1995 

(Pub. L. 104–13), NARA invites the 

general public and other Federal 

agencies to comment on proposed 

information collections. The comments 

and suggestions should address one or 

more of the following points: (a) 

Whether the proposed collection 

information is necessary for the proper 

performance of the functions of NARA; 

(b) the accuracy of NARA’s estimate of 

the burden of the proposed information 

collections; (c) ways to enhance the 

quality, utility, and clarity of the 

information to be collected; and (d) 

ways to minimize the burden of the 

collection of information on 

respondents, including the use of 

information technology; and (e) whether 

small businesses are affected by this 

collection. The comments that are 

submitted will be summarized and 

included in the NARA request for Office 

of Management and Budget (OMB) 

approval. All comments will become a 

matter of public record. In this notice, 

NARA is soliciting comments 

concerning the following information 

collection: 

Title: Request Pertaining to Military 

Records. 

OMB number: 3095–0029. 

Agency form number: SF 180. 

Type of review: Regular. 

Affected public: Veterans, their 

authorized representatives, state and 

local governments, and businesses. 

Estimated number of respondents: 

1,028,769. 

Estimated time per response: 5 

minutes. 

Frequency of response: On occasion 

(when respondent wishes to request 

information from a military personnel 

record). 

Estimated total annual burden hours: 

85,731 hours. 

Abstract: The authority for this 

information collection is contained in 

36 CFR 1233.18(d). In accordance with 

rules issued by the Department of 

Defense (DOD) and Department of 

Homeland Security (DHS, US Coast 

Guard), the National Personnel Records 

Center (NPRC) of the National Archives 

and Records Administration (NARA) 

administers military service records of 

veterans after discharge, retirement, and 

death. When veterans and other 

authorized individuals request 

information from or copies of 

documents in military service records, 

they must provide in forms or in letters 

certain information about the veteran 

and the nature of the request. Federal 

agencies, military departments, 

veterans, veterans’ organizations, and 

the general public use Standard Forms 

(SF) 180, Request Pertaining to Military 

Records, in order to obtain information 

from military service records stored at 

NPRC. Veterans and next-of-kin of 

deceased veterans can also use eVetRecs 

(http://www.archives.gov/ 

research_room/vetrecs/) to order copies. 

Dated: October 20, 2014. 

Swarnali Haldar, 

Executive for Information Services/CIO. 

[FR Doc. 2014–25581 Filed 10–27–14; 8:45 am] 

BILLING CODE 7515–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2014–0239] 

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 October 2, 

2014 to October 15, 2014. The last 

biweekly notice was published on 

October 14, 2014. 

DATES: Comments must be filed by 

November 28, 2014. A request for a 

hearing must be filed by December 29, 

2014. 

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–2014–0239. 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: 

Shirley Rohrer, Office of Nuclear 

Reactor Regulation, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001; telephone: 301–415– 

5411, email: Shirley.Rohrer@nrc.gov 

SUPPLEMENTARY INFORMATION: 


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64220 Federal Register / Vol. 79, No. 208 / Tuesday, October 28, 2014 / Notices 

I. Obtaining Information and 

Submitting Comments 

A. Obtaining Information 

Please refer to Docket ID NRC–2014– 

0239 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–2014–0239. 

• 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–2014– 

0239 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 

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 


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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 10 

days prior to the filing deadline, the 

participant should contact the Office of 

the Secretary by email at 

hearing.docket@nrc.gov, or by telephone 

at 301–415–1677, to request (1) a digital 

identification (ID) certificate, which 

allows the participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and (2) advise the 

Secretary that the participant will be 

submitting a request or petition for 

hearing (even in instances in which the 

participant, or its counsel or 

representative, already holds an NRCissued digital ID certificate). Based upon 

this information, the Secretary will 

establish an electronic docket for the 

hearing in this proceeding if the 

Secretary has not already established an 

electronic docket. 

Information about applying for a 

digital ID certificate is available on the 

NRC’s public Web site at http://

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

getting-started.html. System 

requirements for accessing the ESubmittal server are detailed in the 

NRC’s ‘‘Guidance for Electronic 

Submission,’’ which is available on the 

agency’s public Web site at http://

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

attempt to use other software not listed 

on the Web site, but should note that the 

NRC’s E-Filing system does not support 

unlisted software, and the NRC Meta 

System Help Desk will not be able to 

offer assistance in using unlisted 

software. 

If a participant is electronically 

submitting a document to the NRC in 

accordance with the E-Filing rule, the 

participant must file the document 

using the NRC’s online, Web-based 

submission form. In order to serve 

documents through the Electronic 

Information Exchange System, users 

will be required to install a Web 

browser plug-in from the NRC’s Web 

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

installation of the Web browser plug-in, 

is available on the NRC’s public Web 

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

Once a participant has obtained a 

digital ID certificate and a docket has 

been created, the participant can then 

submit a request for hearing or petition 

for leave to intervene. Submissions 

should be in Portable Document Format 

(PDF) in accordance with NRC guidance 

available on the NRC’s public Web site 

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

complete at the time the documents are 

submitted through the NRC’s E-Filing 

system. To be timely, an electronic 

filing must be submitted to the E-Filing 

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

Time on the due date. Upon receipt of 

a transmission, the E-Filing system 

time-stamps the document and sends 

the submitter an email notice 

confirming receipt of the document. The 

E-Filing system also distributes an email 

notice that provides access to the 

document to the NRC’s Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically using 

the NRC’s adjudicatory E-Filing system 

may seek assistance by contacting the 

NRC Meta System Help Desk through 

the ‘‘Contact Us’’ link located on the 

NRC’s public Web site at http://

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

MSHD.Resource@nrc.gov, or by a tollfree call at 1–866–672–7640. The NRC 

Meta System Help Desk is available 

between 8 a.m. and 8 p.m., Eastern 

Time, Monday through Friday, 

excluding government holidays. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file an 

exemption request, in accordance with 

10 CFR 2.302(g), with their initial paper 

filing requesting authorization to 

continue to submit documents in paper 

format. Such filings must be submitted 

by: (1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, Attention: Rulemaking and 

Adjudications Staff; or (2) courier, 

express mail, or expedited delivery 

service to the Office of the Secretary, 

Sixteenth Floor, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland 20852, Attention: Rulemaking 

and Adjudications Staff. Participants 

filing a document in this manner are 

responsible for serving the document on 

all other participants. Filing is 

considered complete by first-class mail 

as of the time of deposit in the mail, or 

by courier, express mail, or expedited 

delivery service upon depositing the 

document with the provider of the 

service. A presiding officer, having 


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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. 

Duke Energy Florida, Inc. et al. (DEF), 

Docket No. 50–302, Crystal River, Unit 

3, Nuclear Generating Plant (CR–3), 

Citrus County, Florida 

Date of amendment request: October 

29, 2013, as supplemented by letters 

dated May 7, 2014 and June 17, 2014. 

Publicly-available versions are in 

ADAMS under Accession Nos. 

ML13316C083, ML14139A006, and 

ML14178B284. 

Description of amendment request: 

The amendment would revise the CR– 

3 Facility Operating License (FOL) to 

remove and revise certain License 

Conditions. This amendment also 

proposes to extensively revise the CR– 

3 Improved Technical Specifications 

(ITS) in order to create the CR–3 

Permanently Defueled Technical 

Specifications (PDTS). 

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. 

CR–3 has permanently ceased operation. 

The proposed amendment would modify the 

CR–3 FOL and ITS by proposing to delete 

certain License Conditions (LCs) and ITS that 

are no longer applicable to a permanently 

defueled facility, while modifying the 

remaining portions to correspond to the 

permanently shutdown condition. Changes 

proposed to LCs will make them consistent 

with the non-operating status of CR–3. Other 

proposed LCs changes will eliminate LCs that 

were designed for one time implementation 

and have been satisfied, or are no longer 

required due to changes to Part 50 or Part 73 

regulations that accomplish the same result 

or eliminate the requirement for the LC. The 

proposed changes to the ITS are consistent 

with the criteria set forth in 10 CFR 50.36 for 

the contents of ITS. 

Chapter 14 of the CR–3 Final Safety 

Analysis Report (FSAR) described the design 

basis accident (DBA) and transient scenarios 

applicable to CR–3 during power operations. 

