Federal Register / Vol. 75, No. 192 / Tuesday, October 5, 2010 / Notices 61521 

Activity for Which Permit is Requested 

Take, Export from USA, Introduce 

non-indigenous species into Antarctica, 

and Import into USA. The applicant 

plans to collect water samples 

containing marine microbes (algae and 

protozoa) for use in experiments, for 

preservation for future examination, and 

for extraction of nucleic acids for 

diversity and abundance analyses back 

at the home institution. Live cultures of 

marine bacteria, previously collected 

from Antarctic waters, will be used in 

shipboard experiments to study feeding 

rates and transfer of nutrients in 

Antarctic protistan grazers. All 

remaining live cultures will be 

autoclaved before disposal. 

Location 

Ross Sea region, Antarctica. 

Dates 

January 1, 2011 to April 1, 2011. 

Nadene G. Kennedy, 

Permit Officer, Office of Polar Programs. 

[FR Doc. 2010–24865 Filed 10–4–10; 8:45 am] 

BILLING CODE 7555–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0316] 

NUREG/CR–7010, Cable Heat Release, 

Ignition, and Spread in Tray 

Installations During Fire 

(CHRISTIFIRE); Volume 1: Horizontal 

Trays, Draft Report for Comment 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Announcement of issuance for 

public comment, availability. 

SUMMARY: The Nuclear Regulatory 

Commission has issued for public 

comment a document entitled: ‘‘NUREG/ 

CR–7010, Cable Heat Release, Ignition, 

and Spread in Tray Installations During 

Fire (CHRISTIFIRE) Volume 1: 

Horizontal Trays, Draft Report for 

Comment.’’ 

DATES: Please submit comments by 

November 15, 2010. Comments received 

after this date will be considered if it is 

practical to do so, but the NRC staff is 

able to ensure consideration only for 

comments received on or before this 

date. 

ADDRESSES: You may submit comments 

by any one of the following methods. 

Please include Docket ID NRC–2010– 

0316 in the subject line of your 

comments. Comments submitted in 

writing or in electronic form will be 

posted on the NRC website and on the 

Federal rulemaking Web site 

Regulations.gov. Because your 

comments will not be edited to remove 

any identifying or contact information, 

the NRC cautions you against including 

any information in your submission that 

you do not want to be publicly 

disclosed. 

The NRC requests that any party 

soliciting or aggregating comments 

received from other persons for 

submission to the NRC inform those 

persons that the NRC will not edit their 

comments to remove any identifying or 

contact information, and therefore, they 

should not include any information in 

their comments that they do not want 

publicly disclosed. 

Federal Rulemaking Web site: Go to 

http://www.regulations.gov and search 

for documents filed under Docket ID 

NRC–2010–0316. Address questions 

about NRC dockets to Carol Gallagher 

301–492–3668; e-mail 

Carol.Gallagher@nrc.gov. 

Mail comments to: Cindy K. Bladey, 

Chief, Rules, Announcements and 

Directives Branch (RADB), Division of 

Administrative Services, Office of 

Administration, Mail Stop: TWB–05– 

B01M, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, or by fax to RADB at (301) 492– 

3446. 

You can access publicly available 

documents related to this notice using 

the following methods: 

NRC’s Public Document Room (PDR): 

The public may examine and have 

copied for a fee publicly available 

documents at the NRC’s PDR, Public 

File Area O1 F21, One White Flint 

North, 11555 Rockville Pike, Rockville, 

Maryland. 

NRC’s Agencywide Documents Access 

and Management System (ADAMS): 

Publicly available documents created or 

received at the NRC are available 

electronically at the NRC’s Electronic 

Reading Room at http://www.nrc.gov/ 

reading-rm/adams.html. From this page, 

the public can gain entry into ADAMS, 

which provides text and image files of 

NRC’s public documents. If you do not 

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the NRC’s 

PDR reference staff at 1–800–397–4209, 

301–415–4737, or by e-mail to 

pdr.resource@nrc.gov. ‘‘NUREG/CR– 

7010, Cable Heat Release, Ignition, and 

Spread in Tray Installations During Fire 

(CHRISTIFIRE) Volume 1: Horizontal 

Trays’’ is available electronically under 

ADAMS Accession Number 

ML102700336. 

Federal Rulemaking Web site: Public 

comments and supporting materials 

related to this notice can be found at 

http://www.regulations.gov by searching 

on Docket ID: NRC–2010–0316. 

FOR FURTHER INFORMATION CONTACT: 

David Stroup, Division of Risk Analysis, 

Office of Nuclear Regulatory Research, 

U.S. Nuclear Regulatory Commission, 

Washington, DC 20555–0001. 

Telephone: 301–251–7609, e-mail: 

David.Stroup@nrc.gov. 

SUPPLEMENTARY INFORMATION: NUREG/ 

CR–7010, Volume 1 documents the first 

phase of a multi-year program called 

CHRISTIFIRE (Cable Heat Release, 

Ignition, and Spread in Tray 

Installations during FIRE). The overall 

goal of the program is to quantify the 

burning characteristics of grouped 

electrical cables. The first phase of the 

program focuses on horizontal tray 

configurations. The experiments 

conducted range from micro-scale, in 

which very small (5 mg) samples of 

cable materials were burned in a 

calorimeter to determine their heat of 

combustion and other properties; to fullscale, in which horizontal, ladder-back 

trays loaded with varying amounts of 

cable were burned under a large oxygendepletion calorimeter. Other 

experiments include cone calorimetry, 

smoke and effluent characterization in a 

small test furnace, and intermediatescale calorimetry involving a single tray 

of cables underneath a bank of radiant 

panels. The results of the small-scale 

experiments are to serve as input data 

for fire models, while the results of the 

full-scale experiments are to serve as 

validation data for the models. 

Dated at Rockville, Maryland, this 28th day 

of September 2010. 

For the Nuclear Regulatory Commission. 

Mark H. Salley, 

Chief, Fire Research Branch, Division of Risk 

Analysis, Office of Nuclear Regulatory 

Research. 

