77906 Federal Register / Vol. 75, No. 239 / Tuesday, December 14, 2010 / Notices 

DEPARTMENT OF JUSTICE 

Drug Enforcement Administration 

[OMB Number 1117–0043] 

Agency Information Collection 

Activities: Proposed Collection; 

Comments Requested: Drug 

Questionnaire DEA Form 341 

ACTION: 60-Day Notice of Information 

Collection Under Review. 

The Department of Justice (DOJ), Drug 

Enforcement Administration (DEA), will 

be submitting the following information 

collection request to the Office of 

Management and Budget (OMB) for 

review and approval in accordance with 

the Paperwork Reduction Act of 1995. 

The proposed information collection is 

published to obtain comments from the 

public and affected agencies. Comments 

are encouraged and will be accepted 

until February 14, 2011. This process is 

conducted in accordance with 5 CFR 

1320.10. 

If you have comments, especially on 

the estimated public burden or 

associated response time, suggestions, 

or need a copy of the proposed 

information collection instrument with 

instructions or additional information, 

please contact Raymond A. Pagliarini, 

Jr., Assistant Administrator, Human 

Resources Division, Drug Enforcement 

Administration, 8701 Morrissette Drive, 

Springfield, VA 22152. 

Written comments and suggestions 

from the public and affected agencies 

concerning the proposed collection of 

information are encouraged. Your 

comments should address one or more 

of the following four points: 

• Evaluate whether the proposed 

collection of information is necessary 

for the proper performance of the 

functions of the agency, including 

whether the information will have 

practical utility; 

• Evaluate the accuracy of the 

agency’s estimate of the burden of the 

proposed collection of information, 

including the validity of the 

methodology and assumptions used; 

• Enhance the quality, utility, and 

clarity of the information to be 

collected; and 

• Minimize the burden of the 

collection of information on those who 

are to respond, including through the 

use of appropriate automated, 

electronic, mechanical, or other 

technological collection techniques or 

other forms of information technology, 

e.g., permitting electronic submission of 

responses. 

Overview of Information Collection 

1117–0043 

(1) Type of Information Collection: 

Extension of a currently approved 

collection. 

(2) Title of the Form/Collection: Drug 

Questionnaire (DEA Form 341). 

(3) Agency form number, if any, and 

the applicable component of the 

Department sponsoring the collection: 

Form number: DEA Form 341. 

Component: Human Resources 

Division, Drug Enforcement 

Administration, U.S. Department of 

Justice 

(4) Affected public who will be asked 

or required to respond, as well as a brief 

abstract: 

Primary: Individuals. 

Other: None. 

Abstract: DEA Policy states that a past 

history of illegal drug use may be a 

disqualification for employment with 

DEA. This form asks job applicants 

specific questions about their personal 

history, if any, of illegal drug use. 

(5) An estimate of the total number of 

respondents and the amount of time 

estimated for an average respondent to 

respond: It is estimated that 173,800 

respondents will respond annually, 

taking 5 minutes to complete each form. 

(6) An estimate of the total public 

burden (in hours) associated with the 

collection: 14,483 annual burden hours 

If additional information is required 

contact: Lynn Murray, Department 

Clearance Officer, United States 

Department of Justice, Justice 

Management Division, Policy and 

Planning Staff, Patrick Henry Building, 

Suite 1600, 601 D Street NW., 

Washington, DC 20530. 

Dated: December 8, 2010 

Lynn Murray, 

Department Clearance Officer, PRA, U.S. 

Department of Justice. 

[FR Doc. 2010–31280 Filed 12–13–10; 8:45 am] 

BILLING CODE 4410–09–P 

NATIONAL CREDIT UNION 

ADMINISTRATION 

Sunshine Act; Notice of Agency 

Meeting 

TIME AND DATE: 9 a.m., Friday, December 

17, 2010. 

PLACE: Board Room, 7th Floor, Room 

7047, 1775 Duke Street, Alexandria, VA 

22314–3428. 

STATUS: Closed. 

Matters To Be Considered 

1. Consideration of Supervisory 

Activities (3). Closed pursuant to some 

or all of the following: Exemptions (8), 

(9)(A)(ii) and 9(B). 