With the reactor in a permanently defueled 

condition, the spent fuel pool and its cooling 

systems are dedicated only to spent fuel 

storage. In this condition, the spectrum of 

credible accidents is much smaller than for 

an operational plant. As a result of the 

certifications submitted by CR–3 in 

accordance with 10 CFR 50.82(a)(1), and the 

consequent removal of authorization to 

operate the reactor or to place or retain fuel 

in the reactor vessel in accordance with 10 

CFR 50.82(a)(2), the majority of the accident 

scenarios originally postulated in the FSAR 

are no longer possible and have been 

removed from the FSAR under 10 CFR 50.59. 

The definition of safety-related structures, 

systems, and components (SSCs) in 10 CFR 

50.2 states that safety-related SSCs are those 

relied on to remain functional during and 

following design basis events to assure: 

1. The integrity of the reactor coolant 

boundary; 

2. The capability to shutdown the reactor 

and maintain it in a safe shutdown condition; 

or 

3. The capability to prevent or mitigate the 

consequences of accidents which could 

result in potential offsite exposures 

comparable to the applicable guideline 

exposures set forth in 10 CFR 50.34(a)(1) or 

100.11. 

The first two criteria, integrity of the 

reactor coolant pressure boundary and safe 

shutdown of the reactor, are not applicable 

to a plant in a permanently defueled 

condition. The third criterion is related to 

preventing or mitigating the consequences of 

accidents that could result in potential offsite 

exposures exceeding limits. However, after 

the termination of reactor operations at CR– 

3 and the permanent removal of the fuel from 

the reactor vessel (following 4 years of decay 

time after shutdown) and purging of the 

contents of the waste gas decay tanks, none 

of the SSCs at CR–3 are required to be relied 

on for accident mitigation. Therefore, none of 

the SSCs at CR–3 meet the definition of a 

safety-related SSC stated in 10 CFR 50.2 

(with the exception of the passive spent fuel 

pool structure). 

The deletion of ITS definitions and rules 

of usage and application, that are currently 

not applicable in a defueled condition, has 

no impact on facility SSCs or the methods of 

operation of such SSCs. The deletion of 

design features and safety limits not 

applicable to the permanently shutdown and 

defueled status of CR–3 has no impact on the 

remaining DBA [design basis accidents] (the 

Fuel Handling Accident in the Auxiliary 

Building) or the proposed Radioactive Waste 

Handling Accident. The removal of LCOs 

[Limiting Conditions for Operation] or SRs 

[Surveillance Requirements] that are related 

only to the operation of the nuclear reactor 

or accidents do not affect mitigation of the 

applicable DBAs previously evaluated since 

these DBAs are no longer applicable in the 

defueled mode. The safety functions 

involving core reactivity control, reactor heat 

removal, reactor coolant system inventory 

control, and containment integrity are no 

longer applicable at CR–3 as a permanently 

defueled plant. The analyzed accidents 

involving damage to the reactor coolant 

system, main steam lines, reactor core, and 

the subsequent release of radioactive material 

are no longer possible at CR–3 

Since CR–3 has permanently ceased 

operation, the generation of fission products 

has ceased and the remaining source term 

will decay. The radioactive decay of the 

irradiated fuel since shutdown of the reactor 

have reduced the consequences of the Fuel 

Handling Accident (FHA) to levels well 

below those previously analyzed. The 

relevant parameter (water level) associated 

with the fuel pool provides an initial 

condition for the FHA analysis and is 

included in the PDTS. 

The spent fuel pool water level, spent fuel 

pool boron concentration, and spent fuel 

pool storage LCOs are retained to preserve 

the current requirements for safe storage of 

irradiated fuel. 

Fuel pool cooling and makeup related 

equipment and support equipment (e.g., 

electrical power systems) are not required to 

be continuously available since there is 

sufficient time to effect repairs, establish 

alternate sources of makeup flow, or establish 

alternate sources of cooling in the event of a 

loss of cooling and makeup flow to the spent 

fuel pool. 

The deletion and modification of 

provisions of the Administrative Controls do 

not directly affect the design of SSCs 

necessary for the safe storage of irradiated 

fuel or the methods used for handling and 


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storage of such fuel in the fuel pool. Deletion 

of Programs are administrative in nature and 

do not affect any accidents applicable to the 

safe management of irradiated fuel or the 

permanently shutdown and defueled 

condition of the reactor. 

The proposed LC revisions reflect the CR– 

3 functions that are still authorized in the 

permanently defueled condition, and remove 

authorizations that suggest the reactor can be 

placed in operation. LCs that are being 

removed due to their one time applicability 

being previously satisfied have no bearing on 

future functions at CR–3. Other LCs are being 

removed that are not required by regulation 

for a permanently defueled and 

decommissioning plant. These changes 

cannot increase the probability or 

consequences of any accident that remains 

credible. The probability of occurrence of 

previously evaluated accidents is not 

increased, since extended operation in a 

defueled condition is the only operation 

currently allowed, and is therefore bounded 

by the existing analyses. Additionally, the 

occurrence of postulated accidents associated 

with reactor operation is no longer credible 

in a permanently defueled reactor. This 

significantly reduces the scope of applicable 

accidents. 

Therefore, the proposed amendment does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes have no impact on 

facility SSCs affecting the safe storage of 

irradiated fuel, or on the methods of 

operation of such SSCs, or on the handling 

and storage of irradiated fuel itself. The 

removal of ITS that are related only to the 

operation of the nuclear reactor or only to the 

prevention, diagnosis, or mitigation of 

reactor-related transients or accidents cannot 

result in different or more adverse failure 

modes or accidents than previously 

evaluated because the reactor is permanently 

shutdown and defueled, and CR–3 is no 

longer authorized to operate the reactor. 

The proposed deletion of requirements of 

the CR–3 ITS do not affect safe storage of 

nuclear fuel. The proposed PDTS continue to 

require proper control and monitoring of 

safety significant parameters. The proposed 

restriction on the fuel pool level is fulfilled 

by normal operating conditions and 

preserves initial conditions assumed in the 

analyses of the postulated DBA. The spent 

fuel pool water level, spent fuel pool boron 

concentration, and spent fuel pool storage 

LCOs are retained to preserve the current 

requirements for safe storage of irradiated 

fuel. 

The proposed amendment does not result 

in any new mechanisms that could initiate 

damage to the remaining relevant safety 

barriers for defueled plants (i.e., fuel 

cladding and spent fuel cooling). Since 

extended operation in a defueled condition is 

the only operation currently allowed, and 

therefore bounded by the existing analyses, 

such a condition does not create the 

possibility of a new or different kind of 

accident. 

Therefore, the proposed amendment does 

not create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Because the 10 CFR Part 50 license for CR– 

3 no longer authorizes operation of the 

reactor or emplacement or retention of fuel 

into the reactor vessel, as specified in 10 CFR 

50.82(a)(2), the occurrence of postulated 

accidents associated with reactor operation 

are no longer credible. The only remaining 

credible accident is a FHA. The proposed 

amendment does not adversely affect the 

inputs or assumptions of any of the design 

basis analyses that impact a FHA. 

The proposed changes are limited to those 

portions of the LCs and ITS that are not 

related to the safe storage of irradiated fuel. 

The requirements for SSCs that have been 

deleted from the CR–3 ITS are not credited 

in the existing accident analysis for the 

remaining applicable postulated accident; 

and as such, do not contribute to the margin 

of safety associated with the accident 

analysis. Postulated DBAs involving the 

reactor are no longer possible because the 

reactor is permanently shutdown and 

defueled and CR–3 is no longer authorized to 

operate the reactor. 

Therefore, the proposed amendment does 

not involve a significant reduction in a 

margin of safety because the current design 

limits continue to be met for the accident of 

concern. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: Lara S. Nichols, 

550 South Tryon Street, Charlotte NC 

28202. 

NRC Branch Chief: Douglas A. 

Broaddus. 

Entergy Nuclear Indian Point 3, LLC, 

Docket No. 50–286, Indian Point, Unit 3, 

Westchester County, NY 

Date of amendment request: April 1, 

2014. A publicly-available version is in 

ADAMS under Accession No. 

ML14099A333. 

Description of amendment request: 

The amendment would revise Technical 

Specifications (TS) Figure 3.4.3–1, 

Heatup Limitations for Reactor Coolant 

System, Figure 3.4.3–2, Cooldown 

Limitations for Reactor Coolant System, 

and Figure 3.4.3–3, Hydrostatic and 

Inservice Leak Testing Limitations for 

Reactor Coolant System, to indicate that 

the curves are applicable for vacuum 

fill. 