[FR Doc. 2010–24914 Filed 10–4–10; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0309] 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a. (2) of the 

Atomic Energy Act of 1954, as amended 

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

Commission (the Commission or NRC) 

is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

mstockstill on DSKH9S0YB1PROD with NOTICES

61522 Federal Register / Vol. 75, No. 192 / Tuesday, October 5, 2010 / Notices 

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 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 September 

9, 2010, to September 22, 2010. The last 

biweekly notice was published on 

September 21, 2010 (75 FR57521). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

Title 10 of the Code of Federal 

Regulations (10 CFR), Section 50.92, 

this means that operation of the facility 

in accordance with the proposed 

amendment would not (1) involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated; or (2) create the possibility of 

a new or different kind of accident from 

any accident previously evaluated; or 

(3) involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

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. 

Written comments may be submitted 

by mail to the Chief, Rules, 

Announcements and Directives Branch 

(RADB), TWB–05–B01M, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be faxed to the RADB at 301–492– 

3446. Documents may be examined, 

and/or copied for a fee, at the NRC’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. 

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. 

Requests for a hearing and a petition for 

leave to intervene shall be filed in 

accordance with the Commission’s 

‘‘Rules of Practice for Domestic 

Licensing Proceedings’’ in 10 CFR Part 

2. Interested person(s) should consult a 

current copy of 10 CFR 2.309, which is 

available at the Commission’s PDR, 

located at One White Flint North, Public 

File Area O1F21, 11555 Rockville Pike 

(first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

NRC Web site, http://www.nrc.gov/ 

reading-rm/doc-collections/cfr/. If a 

request for a hearing or petition for 

leave to intervene is filed by the above 

date, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the requestor/ 

petitioner seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the requestor/petitioner shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the requestor/petitioner 

intends to rely in proving the contention 

at the hearing. The requestor/petitioner 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the requestor/petitioner intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the requestor/ 

petitioner to relief. A requestor/ 

petitioner who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

mstockstill on DSKH9S0YB1PROD with NOTICES

Federal Register / Vol. 75, No. 192 / Tuesday, October 5, 2010 / Notices 61523 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

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 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 e-mail at 

hearing.docket@nrc.gov, or by telephone 

at (301) 415–1677, to request (1) a 

digital 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 

NRC’s public Web site at http:// 

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

apply-certificates.html. System 

requirements for accessing the ESubmittal server are detailed in NRC’s 

‘‘Guidance for Electronic Submission,’’ 

which is available on the agency’s 

public Web site at http://www.nrc.gov/ 

site-help/e-submittals.html. Participants 

may attempt to use other software not 

listed on the Web site, but should note 

that the 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 EIE, users will be 

required to install a Web browser plugin from the NRC Web site. Further 

information on the Web-based 

submission form, including the 

installation of the Web browser plug-in, 

is available on the NRC’s public Web 

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

e-submittals.html. 

Once a participant has obtained a 

digital ID certificate and a docket has 

been created, the participant can then 

submit a request for hearing or petition 

for leave to intervene. Submissions 

should be in Portable Document Format 

(PDF) in accordance with NRC guidance 

available on the NRC public Web site at 

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

e-submittals.html. A filing is considered 

complete at the time the documents are 

submitted through the 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 e-mail notice 

confirming receipt of the document. The 

E-Filing system also distributes an email notice that provides access to the 

document to the NRC 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 agency’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 Web site at http:// 

www.nrc.gov/site-help/esubmittals.html, by e-mail at 

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

Meta System Help Desk is available 

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

Time, Monday through Friday, 

excluding government holidays. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file an 

exemption request, in accordance with 

10 CFR 2.302(g), with their initial paper 

filing requesting authorization to 

continue to submit documents in paper 

format. Such filings must be submitted 

by: (1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, Attention: Rulemaking and 

Adjudications Staff; or (2) courier, 

express mail, or expedited delivery 

service to the Office of the Secretary, 

Sixteenth Floor, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland 20852, Attention: Rulemaking 

and Adjudications Staff. Participants 

filing a document in this manner are 

responsible for serving the document on 

all other participants. Filing is 

considered complete by first-class mail 

as of the time of deposit in the mail, or 

by courier, express mail, or expedited 

delivery service upon depositing the 

document with the provider of the 

service. A presiding officer, having 

granted an exemption request from 

using E-Filing, may require a participant 

or party to use E-Filing if the presiding 

officer subsequently determines that the 

reason for granting the exemption from 

use of E-Filing no longer exists. 

Documents submitted in adjudicatory 

proceedings will appear in NRC’s 

electronic hearing docket which is 

available to the public at http:// 

ehd.nrc.gov/EHD_Proceeding/home.asp, 

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. 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. Nontimely filings will not be entertained 

absent a determination by the presiding 

officer that the petition or request 

should be granted or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

CFR 2.309(c)(1)(i)–(viii). 

For further details with respect to this 

license amendment application, see the 

application for amendment which is 

available for public inspection at the 

Commission’s PDR, located at One 

White Flint North, Public File Area 

O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. Publicly 

available records will be accessible from 

the ADAMS Public Electronic Reading 

mstockstill on DSKH9S0YB1PROD with NOTICES

61524 Federal Register / Vol. 75, No. 192 / Tuesday, October 5, 2010 / Notices 

Room on the Internet at the NRC Web 

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

adams.html. Persons who do not have 

access to ADAMS or who encounter 

problems in accessing the documents 

located in ADAMS, should contact the 

NRC PDR Reference staff at 1–800–397– 

4209, 301–415–4737, or by e-mail to 

pdr.resource@nrc.gov. 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–369 and 50–370, McGuire 

Nuclear Station, Units 1 and 2, 

Mecklenburg County, North Carolina 

Date of amendment request: May 28, 

2010. 

Description of amendment request: 

The amendments would revise the 

Technical Specifications (TS) to allow 

manual operation of the containment 

spray system (CSS) and to change the 

setpoints for the refueling water storage 

tank (RWST). 

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: 

Criterion 1: Does the proposed amendment 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated? 

Response: No. 

The CSS and RWST are accident mitigation 

equipment. As such, changes in operation of 

these systems cannot have an impact on the 

probability of an accident. 

The RWST will continue to comply with 

all applicable regulatory requirements and 

design criteria following approval of the 

proposed changes (e.g., train separation, 

redundancy, and single failure). The water 

level on the containment floor will be higher 

at the start of transfer to the containment 

sump but will remain below the maximum 

design level analyzed for equipment 

submergence. The change in the sump pH 

will not result in a significant increase in 

radiological consequences of a LOCA [lossof-coolant accident]. Therefore, the design 

functions performed by the equipment are 

not changed. 