2. Personnel. Closed pursuant to 

exemption (2). 

FOR FURTHER INFORMATION CONTACT: 

Mary Rupp, Secretary of the Board, 

Telephone: 703–518–6304. 

Mary Rupp, 

Board Secretary. 

[FR Doc. 2010–31495 Filed 12–10–10; 4:15 pm] 

BILLING CODE P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0382] 

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 

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 November 

18, 2010, to December 1, 2010. The last 

biweekly notice was published on 

November 30, 2010 (75 FR 74091). 

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 


srobinson on DSKHWCL6B1PROD with NOTICES
Federal Register / Vol. 75, No. 239 / Tuesday, December 14, 2010 / Notices 77907 

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, Room 

O1–F21, 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/doccollections/cfr/. If a request for a hearing 

or petition for leave to intervene is filed 

by the above date, the Commission or a 

presiding officer designated by the 

Commission or by the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board Panel, will 

rule on the request and/or petition; and 

the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the requestor/ 

petitioner seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the requestor/petitioner shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the requestor/petitioner 

intends to rely in proving the contention 

at the hearing. The requestor/petitioner 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the requestor/petitioner intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the requestor/ 

petitioner to relief. A requestor/ 

petitioner who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, 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 


srobinson on DSKHWCL6B1PROD with NOTICES
77908 Federal Register / Vol. 75, No. 239 / Tuesday, December 14, 2010 / Notices 

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 Electronic 

Information Exchange System, users 

will be required to install a Web 

browser plug-in 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/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 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 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 1–866–672–7640. The NRC 

Meta System Help Desk is available 

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

Time, Monday through Friday, 

excluding government holidays. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file an 

exemption request, in accordance with 

10 CFR 2.302(g), with their initial paper 

filing requesting authorization to 

continue to submit documents in paper 

format. Such filings must be submitted 

by: (1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, Attention: Rulemaking and 

Adjudications Staff; or (2) courier, 

express mail, or expedited delivery 

service to the Office of the Secretary, 

Sixteenth Floor, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland 20852, Attention: Rulemaking 

and Adjudications Staff. Participants 

filing a document in this manner are 

responsible for serving the document on 

all other participants. Filing is 

considered complete by first-class mail 

as of the time of deposit in the mail, or 

by courier, express mail, or expedited 

delivery service upon depositing the 

document with the provider of the 

service. A presiding officer, having 

granted an exemption request from 

using E-Filing, may require a participant 

or party to use E-Filing if the presiding 

officer subsequently determines that the 

reason for granting the exemption from 

use of E-Filing no longer exists. 

Documents submitted in adjudicatory 

proceedings will appear in 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, Room O1–F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the ADAMS 

Public Electronic Reading 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. 50–269, 50–270, and 50–287, 

Oconee Nuclear Station, Units 1, 2, and 

3, Oconee County, South Carolina 

Date of amendment request: June 29, 

2009, as supplemented June 24, 2010. 

Description of amendment request: 

The proposed amendments would 

approve changes to the updated final 

safety analysis report to allow the use of 

fiber reinforce polymer on masonry 

walls for uniform pressure loads 

resulting from a tornado event. 

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) Involve a Significant Increase in The 

Probability or Consequences of an Accident 

Previously Evaluated 

Response: Physical protection from a 

tornado event is a design basis criterion 

rather than a requirement of a previously 

analyzed [updated final safety analysis 

report] UFSAR accident analysis. The current 


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Federal Register / Vol. 75, No. 239 / Tuesday, December 14, 2010 / Notices 77909 

licensing basis (CLB) for Oconee states that 

systems, structures, and components (SSC’s) 

required to shut down and maintain the units 

in a shutdown condition will not fail as a 

result of damage caused by natural 

phenomena. 

The in-fill masonry walls to be 

strengthened using an FRP system are 

passive, non-structural elements. The use of 

a fiber reinforced polymer [FRP] system on 

existing Auxiliary Building masonry walls 

will allow them to resist uniform pressure 

loads resulting from a tornado and will not 

adversely affect the structure’s ability to 

withstand other design basis events such as 

earthquakes or fires. Therefore, the proposed 

use of FRP on existing masonry walls will 

not significantly increase the probability or 

consequences of an accident previously 

evaluated. 