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. Operation of the facility in accordance 

with the proposed amendment would not 

involve a significant increase in the 

probability of occurrence or consequences of 

an accident previously evaluated. 

The proposed TS changes do not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. There are no physical changes to 

the plant being introduced by the proposed 

changes to the heatup, cooldown and 

hydrostatic inservice leak testing limitation 

curves. The proposed changes do not modify 

the RCS pressure boundary. That is, there are 

no changes in operating pressure, materials, 

or seismic loading. The proposed changes do 

not adversely affect the integrity of the RCS 

pressure boundary such that its function in 

the control of radiological consequences is 

affected. The heatup, cooldown and 

hydrostatic inservice leak testing limitation 

curves were established in compliance with 

the methodology used to calculate and 

predict effects of radiation on embrittlement 

of RPV beltline materials and remain valid 

during vacuum fill. 

Consequently, the proposed changes do not 

involve a significant increase in the 

probability or the consequences of an 

accident previously evaluated. 

2. Operation of the facility in accordance 

with the proposed amendment would not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

The proposed TS changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. No new modes of operation are 

introduced by the proposed changes. The 

proposed changes will not create any failure 

mode not bounded by previously evaluated 

accidents. 

Consequently, the proposed changes do not 

create the possibility of a new or different 

kind of accident, from any accident 

previously evaluated. 

3. Operation of the facility in accordance 

with the proposed amendment would not 

involve a significant reduction in the margin 

of safety. 

The proposed TS changes do not involve 

a significant reduction in the margin of 

safety. The changes clarify that the heatup, 

cooldown and hydrostatic inservice leak 

testing limitation curves remain valid during 

vacuum fill (to 0 psia) in accordance with 

current regulations. Because operation will 

be within these limits, the RCS materials will 

continue to behave in a non-brittle manner 

consistent with the original design bases. 

Therefore, Entergy has concluded that 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 


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proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jeanne Cho, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc., 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Branch Chief: Benjamin Beasley. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and STN 50– 

457, Braidwood Station, Units 1 and 2, 

Will County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–454 and STN 50– 

455, Byron Station, Units 1 and 2, Ogle 

County, Illinois 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station, Unit 1, DeWitt County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, LaSalle 

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Units 1 

and 2, Rock Island County, Illinois 

Date of amendment request: July 14, 

2014. A publicly-available version is in 

ADAMS under Accession No. 

ML14195A172. 

Description of amendment request: 

The proposed amendment would revise 

and add technical specification (TS) 

surveillance requirements to address the 

concerns discussed in NRC Generic 

Letter 2008–01, ‘‘Managing Gas 

Accumulation in Emergency Core 

Cooling, Decay Heat Removal, and 

Containment Spray Systems,’’ dated 

January 11, 2008 (ADAMS Accession 

No. ML072910759). The proposed TS 

changes are based on NRC-approved TS 

Task Force (TSTF) Traveler TSTF–523, 

Revision 2, ‘‘Generic Letter 2008–01, 

Managing Gas Accumulation,’’ dated 

February 21, 2013 (ADAMS Accession 

No. ML13053A075). The NRC staff 

issued a Notice of Availability for 

TSTF–523, Revision 2, for plant-specific 

adoption using the Consolidated Line 

Item Improvement Process, in the 

Federal Register on January 15, 2014 

(79 FR 2700). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises or adds 

Surveillance Requirements (SRs) that require 

verification that the Emergency Core Cooling 

System (ECCS), the Decay Heat Removal 

(DHR)/Residual Heat Removal (RHR)/

Shutdown Cooling (SDC) System, the 

Containment Spray (CS) System, and the 

Reactor Core Isolation Cooling (RCIC) 

System, as applicable, are not rendered 

inoperable due to accumulated gas and to 

provide allowances which permit 

performance of the revised verification. Gas 

accumulation in the subject systems is not an 

initiator of any accident previously 

evaluated. As a result, the probability of any 

accident previously evaluated is not 

significantly increased. The proposed SRs 

ensure that the subject systems continue to 

be capable to perform their assumed safety 

function and are not rendered inoperable due 

to gas accumulation. Thus, the consequences 

of any accident previously evaluated are not 

significantly increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change revises or adds SRs 

that require verification that the ECCS, the 

DHR/RHR/SDC System, the CS System, and 

the RCIC System, as applicable, are not 

rendered inoperable due to accumulated gas 

and to provide allowances which permit 

performance of the revised verification. The 

proposed change does not involve a physical 

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

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. In addition, the proposed 

change does not impose any new or different 

requirements that could initiate an accident. 

The proposed change does not alter 

assumptions made in the safety analysis and 

is consistent with the safety analysis 

assumptions. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

2. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change revises or adds SRs 

that require verification that the ECCS, the 

DHR/RHR/SDC System, the CS System, and 

the RCIC System, as applicable, are not 

rendered inoperable due to accumulated gas 

and to provide allowances which permit 

performance of the revised verification. The 

proposed change adds new requirements to 

manage gas accumulation in order to ensure 

the subject systems are capable of performing 

their assumed safety functions. The proposed 

SRs are more comprehensive than the current 

SRs and will ensure that the assumptions of 

the safety analysis are protected. The 

proposed change does not adversely affect 

any current plant safety margins or the 

reliability of the equipment assumed in the 

safety analysis. Therefore, there are no 

changes being made to any safety analysis 

assumptions, safety limits or limiting safety 

system settings that would adversely affect 

plant safety as a result of the proposed 

change. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: J. Bradley 

Fewell, Esquire, Vice President and 

Deputy General Counsel, Exelon 

Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Travis L. Tate. 

FirstEnergy Nuclear Operating 

Company, et al., Docket Nos. 50–334 

and 50–412, Beaver Valley Power 

Station (BVPS), Units 1 and 2, Beaver 

County, Pennsylvania 

Date of amendment request: 

September 4, 2014. A publicly-available 

version is available in ADAMS under 

Accession No. ML14247A512. 

Description of amendment request: 

The amendment proposes changes to 

align the BVPS Emergency Planning 

Zone (EPZ) boundary with the boundary 

that is currently in use by the 

emergency management agencies of the 

three counties that implement public 

protective actions around BVPS. 

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, along with NRC edits in square 

brackets: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This amendment request would alter 

portions of the outer EPZ boundary defined 

in the BVPS EPP [Emergency Preparedness 

Plan] to align with the EPZ boundaries 

implemented by the Columbiana County, 

Hancock County, and Beaver County 

emergency management agencies. The 

proposed amendment does not involve any 

modifications or physical changes to plant 

systems, structures, or components. The 

proposed amendment does not change plant 

operations or maintenance of plant systems, 

structures, or components. Nor does the 

proposed amendment alter any BVPS EPP 


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facility or equipment. Changing the EPZ 

boundaries cannot increase the probability of 

an accident since emergency plan functions 

would be implemented after a postulated 

accident occurs. The proposed amendment 

does not alter or prevent the ability of the 

BVPS emergency response organization to 

perform intended emergency plan functions 

to mitigate the consequences of and to 

respond adequately to radiological 

emergencies. 

Therefore, the proposed amendment does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

This amendment request alters the EPZ 

boundary described in the BVPS EPP. The 

proposed amendment does not involve any 

design modifications or physical changes to 

the plant, does not change plant operation or 

maintenance of equipment, and does not 

alter BVPS EPP facilities or equipment. The 

proposed amendment to the BVPS EPP does 

not alter any BVPS emergency actions that 

would be implemented in response to 

postulated accident events. The proposed 

amendment does not create any credible new 

failure mechanisms, malfunctions, or 

accident initiators not previously considered. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

This amendment request would alter 

portions of the EPZ boundary defined in the 

BVPS EPP. The proposed amendment does 

not involve any design or licensing basis 

functions of the plant, no physical changes 

to the plant are made, does not impact plant 

operation or maintenance of equipment, and 

does not alter BVPS EPP facilities or 

equipment. This change does not alter any 

BVPS emergency actions that would be 

implemented in response to postulated 

accident events. The BVPS EPP continues to 

meet 10 CFR 50.47 and 10 CFR 50, Appendix 

E requirements for emergency response. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David W. 

Jenkins, FirstEnergy Nuclear Operating 

Company, FirstEnergy Corporation, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: Meena K. Khanna. 

Florida Power and Light Company, et al. 