The proposed change alters the method of 

controlling the safety system following a 

design basis event so that manual actions are 

substituted for automatic actions. 

Calculations and simulator exercises confirm 

these actions will be taken within the 

appropriate scenario sequence timing to 

provide containment cooling and source term 

reduction. 

The delay in CS [containment spray] 

operation will result in an increase in 

containment temperature, containment 

pressure, offsite dose, and control room dose 

during a LOCA or high energy line break 

inside containment. Containment analyses 

have been performed to demonstrate that 

containment pressure and temperature 

remain within the design limits and there is 

no significant impact on the environmental 

qualification for equipment inside 

containment. The reduction in fission 

product removal due to delayed CS operation 

does not result in exceeding the offsite dose 

and control room dose limits in 10 CFR 

50.67. The analysis of the change in 

containment conditions due to a single 

failure of an operating spray pump and the 

suspension of CS determined that the 

pressure remained below the design limits. 

The proposed change to adopt [Technical 

Specification Task Force] TSTF–493, Rev. 4, 

on a limited basis clarifies requirements for 

instrumentation to ensure the 

instrumentation will actuate as assumed in 

the safety analysis. Instruments are not an 

assumed initiator of any accident previously 

evaluated. As a result, the proposed change 

will not increase the probability of an 

accident previously evaluated. The proposed 

change will ensure that the instruments 

actuate as assumed to mitigate the accidents 

previously evaluated. As a result, the 

proposed change will not increase the 

consequences of an accident previously 

evaluated. 

Based on this discussion, the proposed 

amendment does not significantly increase 

the probability or consequences of an 

accident previously evaluated. 

Criterion 2: Does the proposed amendment 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated? 

Response: No. 

The modification to the low level setpoint 

will not install any new plant equipment. 

The setpoint will continue to be included 

within the engineered safeguards features 

instrumentation and monitored according to 

the applicable surveillance requirements. 

The evaluation of the new level setpoint and 

the change in the switchover sequence 

concluded that the equipment aligned to the 

sump will continue to have sufficient suction 

pressure prior to containment sump suction 

switchover. The design of the RWST low 

level instrumentation complies with all 

applicable regulatory requirements and 

design criteria. 

The overall function of the CSS is not 

changed by this proposed amendment. The 

proposed change alters the method of 

controlling the safety system following a 

design basis event so that manual actions are 

substituted for automatic actions. 

Calculations confirm that these actions will 

be taken within the appropriate scenario 

sequence timing to provide containment 

cooling and source term reduction with no 

significant increase in radiological 

consequences and without exceeding 

containment design limits. 

The proposed change to adopt TSTF–493, 

Rev. 4 on a limited basis does not involve a 

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

or different type of equipment will be 

installed) or a change in the methods 

governing normal plant operation. The 

change does not alter assumptions made in 

the safety analysis but ensures that the 

instruments behave as assumed in the 

accident analysis. The proposed change 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 previously 

evaluated. 

Criterion 3: Does the proposed amendment 

involve a significant reduction in a margin of 

safety? 

Response: No. 

The proposed change will increase the 

calculated radiological dose at the site 

boundary and in the control room. However, 

the calculations demonstrate that the dose 

consequences at the site boundary, low 

population zone, and control room remain 

within regulatory acceptance limits of 10 

CFR 50.67. 

Additional analysis concluded: 

• Peak containment pressure for analyzed 

design basis accidents will not be 

significantly increased and containment 

design limits will not be exceeded. 

• Assumptions used in the environmental 

qualification of equipment exposed to the 

containment atmosphere remain bounding. 

• Pumps aligned to the RWST and to the 

containment sump will have adequate 

suction pressure. 

• The CSS will retain its ability to undergo 

all appropriate testing requirements 

following implementation of the proposed 

amendment. These testing requirements are 

conducted in accordance with the McGuire 

Inservice Testing Program and TS 3.6.6. 

It is estimated that the implementation of 

this license amendment request will result in 

an approximate 22% reduction in core 

damage frequency. This amendment request 

is based on the Nuclear Energy Institute (NEI) 

and the Pressurized Water Reactor (PWR) 

Owners Group initiative to extend the postLoss of Coolant Accident (LOCA) injection 

phase and delay the onset of the containment 

sump recirculation phase. 

The proposed change to adopt TSTF–493, 

Rev. 4 on a limited basis clarifies the 

requirements for instrumentation to ensure 

the instrumentation will actuate as assumed 

in the accident analysis. No change is made 

to the accident analysis assumptions and no 

margin of safety is reduced as part of this 

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 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Lara S. Nichols, 

Associate General Counsel, Duke Energy 

Corporation, 526 South Church Street— 

EC07H, Charlotte, NC 28202. 

NRC Branch Chief: Gloria Kulesa. 

Entergy Gulf States Louisiana, LLC, and 

Entergy Operations, Inc., Docket No. 

50–458, River Bend Station, Unit 1, 

West Feliciana Parish, Louisiana 

Date of amendment request: July 22, 

2010. 

Description of amendment request: 

The proposed amendment would revise 

mstockstill on DSKH9S0YB1PROD with NOTICES

Federal Register / Vol. 75, No. 192 / Tuesday, October 5, 2010 / Notices 61525 

Limiting Condition for Operation (LCO) 

3.10.1, ‘‘Inservice Leak and Hydrostatic 

Testing Operation,’’ and the associated 

Bases, to expand its scope to include 

provisions for temperature excursions 

greater than 200 degrees Fahrenheit (°F) 

as a consequence of inservice leak and 

hydrostatic testing, and as a 

consequence of scram time testing 

initiated in conjunction with an 

inservice leak or hydrostatic test, while 

considering operational conditions to be 

in Mode 4. The proposed change is 

consistent with NRC-approved 

Technical Specification Task Force 

(TSTF) Improved Standard Technical 

Specification Traveller, TSTF–484, ‘‘Use 

of TS 3.10.1 for Scram Time Testing 

Activities,’’ that was announced in the 

Federal Register on October 27, 2001 

(71 FR 63050), as part of the 

consolidated Line Item Improvement 

Process (CCIIP). 

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. 

Technical Specifications currently allow 

for operation at > 200 °F while imposing 

MODE 4 requirements in addition to the 

secondary containment requirements 

required to be met. Extending the activities 

that can apply this allowance will not 

adversely impact the probability or 

consequences of an accident previously 

evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of 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. 