(2) Create the Possibility of a New or 

Different Kind of Accident From Any 

Accident Previously Evaluated 

Response: The final state of the FRP system 

is passive in nature and will not initiate or 

cause an accident. More generally, this 

understanding supports the conclusion that 

the potential for new or different kinds of 

accidents is not created. 

(3) Involve a Significant Reduction in a 

Margin of Safety 

Response: The application of an FRP 

system to existing Auxiliary Building 

masonry walls will act to enhance the margin 

of safety, e.g., the West Penetration Room 

walls, by increasing the walls’ ability to resist 

tornado-induced differential pressure. 

Consequently, this change does not involve 

a significant reduction in a margin of safety. 

In summary, based upon the above 

evaluation, Duke has concluded that the 

proposed amendment involves no significant 

hazards consideration. 

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. 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–269, 50–270, and 50–287, 

Oconee Nuclear Station, Units 1, 2, and 

3, Oconee County, South Carolina 

Date of amendment request: July 14, 

2010. 

Description of amendment request: 

The proposed amendments would 

revise the Technical Specifications (TS) 

to adopt NRC Approved Technical 

Specification Task Force (TSTF) Change 

to the Standard TS, TSTF–52 

concerning performance-based 

containment leakage testing 

requirements. 

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. Implementation of these changes will 

provide continued assurance that specified 

parameters associated with containment 

integrity will remain within acceptance 

limits as delineated in [Title 10 of the Code 

of Federal Regulations (10 CFR) Part 50] 10 

CFR Part 50, Appendix J, Option B. The 

changes are consistent with current safety 

analyses. Although some of the proposed 

changes represent minor relaxation to 

existing [Technical Specifications] TS 

requirements, they are consistent with the 

requirements specified by Option B of 10 

CFR Part 50, Appendix J. The systems 

affecting containment integrity related to this 

proposed amendment request are not 

assumed in any safety analyses to initiate any 

accident sequence. Therefore, the probability 

of any accident previously evaluated is not 

increased by this proposed amendment. The 

proposed changes maintain an equivalent 

level of reliability and availability for all 

affected systems. In addition, maintaining 

leakage within analyzed limits assumed in 

accident analyses does not adversely affect 

either onsite or offsite dose consequences. 

Therefore, adopting Appendix J, Option B 

does not significantly increase the probability 

or consequences of any accident previously 

evaluated. 

(2) Does the proposed amendment create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. No changes are being proposed which 

will introduce any physical changes to the 

existing plant design. The proposed changes 

are consistent with the current safety 

analyses. Some of the changes may involve 

revision in the testing of components; 

however, these are in accordance with the 

current safety analyses and provide for 

appropriate testing or surveillance that is 

consistent with 10 CFR Part 50, Appendix J, 

Option B. The proposed changes will not 

introduce new failure mechanisms beyond 

those already considered in the current 

accident analyses. No new modes of 

operation are introduced by the proposed 

changes. The proposed changes maintain, at 

minimum, the present level of operability of 

any system that affects containment integrity. 

Therefore, adoption of Appendix J, Option 

B will not create the possibility of a new or 

different kind of accident from any kind of 

accident previously evaluated. 

(3) Does the proposed amendment involve a 

significant reduction in a margin of safety? 

No. The provisions specified in Option B 

of 10 CFR Part 50, Appendix J allow changes 

to Type B and Type C test intervals based 

upon the performance of past leak rate tests. 

10 CFR Part 50, Appendix J, Option B allows 

longer intervals between leakage tests based 

on performance trends, but does not relax the 

leakage acceptance criteria. Changing test 

intervals from those currently provided in 

the TS to those provided in 10 CFR Part 50, 

Appendix J, Option B does not increase any 

risks above and beyond those that the [U S. 

Nuclear Regulatory Commission] NRC has 

deemed acceptable for the performance based 

option. In addition, there are risk reduction 

benefits associated with reduction in 

component cycling, stress, and wear 

associated with increased test intervals. The 

proposed changes provide continued 

assurance of leakage integrity of containment 

without adversely affecting the public health 

and safety and will not significantly reduce 

existing safety margins. 