(the licensee), Docket Nos. 50–335 and 

50–389, St. Lucie Plant, Units 1 and 2, 

St. Lucie County, Florida 

Date of amendment request: July 14, 

2014. A publicly-available version is in 

ADAMS under Accession No. 

ML14198A074. 

Description of amendment request: 

The amendments would revise or add 

surveillance requirements (SRs) to 

verify that the system locations 

susceptible to gas accumulation are 

sufficiently filled with water and to 

provide allowances that permit 

performance of the verification. The 

licensee proposed the changes to 

address NRC Generic Letter 2008–01, 

‘‘Managing Gas Accumulation in 

Emergency Core Cooling, Decay Heat 

Removal, and Containment Spray 

Systems’’ (ADAMS Accession No. 

ML072910759), as described in Revision 

2 of Technical Specification Task Force 

No. 523, ‘‘Generic Letter 2008–01, 

Managing Gas Accumulation’’ (ADAMS 

Accession No. ML13053A075). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the Proposed Change Involve a 

Significant Increase in the Probability or 

Consequences of an Accident Previously 

Evaluated? 

Response: No. 

The proposed change revises or adds SRs 

that require verification that the Emergency 

Core Cooling Systems (ECCS), Residual Heat 

Removal (RHR) System, Shutdown Cooling 

(SDC) System, and Containment Spray (CS) 

System are not rendered inoperable due to 

accumulated gas and to provide allowances 

which permit performance of the revised 

verification. Gas accumulation in the subject 

systems is not an initiator of any accident 

previously evaluated. As a result, the 

probability of any accident previously 

evaluated is not significantly increased. The 

proposed SRs ensure that the subject systems 

continue to be capable of performing their 

assumed safety function and are not rendered 

inoperable due to gas accumulation. Thus, 

the consequences of any accident previously 

evaluated are not significantly increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the Proposed Change Create the 

Possibility of a New or Different Kind of 

Accident from any Accident Previously 

Evaluated? 

Response: No. 

The proposed change revises or adds SRs 

that require verification that the ECCS, RHR 

System, SDC System, and CS System are not 

rendered inoperable due to accumulated gas 

and to provide allowances which permit 

performance of the revised verification. The 

proposed change does not involve a physical 

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

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. In addition, the proposed 

change does not impose any new or different 

requirements that could initiate an accident. 

The proposed change does not alter 

assumptions made in the safety analysis and 

is consistent with the safety analysis 

assumptions. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the Proposed Change Involve a 

Significant Reduction in a Margin of Safety? 

Response: No. 

The proposed change revises or adds SRs 

that require verification that the ECCS, RHR 

System, SDC System, and CS System are not 

rendered inoperable due to accumulated gas 

and to provide allowances which permit 

performance of the revised verification. The 

proposed change adds new requirements to 

manage gas accumulation in order to ensure 

that the subject systems are capable of 

performing their assumed safety functions. 

The proposed SRs are more comprehensive 

than the current SRs and will ensure that the 

assumptions of the safety analysis are 

protected. The proposed change does not 

adversely affect any current plant safety 

margins or the reliability of the equipment 

assumed in the safety analysis. Therefore, 

there are no changes being made to any safety 

analysis assumptions, safety limits, or 

limiting safety system settings that would 

adversely affect plant safety as a result of the 

proposed change. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: William S. 

Blair, Managing Attorney—Nuclear, 

Florida Power & Light, 700 Universe 

Blvd., MS LAW/JB, Juno Beach, Florida 

33408–0420. 

NRC Acting Branch Chief: Lisa M. 

Regner. 

Florida Power and 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: August 8, 

2014. A publicly-available version is in 

ADAMS under Accession No. 

ML14225A654. 

Description of amendment request: 

The amendments would modify the 

Technical Specifications by removing 

TS 3/4.4.7, ‘‘Chemistry,’’ which 

provides limits on the oxygen, chloride, 

and fluoride content in the reactor 

coolant system to minimize corrosion. 


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The licensee requested that these 

requirements be relocated to the 

Updated Final Safety Analysis Report 

(UFSAR) and controlled in accordance 

with 10 CFR 50.59, ‘‘Changes, tests, and 

experiments.’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee provided its analysis of the 

issue of no significant hazards 

consideration, which is presented as 

follows: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change acts to remove 

current Reactor Coolant System (RCS) 

chemistry limits and monitoring 

requirements from the TS and relocate the 

requirements to the UFSAR. Monitoring and 

maintaining RCS chemistry minimizes the 

potential for corrosion of RCS piping and 

components. Corrosion effects are considered 

a long-term impact on RCS structural 

integrity. Because RCS chemistry will 

continue to be monitored and controlled, 

relocating the current TS requirements to the 

UFSAR will not present an adverse impact to 

the RCS and subsequently, will not impact 

the probability or consequences of an 

accident previously evaluated. Furthermore, 

once relocated to the UFSAR, changes to RCS 

chemistry limits and monitoring 

requirements will be controlled in 

accordance with 10 CFR 50.59. 

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 accident previously 

evaluated? 

Response: No. 

The proposed change acts to remove 

current Reactor Coolant System (RCS) 

chemistry limits and monitoring 

requirements from the TS and relocate the 

requirements to the UFSAR. The proposed 

change does not introduce new modes of 

plant operation and it does not involve 

physical modifications to the plant (no new 

or different type of equipment will be 

installed). There are no changes in the 

method by which any safety related plant 

structure, system, or component (SSC) 

performs its specified safety function. As 

such, the plant conditions for which the 

design basis accident analyses were 

performed remain valid. 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures will be introduced as a result 

of the proposed change. There will be no 

adverse effect or challenges imposed on any 

SSC as a result of the proposed change. 

Therefore, the proposed 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. 

Margin of safety is related to confidence in 

the ability of the fission product barriers to 

perform their accident mitigation functions. 

The proposed change acts to remove current 

Reactor Coolant System (RCS) chemistry 

limits and monitoring requirements from the 

TS and relocate the requirements to the 

UFSAR. The proposed change will maintain 

limits on RCS chemistry parameters and will 

continue to provide associated monitoring 

requirements. The proposed change does not 

physically alter any SSC. There will be no 

effect on those SSCs necessary to assure the 

accomplishment of protection functions. 

There will be no impact on the overpower 

limit, departure from nucleate boiling ratio 

(DNBR) limits, loss of cooling accident peak 

cladding temperature (LOCA PCT), or any 

other margin of safety. The applicable 

radiological dose consequence acceptance 

criteria will continue to be met. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff reviewed the licensee’s 

analysis and, based on this review, it 

appears that the three standards of 

50.92(c) are satisfied. Therefore, the 

NRC staff proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: William S. 

Blair, Managing Attorney—Nuclear, 

Florida Power & Light Company, 700 

Universe Blvd. MS LAW/JB, Juno 

Beach, Florida 33408–0420. 

NRC Acting Branch Chief: Lisa M. 

Regner. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Nuclear Power Plant 

(CPNPP), Units 1 and 2, Somervell 

County, Texas 

Date of amendment request: July 1, 

2014. A publicly-available version is in 

ADAMS under Accession No. 

ML14192A338. 

Description of amendment request. 

The amendments would revise 

Technical Specification (TS) 3.8.1, ‘‘AC 

[Alternating Current] Sources— 

Operating,’’ to extend, on a one-time 

basis, the Completion Time (CT) of 

Required Action A.3 from 72 hours to 

14 days. By letter dated September 18, 

2013 (ADAMS Accession No. 

ML13232A143), the NRC staff issued 

Amendment No. 160 to Facility 

Operating License No. NPF–87 and 

Amendment No. 160 to Facility 

Operating License No. NPF–89 for 

CPNPP, Units 1 and 2, respectively. The 

amendments revised TS 3.8.1 to extend 

the CT for Required Action A.3 on a 

one-time basis from 72 hours to 14 days. 

The CT extension from 72 hours to 14 

days was to be used twice while 

completing the plant modification to 

install alternate startup transformer (ST) 

XST1A and was to expire on March 31, 

2014. 

The first 14-day CT was successfully 

completed, on October 14, 2013. 

However, the licensee inadvertently cut 

the wrong offsite power cable during the 

second 14-day CT resulting in a total 

loss-of-offsite power (LOOP) to both 

units and the modification had to be 

abandoned. Due to the cut-cable event 

and the subsequent efforts to determine 

the causes and corrective actions, the 

modification could not be completed by 

March 31, 2014. The licensee has 

requested an extension of the CT from 

72 hours to 14 days on one-time basis 

to complete the plant modification. 