Technical Specifications currently allow 

for operation at > 200 °F while imposing 

MODE 4 requirements in addition to the 

secondary containment requirements 

required to be met. No new operational 

conditions beyond those currently allowed 

by LCO 3.10.1 are introduced. The extended 

allowances would result from operations that 

commence at reduced temperatures, but 

approach the normal MODE 4 limit of 200 °F 

prior to completion of the inspections or 

testing. The changes do not involve a 

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

or different type of equipment will be 

installed) or a change in the methods 

governing normal plant operation. In 

addition, the changes do not impose any new 

or different requirements or eliminate any 

existing requirements. The changes do not 

alter assumptions made in the safety 

analysis. The proposed changes are 

consistent with the safety analysis 

assumptions and current plant operating 

practice. 

Therefore, the proposed 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. 

Technical Specifications currently allow 

for operation at > 200 °F while imposing 

MODE 4 requirements in addition to the 

secondary containment requirements 

required to be met. Extending the activities 

that can apply this allowance will not 

adversely impact any margin of safety. 

Allowing completion of inspections and 

testing and supporting completion of scram 

time testing initiated in conjunction with an 

inservice leak or hydrostatic test prior to 

power operation, results in enhanced safe 

operations by eliminating unnecessary 

maneuvers to control reactor temperature and 

pressure. 

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: Joseph A. 

Aluise, Assistant General Counsel— 

Nuclear, Entergy Services, Inc., 639 

Loyola Avenue, New Orleans, Louisiana 

70113. 

NRC Branch Chief: Michael T. 

Markley. 

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of amendment request: August 

19, 2010. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specifications to be 

consistent with Standard Technical 

Specifications 3.6.1.8 ‘‘Suppression 

Chamber-to-Drywell Vacuum Breakers’’ 

and 3.6.2.5 ‘‘Drywell-to-Suppression 

Chamber Differential Pressure,’’ along 

with the associated Bases, of NUREG– 

1433, Revision 3, ‘‘Standard Technical 

Specifications General Electric Plants, 

BWR/4,’’ modified to account for plant 

specific design details. 

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 amendment does not 

significantly increase the probability or 

consequences of an accident since it does not 

involve a modification to any plant 

equipment or affect how plant systems or 

components are operated. No design 

functions or design parameters are affected 

by the proposed amendment. The proposed 

amendment involves the operation and 

testing of Primary Containment systems but 

does not impact containment design or 

performance requirements. Therefore, the 

proposed amendment does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve any 

physical alteration of plant equipment and 

does not change the method by which any 

safety-related system performs its function. 

No new or different types of equipment will 

be installed and the basic operation of 

installed equipment is unchanged. The 

methods governing plant operation and 

testing remain consistent with current safety 

analysis assumptions. The proposed 

amendment involves the operation and 

testing of Primary Containment systems but 

does not alter the way that the systems are 

operated or how the tests are performed. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change ensures that the 

safety functions of the pressure suppression 

chamber-drywell vacuum breakers and 

drywell-suppression chamber differential 

pressure are fulfilled by incorporating the 

guidance of NUREG–1433. The proposed 

amendment does not involve a physical 

modification of the plant and does not 

change the design or function of any 

component or system. Therefore, the 

proposed amendment will 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: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 400 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Nancy Salgado. 

mstockstill on DSKH9S0YB1PROD with NOTICES

61526 Federal Register / Vol. 75, No. 192 / Tuesday, October 5, 2010 / Notices 

Entergy Operations, Inc., Docket No. 

50–313, Arkansas Nuclear One, Unit 

No. 1, Pope County, Arkansas 

Date of amendment request: August 

10, 2010. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 3.9.3, 

‘‘Reactor Building Penetrations,’’ to 

allow reactor building flow path(s) 

providing direct access from the reactor 

building atmosphere to the outside 

atmosphere to be unisolated under 

administrative control, during 

movement of irradiated fuel assemblies. 

The proposed change is consistent with 

Technical Specification Task Force 

(TSTF) Technical Change Traveler 312, 

Revision 1, ‘‘Administratively Control 

Containment Penetrations.’’ 

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 status of the penetration flow paths 

during fuel movement in the reactor building 

has no affect on the probability of the 

occurrence of any accident previously 

evaluated. The proposed change does not 

alter any plant equipment or operating 

practices in such a manner that the 

probability of an accident is increased. Since 

the consequences of a fuel handling accident 

(FHA) inside the reactor building with open 

penetrations flow paths is bounded by the 

current FHA analyses and the probability of 

an accident is not affected by the status of the 

penetration flow paths, 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 open reactor building penetration flow 

paths are not accident initiators. The 

proposed allowance to open the reactor 

building penetrations during fuel movement 

inside the reactor building will not adversely 

affect plant safety functions or equipment 

operating practices such that a new or 

different accident could be created. 

Therefore, the proposed change does not 

create the possibility of an accident of a 

different kind than previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

Technical Specification (TS) 3.9.3 closure 

requirements for reactor building 

penetrations ensure that the consequences of 

a postulated FHA inside the reactor building 

during irradiated fuel handling activities are 

minimized. The Limiting Condition for 

Operation establishes reactor building 

closure requirements, which limit the 

potential escape paths for fission products by 

ensuring that there is at least one integral 

barrier to the release of radioactive material. 

The proposed change to allow the reactor 

building penetration flow paths to be open 

during refueling operations under 

administrative controls does not significantly 

affect the expected dose consequences of a 

FHA because the limiting FHA does not 

credit reactor building closure or filtration. 

The proposed administrative controls 

provide assurance that prompt closure of the 

penetration flow paths will be accomplished 

in the event of a[n] FHA inside the reactor 

building. The provisions to promptly isolate 

open penetration flow paths provide 

assurance that the offsite dose consequences 

of a[n] FHA inside containment will be 

minimized. Therefore, this proposed change 

does not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Joseph A. 

Aluise, Assistant General Counsel— 

Nuclear, Entergy Services, Inc., 639 

Loyola Avenue, New Orleans, Louisiana 

70113. 

NRC Branch Chief: Michael T. 

Markley. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and STN 50– 

457, Braidwood Station, Units 1 and 2, 

Will County, Illinois; Docket Nos. STN 

50–454 and STN 50–455, Byron Station, 

Unit Nos. 1 and 2, Ogle County, Illinois 

Date of amendment request: June 29, 

2010, as supplemented on August 24, 

2010. 