Therefore, adoption of Appendix J, option 

B 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. 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–269, 50–270, and 50–287, 

Oconee Nuclear Station (ONS), Units 1, 

2, and 3, Oconee County, South 

Carolina; Docket Nos. 50–369 and 50– 

370, McGuire Nuclear Station (MNS), 

Units 1 and 2, Mecklenburg County, 

North Carolina; Docket Nos. 50–413 and 

50–414, Catawba Nuclear Station (CNS), 

Units 1 and 2, York County, South 

Carolina 

Date of amendment request: 

September 16, 2010. 

Description of amendment request: 

The proposed amendments would 

revise the Technical Specifications to 

update the qualification requirements 

for the Station Manager and Radiation 

Protection Manager to meet or exceed 

the minimum qualifications in ANSI/ 

ANS–3.1–1993, ‘‘Selection, 

Qualification, and Training of Personnel 

for Nuclear Power Plants.’’ 

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. 


srobinson on DSKHWCL6B1PROD with NOTICES
77910 Federal Register / Vol. 75, No. 239 / Tuesday, December 14, 2010 / Notices 

The proposed change to [Technical 

Specifications] TS 5.3.1 is an administrative 

change to update the minimum qualification 

requirements for Station Manager and 

Radiation Protection Manager to meet or 

exceed ANSI/ANS 3.1–1993 as endorsed by 

Regulatory Guide 1.8, Revision 3, dated May 

2000. This update for Station Manager and 

Radiation Protection Manager qualifications 

will also provide Oconee, McGuire, and 

Catawba the needed flexibility to appoint 

Station Managers and Radiation Protection 

Managers from a larger candidate pool. The 

current qualification requirements restrict the 

pool of personnel capable of performing the 

Station Manager and Radiation Protection 

Manager functions. This change will also 

revise the current Oconee, McGuire, and 

Catawba TS 5.3.1 qualification requirements 

for Station Manager and Radiation Protection 

Manager to be consistent among all three 

stations. The proposed change does not 

impact the physical configuration or function 

of plant structures, systems, or components 

or the manner in which structures, systems, 

or components are operated, maintained, 

modified, tested, or inspected. Updating the 

minimum qualification requirements for 

Station Manager and Radiation Protection 

Manager is not an initiator of any accident 

previously evaluated. Updating the minimum 

qualification requirements for Station 

Manager and Radiation Protection Manager is 

not an assumption in the consequence 

mitigation of any accident previously 

evaluated. Therefore, it is concluded that this 

change does not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

2. Does the proposed amendment create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change to TS 5.3.1 is an 

administrative change to update the 

minimum qualification requirements for 

Station Manager and Radiation Protection 

Manager to meet or exceed ANSI/ANS 3.1– 

1993 as endorsed by RG 1.8, Revision 3, 

dated May 2000. This represents an update 

to current guidance. This update for Station 

Manager and Radiation Protection Manager 

qualifications will also provide Oconee, 

McGuire, and Catawba the needed flexibility 

to appoint Station Manager and Radiation 

Protection Manager from a larger candidate 

pool. The current qualification requirements 

restrict the pool of personnel capable of 

performing the Station Manager and 

Radiation Protection Manager functions. This 

change will also revise the current Oconee, 

McGuire and Catawba TS 5.3.1 qualification 

requirements for Station Manager and 

Radiation Protection Manager to be 

consistent among all three stations. 

The proposed change does not impact the 

physical configuration or function of plant 

structures, systems, or components or the 

manner in which structures, systems, or 

components are operated, maintained, 

modified, tested, or inspected. In addition, 

there is no change in the types or increases 

in the amounts of effluents that may be 

released offsite, and there is no increase in 

individual or cumulative occupational 

radiation exposure. 