Installation of the alternate ST XST1A 

will result in improved offsite power 

system reliability. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change will revise the CT for 

the loss of one offsite source from 72 hours 

to 14 days to allow a one-time, 14-day CT. 

The proposed one-time extension of the CT 

for the loss of one offsite power circuit does 

not significantly increase the probability of 

an accident previously evaluated. The TS 

will continue to require equipment that will 

power safety related equipment necessary to 

perform any required safety function. The 

one-time extension of the CT to 14 days does 

not affect the design of the STs, the interface 

of the STs with other plant systems, the 

operating characteristic of the STs, or the 

reliability of the STs. 

The consequence of a LOOP event has been 

evaluated in the CPNPP Final Safety Analysis 

Report (Reference 8.1 [of the licensee’s letter 

dated July 1, 2014]) and the Station Blackout 

evaluation. Increasing the CT for one offsite 

power source on a one-time basis from 72 

hours to 14 days does not increase the 

consequences of a LOOP event nor change 

the evaluation of LOOP events. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes 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 electrical 

distribution subsystems provide plant 

protection. The proposed change will only 

affect the time allowed to restore the 

operability of the offsite power source 

through a ST. The proposed change does not 

affect the configuration, or operation of the 


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plant. The proposed change to the CT will 

facilitate installation of a plant modification 

which will improve plant design and will 

eliminate the necessity to shut down both 

Units if XST1 fails or requires maintenance 

that goes beyond the current TS CT of 72 

hours. This change will improve the longterm reliability of the 138 [kiloVolt (kV)] 

offsite circuit ST which is common to both 

CPNPP Units. 

There are no changes to the STs or the 

supporting systems operating characteristics 

or conditions. The change to the CT does not 

change any existing accident scenarios, nor 

create any new or different accident 

scenarios. In addition, the change does not 

impose any new or different requirements or 

eliminate any existing requirements. The 

change does not alter any of 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 previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change does not affect the 

acceptance criteria for any analyzed event 

nor is there a change to any safety limit. The 

proposed change does not alter the manner 

in which safety limits, limiting safety system 

settings, or limiting conditions for operation 

are determined. Neither the safety analyses 

nor the safety analysis acceptance criteria are 

affected by this change. The proposed change 

will not result in plant operation in a 

configuration outside the current design 

basis. The proposed activity only increases a 

one-time pre-planned occurrence, the period 

when the plant may operate with one offsite 

power source. The margin of safety is 

maintained by maintaining the ability to 

safely shut down the plant and remove 

residual heat. 

Therefore, the proposed change does not 

involve a reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the standards of 

10 CFR 50.92(c) are satisfied. Therefore, 

the NRC staff proposes to determine that 

the request for amendments involves no 

significant hazards consideration. 

Attorney for licensee: Timothy P. 

Matthews, Esq., Morgan, Lewis and 

Bockius, 1111 Pennsylvania Avenue 

NW., Washington, DC 20004. 

NRC Branch Chief: Michael T. 

Markley. 

Northern States Power Company— 

Minnesota, Docket Nos. 50–282 and 50– 

306, Prairie Island Nuclear Generating 

Plant (PINGP), Units 1 and 2, Goodhue 

County, Minnesota 

Date of amendment request: August 

21, 2014. A publicly-available version is 

in ADAMS under Accession No. 

ML14233A431. 

Description of amendment request: 

The proposed amendments would 

revise the PINGP, Units 1 and 2, 

licensing basis analysis for waste gas 

decay tank rupture. 

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. 

This license amendment request proposes 

to revise the licensing basis waste gas decay 

tank rupture analysis. The proposed analysis 

was updated to include the current fuel type, 

current fuel cycle lengths and plant operation 

to sixty years. 

The proposed waste gas decay tank rupture 

analysis changes are not accident initiators, 

and therefore the proposed changes do not 

involve an increase in the probability of an 

accident. 

The original waste gas decay tank rupture 

analysis demonstrated that the doses were a 

small fraction of the regulatory guidelines 

and that the waste gas system design 

prevents release of undue amounts of 

radioactivity. The revised waste gas decay 

tank rupture analysis demonstrates that the 

doses are well within the regulatory 

guidelines and that the waste gas system 

design continues to prevent release of undue 

amounts of radioactivity, and thus the 

proposed changes do not involve a 

significant increase in the consequences of an 

accident. 

Therefore, the proposed licensing basis 

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. 

This license amendment request proposes 

to revise the licensing basis waste gas decay 

tank rupture analysis. The proposed analysis 

was updated to include the current fuel type, 

current fuel cycle lengths and plant operation 

to sixty years. 

The proposed waste gas decay tank rupture 

analysis includes plant changes that have 

previously been evaluated. This analysis 

applies the same methodology as the 

previous analysis. The proposed revision to 

the waste gas decay tank rupture analysis 

does not change any system operations or 

maintenance activities. The changes do not 

involve physical alteration of the plant; that 

is, no new or different type of equipment will 

be installed. These changes do not create new 

failure modes or mechanisms which are not 

identifiable during testing and no new 

accident precursors are generated. 

Therefore, the proposed licensing basis 

change does not create the possibility of a 

new or different kind of accident from any 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

This license amendment request proposes 

to revise the licensing basis waste gas decay 

tank rupture analysis. The proposed analysis 

was updated to include the current fuel type, 

current fuel cycle lengths and plant operation 

to sixty years. 

This revised analysis applies the same 

methodology as the original waste gas decay 

tank rupture analysis. The original waste gas 

decay tank rupture analysis demonstrated 

that the doses were a small fraction of the 

regulatory guidelines and that the waste gas 

system design prevents release of undue 

amounts of radioactivity. The revised waste 

gas decay tank rupture analysis demonstrates 

that the doses are well within the regulatory 

guidelines and that the waste gas system 

design continues to prevent release of undue 

amounts of radioactivity. 

Therefore, the proposed licensing basis 

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 requests involve no 

significant hazards consideration. 

Attorney for licensee: Peter M. Glass, 

Assistant General Counsel, Xcel Energy 

Services, Inc., 414 Nicollet Mall, 

Minneapolis, MN 55401 

NRC Branch Chief: David L. Pelton. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station (SNGS), Units 1 and 2, Salem 

County, New Jersey 

Date of amendment request: July 28, 

2014. A publicly-available version is in 

ADAMS under Accession No. 

ML14210A484. 

Description of amendment request: 

The proposed amendment would 

modify technical specification (TS) 

requirements regarding steam generator 

tube inspections and reporting as 

described in Technical Specification 

Task Force (TSTF) traveler TSTF–510, 

Revision 2, ‘‘Revision to Steam 

Generator Program Inspection 

Frequencies and Tube Sample 

Selection.’’ In addition, the proposed 

amendment would revise the SNGS, 

Unit 2, TSs 6.8.4.i, ‘‘Steam Generator 

(SG) Program,’’ TS 6.9.1.10, ‘‘Steam 

Generator Tube Inspection Report,’’ and 

the bases section of 3/4.4.6, ‘‘Steam 

Generator (SG) Tube Integrity,’’ to 

remove unnecessary information related 

to the original Salem Unit 2 

Westinghouse steam generators. 

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 


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below, along with NRC edits in square 

brackets: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change revises the Steam 

Generator (SG) Program to modify the 

frequency of verification of SG tube integrity 

and SG tube sample selection. A steam 

generator tube rupture (SGTR) event is one of 

the design basis accidents that are analyzed 

as part of a plant’s licensing basis. The 

proposed SG tube inspection frequency and 

sample selection criteria will continue to 

ensure that the SG tubes are inspected such 

that the probability of a SGTR is not 

increased. The consequences of a SGTR are 

bounded by the conservative assumptions in 

the design basis accident analysis. The 

proposed change will not cause the 

consequences of a SGTR to exceed those 

assumptions. 

Therefore, it is concluded that this change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

The proposed changes to the Salem Unit 2 

Technical Specifications (TS) that are not 

associated with TSTF–510, removing 

unnecessary information related to W* 

[pronounced ‘‘W star,’’ which refers to the 

length of the steam generator tube required 

to be inspected within the hot-leg tube sheet] 

that is only applicable to Westinghouse 

steam generators, is an administrative change 

that does not involve a significant increase in 

the probability or consequences of an 

accident previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes to the Steam 

Generator Program will not introduce any 

adverse changes to the plant design basis or 

postulated accidents resulting from potential 

tube degradation. The proposed change does 

not affect the design of the SGs or their 

method of operation. In addition, the 

proposed change does not impact any other 

plant system or component. 