Description of amendment request: 

The proposed amendments would 

revise Technical Specifications (TS) 

Section 3.4.12, ‘‘Low Temperature 

Overpressure Protection (LTOP) 

System,’’ to correct an inconsistency 

between the TS, and implementation of 

procedures and administrative controls 

for Safety Injection (SI) pumps required 

to mitigate a postulated loss of decay 

heat removal during mid-loop operation 

as discussed in NRC Generic Letter (GL) 

88–17, ‘‘Loss of Decay Heat Removal.’’ 

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 does not result in 

any physical changes to safety related 

structures, systems, or components. The 

proposed change revises TS 3.4.12 to correct 

an inconsistency between the TS, and 

implementation of procedures and 

administrative controls for SI pumps 

required to mitigate a postulated loss of 

decay heat removal during mid-loop 

operation as discussed in GL 88–17. 

Specifically, the proposed change adds a note 

to TS LCO [limiting condition for operation] 

3.4.12 that states: ‘‘For the purpose of 

protecting the decay heat removal function, 

one or more SI pumps may be capable of 

injecting into the RCS in MODE 5 and MODE 

6 when the reactor vessel head is on 

provided pressurizer level is ≤ 5 percent.’’ 

The proposed change corrects an oversight 

introduced during the conversion of the 

Braidwood Station and Byron Station TS to 

the ITS [Improved TS]. 

The probability of occurrence of an 

accident is not increased since the proposed 

change will continue to require that no SI 

pumps are capable of injecting into the RCS 

in Modes 5 and 6 with pressurizer level 

greater than 5 percent. 

The NRC has previously evaluated the 

allowance for one or more SI pumps to be 

capable of injecting into the RCS in Mode 5 

or Mode 6 when the reactor vessel head is 

on provided pressurizer level is ≤ 5 percent 

for the Braidwood Station and Byron Station. 

In a safety evaluation dated August 31, 1990, 

related to Braidwood Station, Units 1 and 2, 

Amendment 25, and Byron Station, Units 1 

and 2, Amendment 38, the NRC concluded 

that allowing SI pump capability to inject 

into the RCS in Mode 5 or Mode 6 when the 

reactor vessel head is on provided 

pressurizer level is ≤ 5 percent was 

acceptable. The availability of SI pumps 

under these circumstances does not present 

a concern regarding cold overpressure 

protection since sufficient air volume exists 

which allows Operations personnel time to 

mitigate the transient. This is in contrast to 

the analyzed cold overpressure transients, in 

which the RCS is assumed to be water solid 

at the onset of the event. 

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 TS 3.4.12 to 

correct an inconsistency between the TS, and 

implementation of procedures and 

administrative controls for SI pumps 

required to mitigate a postulated loss of 

decay heat removal during mid-loop 

operation as discussed in GL 88–17. 

Specifically, the proposed change adds a note 

to TS LCO 3.4.12 that states: ‘‘For the purpose 

of protecting the decay heat removal 

function, one or more SI pumps may be 

mstockstill on DSKH9S0YB1PROD with NOTICES

Federal Register / Vol. 75, No. 192 / Tuesday, October 5, 2010 / Notices 61527 

capable of injecting into the RCS in MODE 

5 and MODE 6 when the reactor vessel head 

is on provided pressurizer level is ≤ 5 

percent.’’ The proposed change corrects an 

oversight introduced during the conversion 

of the Braidwood Station and Byron Station 

TS to the ITS. 

The proposed change is necessary for the 

purpose of mitigating the consequences of a 

loss of decay heat removal during mid-loop 

operations. Operation of at least one SI pump 

is required in some cases to prevent the core 

from uncovering. The only new configuration 

allowed by the proposed change is the 

potential of having one or more SI pumps 

available in Modes 5 and 6 with pressurizer 

level ≤ 5 percent. The potential 

overpressurization accident has been 

analyzed and accounted for by requiring 

pressurizer level to be ≤ 5 percent if one or 

more SI pumps are available. 

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 TS 3.4.12 to 

correct an inconsistency between the TS, and 

implementation of procedures and 

administrative controls for SI pumps 

required to mitigate a postulated loss of 

decay heat removal during mid-loop 

operation as discussed in GL 88–17. 

Specifically, the proposed change adds a note 

to TS LCO 3.4.12 that states: ‘‘For the purpose 

of protecting the decay heat removal 

function, one or more SI pumps may be 

capable of injecting into the RCS in MODE 

5 and MODE 6 when the reactor vessel head 

is on provided pressurizer level is ≤ 5 

percent.’’ The proposed change corrects an 

oversight introduced during the conversion 

of the Braidwood Station and Byron Station 

TS to the ITS. 

The proposed note allows one or more SI 

pumps to be capable of injecting into the RCS 

only when pressurizer level is ≤ 5 percent in 

Mode 5 and Mode 6 when the reactor vessel 

head is on. This provides protection to limit 

coolant input capacity during shutdown in 

which a pressure fluctuation due to coolant 

input from the SI pumps could occur more 

quickly than an operator could react, while 

providing an allowance for one or more SI 

pumps to be capable of injecting into the RCS 

during conditions in which a loss of decay 

heat removal could result in rapid core 

uncovery. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The Nuclear Regulatory Commission 

(NRC) staff has reviewed the licensee’s 

analysis and, based on this review, it 

appears that the three standards of 10 

CFR 50.92(c) are satisfied. Therefore, the 

NRC staff proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Robert D. Carlson. 

Florida Power and Light Company 

(FPL), Docket Nos. 50–250 and 50–251, 

Turkey Point Plant, Units 3 and 4, 

Miami-Dade County, Florida 

Date of amendment request: August 5, 

2010. 

Description of amendment request: 

The proposed amendments would 

revise technical specification (TS) 5.5.1 

Fuel Storage—Criticality, to include 

new spent fuel storage patterns that 

account for both the increase in fuel 

maximum enrichment from 4.5 weight 

percentage (wt%) U–235 to 5.0 wt% 

U–235 and the impact on the fuel of 

higher power operation proposed under 

the Extended Power Uprate (EPU) 

project. Although the fuel storage has 

been analyzed at the higher fuel 

enrichment in the new criticality 

analysis, the fuel enrichment limit of 4.5 

wt% U–235 specified in TS 5.5.1 will 

not be changed under this license 

amendment request. The proposed TS 

changes and a new supporting criticality 

analysis are being submitted to revise 

the current licensing basis analysis for 

both new fuel and spent fuel pool 

storage. 