The proposed changes to the Salem Unit 2 

Technical Specifications (TS) that are not 

associated with TSTF–510, removing 

unnecessary information related to W* that is 

only applicable to Westinghouse steam 

generators, is an administrative change that 

does not affect the design of the SGs or their 

method of operation. 

Therefore, it is concluded that these 

changes do not create the possibility of a new 

of different kind of accident from any 

accident previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The SG tubes in pressurized water reactors 

are an integral part of the reactor coolant 

pressure boundary and, as such, are relied 

upon to maintain the primary system’s 

pressure and inventory. As part of the reactor 

coolant pressure boundary, the SG tubes are 

unique in that they are also relied upon as 

a heat transfer surface between the primary 

and secondary systems such that residual 

heat can be removed from the primary 

system. In addition, the SG tubes also isolate 

the radioactive fission products in the 

primary coolant from the secondary system. 

In summary, the safety function of a SG is 

maintained by ensuring the integrity of its 

tubes. 

Steam generator tube integrity is a function 

of the design, environment, and the physical 

condition of the tube. The proposed change 

does not affect tube design or operating 

environment. The proposed change will 

continue to require monitoring of the 

physical condition of the SG tubes such that 

there will not be a reduction in the margin 

of safety compared to the current 

requirements. 

The proposed changes to the Salem Unit 2 

Technical Specifications (TS) that are not 

associated with TSTF–510, removing 

unnecessary information related to W* that is 

only applicable to Westinghouse steam 

generators, is an administrative change that 

does not involve a significant reduction in a 

margin of safety. 

Therefore, it is concluded that the 

proposed changes do not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jeffrie J. Keenan, 

PSEG Nuclear LLC—N21, P.O. Box 236, 

Hancocks Bridge, NJ 08038. 

NRC Branch Chief: Meena K. Khanna. 

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: June 20, 

2014, as supplemented by letter dated 

August 6, 2014. Publicly-available 

versions are in ADAMS under 

Accession Nos. ML14174B176 and 

ML14218A809. 

Description of amendment request: 

The proposed license amendment 

request (LAR) would revise the Updated 

Final Safety Analysis Report (UFSAR) 

in regard to Tier 2* information related 

to fire area boundaries. These changes 

add three new fire zones in the middle 

annulus to provide enclosures for the 

Class 1E electrical containment 

penetrations in accordance with UFSAR 

Appendix 9A, Subsection 9A.3.1.1.15. 

The addition of the three new fire zones 

extended the fire area boundaries for 

three existing fire areas and, therefore, 

constitutes a change to Tier 2* 

information. Additionally, the licensee 

proposed changes that require revisions 

to UFSAR Tier 2 information involving 

changes to plant-specific Tier 2* 

information. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed middle annulus fire barrier 

reconfiguration for the electrical penetrations 

would not adversely affect any safety-related 

equipment or function. The modified 

configuration for the Class 1E electrical 

containment penetration enclosures will 

maintain the fire protection function (i.e., 

barrier) as evaluated in Updated Final Safety 

Analysis Report (UFSAR), thus, the 

probability of a Class 1E electrical 

containment penetration failure is not 

significantly increased. The safe shutdown 

fire analysis is not affected, and the fire 

protection analysis results are not adversely 

affected. The proposed changes do not 

involve any accident, initiating event or 

component failure; thus, the probabilities of 

previously evaluated accidents are not 

affected. The maximum allowable leakage 

rate specified in the Technical Specifications 

is unchanged, and radiological material 

release source terms are not affected; thus, 

the radiological releases in the accident 

analyses are not affected. 

Therefore, the proposed amendment does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The addition of enclosures constructed of 

three-hour rated fire barriers to separate the 

fire zones in the middle annulus for the Class 

1E electrical penetration assemblies will 

maintain the fire protection function as 

evaluated in the UFSAR. The addition of the 

fire barriers does not affect the function of 

the Class 1E electrical containment 

penetrations or electrical penetration 

assemblies, and thus, does not introduce a 

new failure mode. The addition of the fire 

barriers does not create a new fault or 

sequence of events that could result in a 

radioactive material release. 

Therefore, the proposed amendment does 

not create the possibility of a new or different 

kind of accident. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The use of enclosures constructed of threehour rated fire barriers to separate the fire 

zones in the middle annulus for the Class 1E 

electrical penetration assemblies will 

maintain the fire protection function as 

evaluated in the UFSAR. The use of the fire 

barriers does not affect the ability of the Class 


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1E electrical containment penetrations, 

electrical penetration assemblies, or the 

containment to perform their design 

function. The Class 1E electrical containment 

penetrations and electrical penetration 

assemblies within the enclosures continue to 

comply with the existing design codes and 

regulatory criteria, and do not affect any 

safety limit. The use of fire barriers and 

enclosures to separate the Class 1E electrical 

penetration assemblies does not adversely 

affect any margin of safety. 

Therefore, the proposed amendment does 

not involve a significant reduction in the 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis, and based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ms. Kathryn M. 

Sutton, Morgan, Lewis & Bockius LLC, 

1111 Pennsylvania Avenue NW., 

Washington, DC 20004–2514. 

NRC Branch Chief: Lawrence J. 

Burkhart. 

STP Nuclear Operating Company, 

Docket Nos. 50–498 and 50–499, South 

Texas Project (STP), Units 1 and 2, 

Matagorda County, Texas 

Date of amendment request: May 15, 

2014, as supplemented by letter dated 

July 10, 2014. Publicly-available 

versions are in ADAMS under 

Accession Nos. ML14164A341 and 

ML14282A185, respectively. 

Description of amendment request: 

The amendment would update the 

Emergency Action Levels (EALs) used at 

STP, Units 1 and 2 from the current 

scheme based on Nuclear Management 

and Resources Council, Inc. 

(NUMARC)/Nuclear Environmental 

Studies Project (NESP) report 

NUMARC/NESP–007, Revision 2, 

‘‘Methodology for Development of 

Emergency Action Levels,’’ dated 

January 1992 (ADAMS Accession No. 

ML041120174), to the NRC-endorsed 

scheme contained in Nuclear Energy 

Institute (NEI) 99–01, Revision 6, 

‘‘Development of Emergency Action 

Levels for Non-Passive Reactors,’’ dated 

November 2012 (ADAMS Accession No. 

ML12326A805). The EAL scheme in NEI 

99–01, Revision 6 includes an EAL for 

Independent Spent Fuel Storage 

Installations (ISFSI), which is needed in 

order to implement dry cask storage 

operations at STP Units 1 and 2. 

Additionally, there are three EALs that 

require Spent Fuel Pool level 

instrument values which are designed to 

address lessons learned from the 

Fukushima Dai-ichi accident. 

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 

consequence of an accident previously 

evaluated? 

Response: No. 

The change revises the STPNOC [STP 

Nuclear Operating Company] Emergency 

Action Levels to be consistent with the NRC 

endorsed EAL scheme contained in NEI 99– 

01, Revision 6, ‘‘Methodology for 

Development of Emergency Action Levels,’’ 

but does not alter any of the requirements of 

the Operating License or the Technical 

Specifications. In addition to replacing the 

current STP EALs, the new EAL scheme 

includes an EAL related to the planned STP 

Independent Spent Fuel Storage Installation, 

and EALs related to planned changes to the 

Spent Fuel Pool level instrumentation that 

will address lessons learned from Fukushima 

Daiichi. The proposed change does not 

modify any plant equipment and does not 

impact any failure modes that could lead to 

an accident. Additionally, the proposed 

change has no effect on the consequences of 

any analyzed accident since the change does 

not affect any equipment related to accident 

mitigation. Based on this discussion, the 

proposed amendment does not increase the 

probability or consequence 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 change revises the STPNOC 

Emergency Action Levels to be consistent 

with the NRC endorsed EAL scheme 

contained in NEI 99–01, Revision 6, 

‘‘Methodology for Development of 

Emergency Action Levels,’’ but does not alter 

any of the requirements of the Operating 

License or the Technical Specifications. In 

addition to replacing the current STP EALs, 

the new EAL scheme includes an EAL related 

to the planned STP Independent Spent Fuel 

Storage Installation, and EALs related to 

planned changes to the Spent Fuel Pool level 

instrumentation that will address lessons 

learned from Fukushima Daiichi. The 

proposed change does not modify any plant 

equipment and there is no impact on the 

capability of the existing equipment to 

perform their intended functions. No system 

setpoints are being modified. No new failure 

modes are introduced by the proposed 

change. The proposed amendment does not 

introduce any accident initiators or 

malfunctions that would cause a new or 

different kind of accident. 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 change revises the STPNOC 

Emergency Action Levels to be consistent 

with the NRC endorsed EAL scheme 

contained in NEI 99–01, Revision 6, 

‘‘Methodology for Development of 

Emergency Action Levels,’’ but does not alter 

any of the requirements of the Operating 

License or the Technical Specifications. In 

addition to replacing the current STP EALs, 

the new EAL scheme includes an EAL related 

to the planned STP Independent Spent Fuel 

Storage Installation, and EALs related to 

planned changes to the Spent Fuel Pool level 

instrumentation that will address lessons 

learned from Fukushima Daiichi. The 

proposed change does not affect any of the 

assumptions used in the accident analysis, 

nor does it affect any operability 

requirements for equipment important to 

plant safety. Therefore, the proposed change 

will not result in a significant reduction in 

the margin of safety in operation of the 

facility as discussed in this license 

amendment request. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the standards of 

10 CFR 50.92(c) are satisfied. Therefore, 

the NRC staff proposes to determine that 

the request for amendments involves no 

significant hazards consideration. 

Attorney for licensee: A. H. 

Gutterman, Esq., Morgan, Lewis & 

Bockius, 1111 Pennsylvania Avenue 

NW., Washington, DC 20004. 

NRC Branch Chief: Michael T. 

Markley. 

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 


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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. 

50–261, H. B. Robinson Steam Electric 

Plant (HBRSEP), Unit 2, Darlington 

County, South Carolina 

Date of application for amendment: 

September 30, 2013, as supplemented 

by letter dated August 6, 2014. 

Brief description of amendment: The 

amendment implements the NRCapproved Technical Specification Task 

Force (TSTF) Standard Technical 

Specification Change Traveler, TSTF– 

491, ‘‘Removal of Main Steam and Main 

Feedwater Valve Isolation Times from 

Technical Specifications,’’ via the 

Consolidated Line Item Improvement 

Process. This amendment modifies the 

current Technical Specifications (TSs) 

3.7.2, Main Steam Isolation Valves and 

3.7.3, Main Feedwater Isolation Valves, 

Main Feedwater Regulation Valves and 

Bypass Valves by relocating the specific 

isolation time for the isolation valves 

from the associated Surveillance 

Requirements (SRs). The isolation time 

in the TS SRs is replaced with the 

requirement to verify the valve isolation 

time is ‘‘within limits.’’ The specific 

isolation times will be maintained in the 

HBRSEP, Unit 2, Technical 

Requirements Manual. 

Date of issuance: October 10, 2014. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days of issuance. 

Amendment No.: 237. A publiclyavailable version is in ADAMS under 

Accession No. ML14252A221; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. DPR–23. Amendment revised the 

Facility Operating License and TSs. 

Date of initial notice in Federal 

Register: December 10, 2013 (78 FR 

74180). The supplemental letter dated 

August 6, 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 October 10, 

2014. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–333, James A. FitzPatrick 

Nuclear Power Plant, Oswego County, 

New York 

Date of amendment request: May 1, 

2014, as supplemented by letter dated 

August 21, 2014. 

Brief description of amendment: The 

amendment revises Technical 

Specification 2.0, ‘‘Safety Limits (SLs),’’ 

by changing the safety limit minimum 

critical power ratio for both single and 

dual recirculation loop operation. 

Date of issuance: September 30, 2014. 

Effective date: As of the date of 

issuance, and shall be implemented 

within [30] days. 

Amendment No.: 307. A publiclyavailable version is in ADAMS under 

Accession No. ML14258B201; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. DPR–59: The amendment revised 

the License and the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 5, 2014 (79 FR 45487). 

The supplemental letter dated August 

21, 2014, provided additional 

information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the NRC staff’s 

original proposed no significant hazards 

consideration (NSHC) determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment and final NSHC 

determination is contained in a Safety 

Evaluation dated September 30, 2014. 

No significant hazards consideration 

comments received. No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, LaSalle 

County Station (LSCS), Units 1 and 2, 

LaSalle County, Illinois 

Date of amendment request: 

September 20, 2013, as supplemented 

by letter dated June 30, 2014. 

Brief description of amendment: The 

amendment revised the LSCS, Units 1 

and 2, allowable values for the loss of 

voltage relay voltage setpoints in 

Technical Specification Table 3.3.8.1–1, 

‘‘Loss of Power Instrumentation.’’ 

Date of issuance: September 29, 2014. 

Effective date: As of the date of 

issuance. For LSCS Unit 1, the 

amendment shall be implemented prior 

to entering MODE 4 following the spring 

2016 refueling outage (L1R16). For 

LSCS, Unit 2, the amendment shall be 

implemented prior to entering MODE 4 

following the spring 2015 refueling 

outage (L2R15). 

Amendment No.: 209 and 196. A 

publicly-available version is in ADAMS 

under Accession No. ML14252A913; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Facility Operating License Nos. NPF– 

11 and NPF–18: Amendment revised the 

Facility Operating License and 

Technical Specifications. 

Date of initial notice in Federal 

Register: December 10, 2013 (78 FR 

74182). The supplemental letter dated 

June 30, 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 September 29, 

2014. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket No. 50–289, Three Mile Island 

Nuclear Station, Unit (TMI–1), Dauphin 

County, Pennsylvania 

Date of application for amendment: 

May 7, 2014. 

Brief description of amendment: The 

amendment revises Technical 

Specification (TS) Surveillance 

Requirements (SRs) 4.12.1, ‘‘Emergency 

Control Room Air Treatment System,’’ 

and 4.12.4, ‘‘Fuel Handling Building 

[Engineered Safety Feature] ESF Air 

Treatment System.’’ The amendment 

revised the TSs to replace the existing 

SRs to operate ventilation systems with 

charcoal filters for a 10-hour period at 

a frequency controlled in accordance 

with the Surveillance Frequency 

Control Program (SFCP) with a 

requirement to operate the systems for 

greater than or equal to 15 continuous 

minutes at a frequency controlled in 

accordance with the SFCP. These 

changes are consistent with Technical 

Specification Task Force (TSTF) 

Traveler TSTF–522, Revision 0, ‘‘Revise 

Ventilation System Surveillance 

Requirements to Operate for 10 hours 

per Month.’’ 

Date of issuance: October 14, 2014. 


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Federal Register / Vol. 79, No. 208 / Tuesday, October 28, 2014 / Notices 64231 

Effective date: As of its date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 282. A publiclyavailable version is in ADAMS under 

Accession No. ML14240A348; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Renewed Facility Operating License 

No. DPR–50: Amendment revised the 

license and the technical specifications. 

Date of initial notice in Federal 

Register: August 5, 2014 (79 FR 45476). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 14, 

2014. 

No significant hazards consideration 

comments received: No. 

Indiana Michigan Power Company, 

Docket Nos. 50–315 and 50–316, Donald 

C. Cook Nuclear Plant (CNP), Units 1 

and 2, Berrien County, Michigan 

Date of amendment requests: April 9, 

2014, as supplemented by letter dated 

August 15, 2014. 

Brief description of amendments: The 

amendments revised the CNP Technical 

Specifications (TSs) 3.4.3, ‘‘[Reactor 

Coolant System] RCS Pressure and 

Temperature Limits.’’ The changes to 

TSs clarify that pressure limits are 

considered to be met for pressures that 

are below 0 psig (i.e., up to and 

including full vacuum conditions). 

Vacuum fill operations for the RCS can 

result in system pressures below 0 psig. 

Date of issuance: October 1, 2014. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment Nos.: 323 (Unit 1) and 

306 (Unit 2). A publicly-available 

version is in ADAMS under Accession 

No. ML14259A549; documents related 

to these amendments are listed in the 

Safety Evaluation enclosed with the 

amendments. 

Facility Operating License Nos. DPR– 

58 and DPR–74: Amendments revise the 

Facility Operating Licenses and 

Technical Specifications. 