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? 

No. The proposed amendments do not 

change or modify the fuel, fuel handling 

processes, fuel storage racks, number of fuel 

assemblies that may be stored in the spent 

fuel pool (SFP), decay heat generation rate, 

or the spent fuel pool cooling and cleanup 

system. The proposed amendment was 

evaluated for impact on the following 

previously evaluated events and accidents: 

a. A fuel handling accident (FHA), 

b. A cask drop accident, 

c. A fuel mispositioning event, 

d. A spent fuel pool boron dilution event, 

e. A seismic event, and 

f. A loss of spent fuel pool cooling event. 

Although the proposed amendment will 

require increased handling of the fuel, the 

probability of a FHA is not significantly 

increased because the implementation of the 

proposed amendment will employ the same 

equipment and process to handle fuel 

assemblies that is currently used. Also, tests 

have confirmed that the Metamic inserts can 

be installed and removed without damaging 

the host fuel assemblies. The FHA 

radiological dose consequences associated 

with fuel enrichment at this level were 

addressed in LAR [license amendment 

request] 196 on Alternative Source Term 

implementation at EPU conditions and 

remain unchanged. Therefore, the proposed 

amendments do not significantly increase the 

probability or consequences of a FHA. 

The proposed amendments do not increase 

the probability of dropping a fuel transfer 

cask because they do not introduce any new 

heavy loads to the SFP and do not affect 

heavy load handling processes. Also, the 

insertion of Metamic rack inserts does not 

increase the consequences of the cask drop 

accident because the radiological source term 

of that accident is developed from a nonmechanistically derived quantity of damaged 

fuel stored in the spent fuel pool. Therefore, 

the proposed amendments do not 

significantly increase the probability or 

consequences of a cask drop accident. 

Operation in accordance with the proposed 

amendment will not change the probability 

of a fuel mispositioning event because fuel 

movement will continue to be controlled by 

approved fuel handling procedures. These 

procedures continue to require identification 

of the initial and target locations for each fuel 

assembly that is moved. The consequences of 

a fuel mispositioning event are not changed 

because the reactivity analysis demonstrates 

that the same subcriticality criteria and 

requirements continue to be met for the 

worst-case fuel mispositioning event. 

Operation in accordance with the proposed 

amendment will not change the probability 

of a boron dilution event because the systems 

and events that could affect spent fuel pool 

soluble boron are unchanged. The 

consequences of a boron dilution event are 

unchanged because the proposed amendment 

reduces the soluble boron requirement below 

the currently required value and the 

maximum possible water volume displaced 

by the inserts is an insignificant fraction of 

the total spent fuel pool water volume. 

Operation in accordance with the proposed 

amendment will not change the probability 

of a seismic event. The consequences of a 

seismic event are not significantly increased 

because the forcing functions for seismic 

excitation are not increased and because the 

mass of storage racks with Metamic inserts is 

not appreciably increased. Seismic analyses 

demonstrate adequate stress levels in the 

storage racks when inserts are installed. 

Operation in accordance with the proposed 

amendment will not change the probability 

of a loss of SFP cooling event because the 

systems and events that could affect SFP 

cooling are unchanged. The consequences are 

not significantly increased because there are 

no changes in the SFP heat load or SFP 

cooling systems, structures or components. 

Furthermore, conservative analyses indicate 

that the current design requirements and 

criteria continue to be met with the Metamic 

inserts installed. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. The proposed amendments do not 

change or modify the fuel, fuel handling 

processes, fuel racks, number of fuel 

assemblies that may be stored in the pool, 

mstockstill on DSKH9S0YB1PROD with NOTICES

61528 Federal Register / Vol. 75, No. 192 / Tuesday, October 5, 2010 / Notices 

decay heat generation rate, or the spent fuel 

pool cooling and cleanup system. The effects 

of operating with the proposed amendment 

are listed below. The proposed amendments 

were evaluated for the potential of each effect 

to create the possibility of a new or different 

kind of accident: 

a. Addition of inserts to the fuel storage 

racks, 

b. New storage patterns, 

c. Additional weight from the inserts, 

d. Insert movement above fuel, and 

e. Displacement of fuel pool water by the 

inserts. 

Each insert will be placed between a fuel 

assembly and the storage cell wall, taking up 

some of the space available on two sides of 

the fuel assembly. Tests confirm that the 

insert can be installed and removed without 

damaging the fuel assembly. Analyses 

demonstrate that the presence of the inserts 

does not adversely affect spent fuel cooling, 

seismic capability, or subcriticality. The 

aluminum (alloy 6061) and boron carbide 

materials of construction have been shown to 

be compatible with nuclear fuel, storage 

racks and spent fuel pool environments, and 

generate no adverse material interactions. 

Therefore, placing the inserts into the spent 

fuel pool storage racks cannot cause a new 

or different kind of accident. 

Operation with the proposed fuel storage 

patterns will not create a new or different 

kind of accident because fuel movement will 

continue to be controlled by approved fuel 

handling procedures. These procedures 

continue to require identification of the 

initial and target locations for each fuel 

assembly that is moved. There are no changes 

in the criteria or design requirements 

pertaining to fuel storage safety, including 

subcriticality requirements, and analyses 

demonstrate that the proposed storage 

patterns meet these requirements and criteria 

with adequate margins. Therefore, the 

proposed storage patterns cannot cause a new 

or different kind of accident. 

Operation with the added weight of the 

Metamic inserts will not create a new or 

different accident. The net effect of the 

adding the maximum number of inserts is to 

add less than one percent to the weight of the 

loaded racks. Furthermore, the analyses of 

the racks with Metamic inserts installed 

demonstrate that the stress levels in the rack 

modules continue to be considerably less 

than allowable stress limits. Therefore, the 

added weight from the inserts cannot cause 

a new or different kind of accident. 

Operation with insert movement above 

stored fuel will not create a new or different 

kind of accident. The insert with its handling 

tool weighs considerably less than the weight 

of a single fuel assembly. Single fuel 

assemblies are routinely moved safely over 

fuel assemblies and the same level of safety 

in design and operation will be maintained 

when moving the inserts. Furthermore, the 

effect of a dropped insert to block the top of 

a storage cell has been evaluated in thermalhydraulic analyses. Therefore, the movement 

of inserts cannot cause a new or different 

kind of accident. 

Whereas the installed rack inserts will 

displace a very small fraction of the fuel pool 

water volume and impose a very small 

reduction in operator response time to 

previously-evaluated SFP accidents, the 

reduction will not promote a new or different 

kind of accident. Also, displacement of water 

along two sides of a stored fuel assembly may 

have some local reduction in the peripheral 

cooling flow; however, this effect would be 

small compared to the flow induced through 

the fuel assembly and would in no way 

promote a new or different kind of accident. 

The accidents and events previously 

analyzed and presented in the Boraflex 

Remedy and Alternative Source Term LARs 

remain bounding. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in the margin of 

safety? 

No. The proposed change was evaluated 

for its effect on current margins of safety as 

they relate to criticality, structural integrity, 

and spent fuel heat removal capability. 

The margin of safety for subcriticality 

required by 10 CFR 50.68(b)(4) is unchanged. 

New criticality analysis confirms that 

operation in accordance with the proposed 

amendment continues to meet the required 

subcriticality margins. 

The structural evaluations for the racks and 

spent fuel pool with Metamic inserts 

installed show that the rack and spent fuel 

pool are unimpaired by loading combinations 

during seismic motion, and there is no 

adverse seismic-induced interaction between 

the rack and Metamic inserts. 

The proposed change does not affect spent 

fuel heat generation or the spent fuel pool 

cooling systems. A conservative analysis 

indicates that the design basis requirements 

and criteria for spent fuel cooling continue to 

be met with the Metamic inserts in place, and 

displacing coolant. Thermal hydraulic 

analysis of the local effects of an installed 

rack insert blocking peripheral flow show a 

small increase in local water and fuel clad 

temperatures, but will remain within 

acceptable limits including no departure 

from nucleate boiling. 

Therefore, the proposed changes do not 

involve a significant reduction in the margin 

of safety. 

Based on the above discussion, FPL has 

determined that the proposed change does 

not involve a significant hazards 

consideration. 

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: M.S. Ross, 

Attorney, Florida Power & Light, P.O. 

Box 14000, Juno Beach, Florida 33408– 

0420. 

NRC Branch Chief: Douglas A. 

Broaddus. 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice. 

Virginia Electric and Power Company: 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Unit Nos. 1 and 2, 

Located in Louisa County, Virginia; and 

50–280 and 50–281, Surry Power 

Station, Unit Nos. 1 and 2, Located in 

Surry County, Virginia 

Date of amendment request: May 6 

and February 10, 2010. 

Brief description of amendment 

request: The proposed amendments will 

add Optimized ZIRLO as an acceptable 

fuel rod cladding material and in 

addition, propose adding the 

Westinghouse topical report for 

Optimized ZIRLO to the analytical 

methods used to determine the core 

operating limits listed in the Technical 

Specifications. 

Date of publication of individual 

notice in Federal Register: August 27, 

2010 (75 FR 52781). 

Expiration date of individual notice: 

Comments, September 27, 2010; 

Hearing, October 26, 2010. 

Notice of Issuance of Amendments to 

Facility Operating 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. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

mstockstill on DSKH9S0YB1PROD with NOTICES

Federal Register / Vol. 75, No. 192 / Tuesday, October 5, 2010 / Notices 61529 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

System (ADAMS) Public Electronic 

Reading Room on the Internet at the 

NRC Web site, http://www.nrc.gov/ 

reading-rm/adams.html. If you do not 

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1–(800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr.resource@nrc.gov. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Unit Nos. 1, 2, and 

3, Maricopa County, Arizona 

Date of application for amendment: 

October 30, 2009, as supplemented by 

letters dated April 29 and August 24, 

2010. 

Brief description of amendment: The 

amendments consisted of administrative 

changes to update the licenses and the 

technical specifications as a result of 

changes that were approved in 

previously issued amendments. The 

amendments removed requirements that 

are no longer applicable due to the 

completion of power uprates, the 

replacement of steam generators, the 

removal of part-length control element 

assemblies, the completion of the core 

protection calculator upgrade, and made 

a minor administrative change to the 

nomenclature of the containment sump 

trash racks and screens. 

Date of issuance: September 10, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: Unit 1–179; Unit 2– 

179; Unit 3–179. 

Facility Operating License Nos. NPF– 

41, NPF–51, and NPF–74: The 

amendment revised the Operating 

Licenses and Technical Specifications. 

Date of initial notice in Federal 

Register: January 26, 2010 (75 FR 

4113). The supplemental letters dated 

April 29 and August 24, 2010, 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. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 10, 

2010. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit No. 2, 

Westchester County, New York 

Date of application for amendment: 

November 19, 2009. 

Brief description of amendment: The 

amendment revises the charcoal testing 

criteria in Technical Specification 5.5.9, 

‘‘Ventilation Filter Testing Program.’’ 

Date of issuance: September 13, 2010. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 30 days. 

Amendment No.: 265. 

Facility Operating License No. DPR– 

26: The amendment revised the License 

and the Technical Specifications. 

Date of initial notice in Federal 

Register: January 26, 2010 (75 FR 

4115). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated January 13, 

2010. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 

50–382, Waterford Steam Electric 

Station, Unit 3, St. Charles Parish, 

Louisiana 

Date of amendment request: 

September 9, 2009. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) 3/4 .9.7, ‘‘Crane 

Travel—Fuel Handling Building,’’ to 

permit certain operations needed for dry 

cask storage of spent nuclear fuel. 

Specifically, the proposed change to this 

TS, while continuing to prohibit travel 

of a heavy load over irradiated fuel 

assemblies in the spent fuel pool, would 

permit travel of loads in excess of 2,000 

pounds over a transfer cask containing 

irradiated fuel assemblies, provided a 

single-failure-proof handling system is 

used. 

Date of issuance: September 13, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to the start of the dry cask storage 

operations. 

Amendment No.: 227. 

Facility Operating License No. NPF– 

38: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: November 17, 2009 (74 FR 

59261). The supplemental letters dated 

June 8 and July 22, 2010, 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 13, 

2010. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 

50–382, Waterford Steam Electric 

Station, Unit 3, St. Charles Parish, 

Louisiana 

Date of amendment request: October 

22, 2009. 