Date of initial notice in Federal 

Register: July 8, 2014 (79 FR 38591). 

The supplemental letter dated August 

15, 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 October 1, 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: April 18, 

2014, as supplemented by the letter 

dated July 30, 2014. 

Brief description of amendment: The 

license amendment revises the Updated 

Final Safety Analysis Report (UFSAR) 

in regard to Tier 2* information related 

to fire area boundaries. These changes 

add three new fire zones in the middle 

annulus to provide enclosures for the 

Class 1E electrical containment 

penetrations in accordance with UFSAR 

Appendix 9A, Subsection 9A.3.1.1.15. 

Additionally, the license amendment 

revises UFSAR Tier 2 information 

involving changes to plant-specific Tier 

2* information. 

Date of issuance: October 8, 2014. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 25. A publiclyavailable version is in ADAMS under 

Accession No. ML14248A243; 

documents related to this amendment 

are listed in the Safety Evaluation 

enclosed with the amendment. 

Facility Combined Licenses No. NPF– 

91 and NPF–92: Amendment revised the 

Facility Combined Licenses. 

Date of initial notice in Federal 

Register: May 27, 2014 (79 FR 30189). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 8, 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 

27, 2014, as supplemented by the letter 

dated July 23, 2014. 

Brief description of amendment: The 

amendment revises the VEGP Units 3 

and 4 Emergency Plan and changes the 

combined licenses (COL), Appendix C, 

plant-specific emergency planning 

inspections, tests, analyses, and 

acceptance criteria (ITAAC) to reflect 

the relocation of the Operations Support 

Centers and changes the description of 

the plant monitoring system. 

Date of issuance: October 7, 2014 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 24. A publiclyavailable version is in ADAMS under 

Accession No. ML14245A075; 

documents related to these amendments 

are listed in the Safety Evaluation 

enclosed with the amendments. 

Facility Combined Licenses Nos. NPF– 

91 and NPF–92: Amendment revised the 

Facility Combined Licenses. 

Date of initial notice in Federal 

Register: May 13, 2014 (79 FR 27345). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated October 7, 2014. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket 

Nos. 50–327 and 50–328, Sequoyah 

Nuclear Plant, Units 1 and 2, Hamilton 

County, Tennessee 

Date of amendment request: April 29, 

2014. A redacted version was provided 

by letter dated May 27, 2014. 

Brief description of amendment: The 

amendments revised the Cyber Security 

Plan Implementation Milestone No. 8 

completion date and the physical 

protection license condition. 

Date of issuance: September 29, 2014. 

Effective date: As of its date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: Unit 1–333 and 

Unit 2–326. A publicly-available version 

is in ADAMS under Accession No. 

ML14245A179; documents related to 

these amendments are listed in the 

Safety Evaluation enclosed with the 

amendments. 

Facility Operating License Nos. DPR– 

77 and DPR–79. The amendments 

revised the Operating License. 

Date of initial notice in Federal 

Register: July 8, 2014 (79 FR 38582). 

The supplemental letter dated May 27, 

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 September 29, 

2014. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket 

Nos. 50–327 and 50–328, Sequoyah 

Nuclear Plant (SQN), Units 1 and 2, 

Hamilton County, Tennessee 

Date of amendment request: April 29, 

2014. A redacted version was provided 

by letter dated May 27, 2014. 

Brief description of amendment: The 

amendments revised the Cyber Security 

Plan Implementation Milestone No. 8 

completion date and the physical 

protection license condition. 


asabaliauskas on DSK4SPTVN1PROD with NOTICES
64232 Federal Register / Vol. 79, No. 208 / Tuesday, October 28, 2014 / Notices 

Date of issuance: September 29, 2014. 

Effective date: As of its date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: Unit 1–333 and 

Unit 2–326. A publicly-available version 

is in ADAMS under Accession No. 

ML14245A179; documents related to 

these amendments are listed in the 

Safety Evaluation enclosed with the 

amendments. 

Facility Operating License Nos. DPR– 

77 and DPR–79. The amendments 

revised the Operating License. 

Date of initial notice in Federal 

Register: July 8, 2014 (79 FR 38582). 

The supplemental letter dated May 27, 

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 September 29, 

2014. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket No. 

50–390 Watts Bar Nuclear Plant, Unit 1, 

Rhea County, Tennessee 

Date of amendment request: April 29, 

2014, as supplemented by letter dated 

May 27, 2014. 

Brief description of amendment: The 

amendment revised the Cyber Security 

Plan Implementation Milestone No. 8 

completion date and the physical 

protection license condition. 

Date of issuance: September 29, 2014. 

Effective date: As of its date of 

issuance and shall be implemented 

within 30 days of issuance. 

Amendment No.: 97. A publiclyavailable version is in ADAMS under 

Accession No. ML14255A152; 

documents related to this amendment 

are listed in the Safety Evaluation (SE) 

enclosed with the amendment. 

Facility Operating License No. NPF– 

90. Amendment revised the Operating 

License. 

Date of initial notice in Federal 

Register: July 8, 2014 (79 FR 38581). 

The supplemental letter dated May 27, 

2014, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in the SE 

dated September 29, 2014. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 16th day 

of October 2014. 

For the Nuclear Regulatory Commission. 

Michele G. Evans, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2014–25357 Filed 10–27–14; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2014–0001] 

Sunshine Act Meeting Notice 

DATE: Weeks of October 27, November 3, 

10, 17, 24, December 1, 2014. 

PLACE: Commissioners’ Conference 

Room, 11555 Rockville Pike, Rockville, 

Maryland. 

STATUS: Public and Closed. 

Week of October 27, 2014 

Wednesday, October 29, 2014 

9:00 a.m. Briefing on Security Issues 

(Closed—Ex. 1) 

1:30 p.m. Briefing on Security Issues 

(Closed—Ex. 1) 

Thursday, October 30, 2014 

9:00 a.m. Briefing on Watts Bar Unit 2 

License Application Review (Public 

Meeting) (Contact: Justin Poole, 

301–415–2048) 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov/ . 

Week of November 3, 2014—Tentative 

Wednesday, November 5, 2014 

1:00 p.m. Briefing on Small Modular 

Reactors (Public Meeting) (Contact: 

Rollie D. Berry, III, 301–415–8162) 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov/ . 

Week of November 10, 2014—Tentative 

Thursday, November 13, 2014 

9:30 a.m. Strategic Programmatic 

Overview of the Nuclear Material 

Users and the Fuel Facilities 

Business Lines (Public Meeting) 

(Contact: Cinthya Roman, 301–287– 

9091) 

This meeting will be webcast live at 

the Web address—http://www.nrc.gov/ . 

1:30 p.m. Discussion of Management 

and Personnel Issues (Closed—Ex. 2 

and 6) 

Week of November 17, 2014—Tentative 

Thursday, November 20, 2014 

9:30 a.m. Briefing on Project Aim 2020 

(Closed—Ex. 2) 

Week of November 24, 2014—Tentative 

There are no meetings scheduled for 

the week of November 24, 2014. 

Week of December 1, 2014—Tentative 

There are no meetings scheduled for 

the week of December 1, 2014. 

* * * * * 

The schedule for Commission 

meetings is subject to change on short 

notice. For more information or to verify 

the status of meetings, contact Glenn 

Ellmers at (301) 415–0442 or via email 

at Glenn.Ellmers@nrc.gov. 

* * * * * 

The NRC Commission Meeting 

Schedule can be found on the Internet 

at: http://www.nrc.gov/public-involve/

public-meetings/schedule.html. 

* * * * * 

The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g. 

braille, large print), please notify 

Kimberly Meyer, NRC Disability 

Program Manager, at 301–287–0727, by 

videophone at 240–428–3217, or by 

email at Kimberly.Meyer-Chambers@

nrc.gov . Determinations on requests for 

reasonable accommodation will be 

made on a case-by-case basis. 

* * * * * 

Members of the public may request to 

receive this information electronically. 

If you would like to be added to the 

distribution, please contact the Office of 

the Secretary, Washington, DC 20555 

(301–415–1969), or send an email to 

Patricia.Jimenez@nrc.gov or 

Brenda.Akstulewicz@nrc.gov. 

Dated: October 23, 2014. 

Glenn Ellmers, 

Policy Coordinator, Office of the Secretary. 

[FR Doc. 2014–25721 Filed 10–24–14; 4:15 pm] 

BILLING CODE 7590–01–P 


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