Brief description of amendment: The 

amendment modified the Technical 

Specifications (TS) Table 2.2–1 and 

Table 3.3–1. Specifically, the TS 

changes clarify TS Table 2.2–1 Notes (1) 

and (5), TS Table 3.3–1 Notes (a) and 

(c), and TS Table 3.3–1 Actions 2 and 

3, which have resulted in Plant 

Protection System redundancy issues 

with respect to verbatim compliance. 

While the changes modified the table 

notations for the 10¥4 percent Bistable 

in the Tables, they still maintain the 

safety function associated with the Core 

Protection Calculators and High 

Logarithmic Power trip functions, and 

with the small hysteresis for the 10¥4

percent Bistable, there is a negligible 

impact on the Control Element 

Assembly withdrawal analysis. 

Additionally, the calculated peak power 

and heat flux are not significantly 

changed. 

Date of issuance: September 13, 2010. 

mstockstill on DSKH9S0YB1PROD with NOTICES

61530 Federal Register / Vol. 75, No. 192 / Tuesday, October 5, 2010 / Notices 

Effective date: As of the date of 

issuance and shall be implemented 90 

days from the date of issuance. 

Amendment No.: 228. 

Facility Operating License No. NPF– 

38: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: December 15, 2009 (74 FR 

66384). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 13, 

2010. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and STN 50– 

457, Braidwood Station, Units 1 and 2 

(Braidwood), Will County, Illinois; 

Docket Nos. STN 50–454 and STN 50– 

455, Byron Station, Unit Nos. 1 and 2 

(Byron), Ogle County, Illinois 

Date of application for amendment: 

March 29, 2010. 

Brief description of amendment: The 

amendments revise Technical 

Specification (TS) 5.5.7, ‘‘Reactor 

Coolant Pump Flywheel Inspection 

Program,’’ to extend the reactor coolant 

pump (RCP) motor flywheel 

examination frequency from the 

currently-approved 10-year inspection 

interval to an interval not to exceed 20 

years for certain Braidwood and Byron 

RCPs. These changes are consistent with 

TS Task Force (TSTF) traveler TSTF– 

421, ‘‘Revision to RCP Flywheel 

Inspection Program (WCAP–15666),’’ 

Revision 0, that has been approved 

generically for the Westinghouse 

Standard Technical Specifications, 

NUREG–1431. A notice announcing the 

availability of this proposed TS change 

using the Consolidated Line Item 

Improvement Process was published in 

the Federal Register on October 22, 

2003 (68 FR 60422). 

Date of issuance: September 16, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: Braidwood Unit 1– 

163; Braidwood Unit 2–163; Byron Unit 

No. 1–169; and Byron Unit No. 2–169. 

Facility Operating License Nos. NPF– 

72, NPF–77, NPF–37, and NPF–66: The 

amendments revise the TSs and 

Licenses. 

Date of initial notice in Federal 

Register: May 18, 2010 (75 FR 27827). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 16, 

2010. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station, Unit No. 1, DeWitt County, 

Illinois 

Date of application for amendment: 

April 2, 2010, as supplemented by 

letters dated June 19, 2009, and March 

31, 2010. 

Brief description of amendment: The 

amendment revises the Exelon Nuclear 

Radiological Emergency Plan Annex for 

Clinton Station, Table B–1, ‘‘Minimum 

Staffing Requirements for the On-Shift 

Clinton Station ERO,’’ to increase the 

Non-Licensed Operator staffing from 

two to four, allow in-plant protective 

actions to be performed by personnel 

assigned to other functions, and replace 

a Mechanical Maintenance person with 

a Non-Licensed Operator. 

Date of issuance: September 21, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment No.: 191. 

Facility Operating License No. NPF– 

62: The amendment revised the Facility 

Operating License. 

Date of initial notice in Federal 

Register: June 1, 2010 (75 FR 30445). 

The June 19, 2009, and March 31, 

2010, supplement, contained clarifying 

information and did not change the NRC 

staff’s initial proposed finding of no 

significant hazards consideration. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated September 21, 

2010. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, 

LaSalle County Station, Units 1 and 2, 

LaSalle County, Illinois 

Date of application for amendments: 

January 27, 2010, as supplemented by 

letters dated May 12, and May 13, 2010. 

Brief description of amendments: The 

amendments would revise the Operating 

License and technical Specifications to 

implement an increase of approximately 

1.65 percent in rated thermal power 

from the current licensed thermal power 

of 3489 megawatts thermal (MWt) to 

3546 MWt. 

Date of issuance: September 16, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days for Unit 1 and within 90 

days of completion of refueling outage 

L2R13, which is currently scheduled for 

March 2011, for Unit 2. 

Amendment Nos.: 198/185. 

Facility Operating License Nos. NPF– 

11 and NPF–18: The amendments 

revised the Technical Specifications and 

License. 

Date of initial notice in Federal 

Register: May 11, 2010 (75 FR 26289). 

The May 12, and May 13, 2010, 

supplements, contained clarifying 

information and did not change the NRC 

staff’s initial proposed finding of no 

significant hazards consideration. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated September 16, 

2010. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 23rd day 

of September 2010. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2010–24815 Filed 10–4–10; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0312] 

Issuance of Regulatory Guides 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Notice. 

SUMMARY: The Nuclear Regulatory 

Commission (NRC) is providing notice 

of the issuance and availability of 

Regulatory Guides 1.84, Rev. 35, 

‘‘Design, Fabrication, and Materials 

Code Case Acceptability, ASME Section 

III,’’ and RG 1.147, Rev. 16, ‘‘Inservice 

Inspection Code Case Acceptability, 

ASME Section XI, Division 1.’’ 

FOR FURTHER INFORMATION CONTACT: 

Wallace E. Norris, Component Integrity 

Branch, Division of Engineering, Office 

of Nuclear Regulatory Research, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, telephone 

(301) 251–7650 or e-mail 

Wallace.Norris@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Introduction 

The NRC is issuing two final 

Regulatory Guides (RGs) in the agency’s 

‘‘Regulatory Guide’’ series: RG 1.84 and 

RG 1.147. This series was developed to 

describe and make available to the 

public specific program information. 

This information includes methods 

acceptable to the NRC staff for 

implementing specific parts of the 

agency’s regulations, techniques the 

staff uses in evaluating specific 

problems or postulated accidents, and 

data the staff needs in its review of 

applications for permits and licenses. 

mstockstill on DSKH9S0YB1PROD with NOTICES