As the proposed change is administrative 

in nature, operation of the facility in 

accordance with 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 proposed change to TS 5.3.1 is an 

administrative change to update the 

minimum qualification requirements for 

Station Manager and Radiation Protection 

Manager to meet or exceed ANSI/ANS 3.1– 

1993 as endorsed by RG 1.8, Revision 3, 

dated May 2000. This update for Station 

Manager and Radiation Protection Manager 

qualifications will also provide Oconee, 

McGuire, and Catawba the needed flexibility 

to appoint Station Manager and Radiation 

Protection Manager from a larger candidate 

pool. The current qualification requirements 

restrict the pool of personnel capable of 

performing the Station Manager and 

Radiation Protection Manager functions. This 

change will also revise the current ONS, 

MNS, and CNS TS 5.3.1 qualification 

requirements for Station Manager and 

Radiation Protection Manager to be 

consistent among all three stations. The 

proposed change does not impact the 

physical configuration or function of plant 

structures, systems, or components or the 

manner in which structures, systems, or 

components are operated, maintained, 

modified, tested, or inspected. The proposed 

change does not alter the manner in which 

safety limits, limiting safety system settings 

or limiting conditions for operation are 

determined. The safety analysis acceptance 

criteria are not affected by this change. The 

proposed change will not result in plant 

operation in a configuration outside the 

design basis. The proposed change does not 

adversely affect systems that respond to 

safely shutdown the plant and to maintain 

the plant in a safe shutdown condition. The 

proposed change is administrative in nature; 

thus operation of the facility in accordance 

with the proposed amendment does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Lara S. Nichols, 

Associate General Counsel, Duke Energy 

Corporation, 526 South Church Street— 

EC07H, Charlotte, NC 28202. 

NRC Branch Chief: Gloria Kulesa. 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–269, 50–270, and 50–287, 

Oconee Nuclear Station, Units 1, 2, and 

3, Oconee County, South Carolina 

Date of amendment request: 

November 8, 2010. 

Description of amendment request: 

The proposed amendments would 

approve revisions to the updated final 

safety analysis report to incorporate the 

licensee’s reactor vessel internals 

inspection plan. 

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) Involve a significant increase in the 

probability or consequences of an accident 

previously evaluated 

No. The proposed license amendment 

request provides the Reactor Vessel Internals 

Inspection Plan report. The report also 

provides a description of the inspection plan 

as it relates to the management of aging 

effects consistent with previous 

commitments. The inspection plan is based 

on MRP–227, Revision 0, ‘‘Pressurized Water 

Reactors Internals Inspection and Evaluation 

Guidelines’’ and describes using the ten 

Aging Management Program (AMP) elements 

in the current revision of NUREG–1801 

‘‘Generic Aging Lessons Learned’’ (GALL, 

Revision 1) report. 

The inspection plan contains a discussion 

of the background of the Babcock and Wilcox 

designed plant Reactor Vessel Internals 

programs, first sponsored by the utilities 

through the Babcock and Wilcox Owner’s 

Group and later by the Pressurized Water 

Reactor Owner’s Group, culminating in a 

submittal to the Nuclear Regulatory 

Commission through the Electric Power 

Research Institute Materials Reliability 

Program. The inspection plan also contains a 

discussion of operational experience, timelimited aging analyses, and relevant existing 

programs. 

The Reactor Vessel Internals Aging 

Management Program includes the 

inspection plan and demonstrates that the 

program adequately manages the effects of 

aging for Reactor Vessel Internals 

components and establishes the basis for 

providing reasonable assurance the Reactor 

Vessel Internals components will remain 

functional through the license renewal 

period of extended operation. 

This license amendment request provides 

an inspection plan based on industry work 

and experiences as agreed to in Duke 

Energy’s license renewal commitments for 

Reactor Vessel Internals Inspection. It is not 

an accident initiator; therefore, it will not 

increase the probability or consequences of 

an accident previously evaluated. 

(2) Create the possibility of a new or different 

kind of accident from any accident 

previously evaluated 

No. The proposed Reactor Vessel Internals 

Inspection Plan does not change the methods 

governing normal plant operation, nor are the 

methods utilized to respond to plant 

transients altered. The revised inspection 

plan is not an accident/event initiator. No 

new initiating events or transients result from 

the use of the Reactor Vessel Internals 

Inspection plan. 


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(3) Involve a significant reduction in a 

margin of safety 

No. The proposed safety limits have been 

preserved. The License Amendment Request 

requests review and approval for the Reactor 

Vessel Internals Inspection plan that Duke 

Energy committed to provide prior to 

commencing inspections. 

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. 

Duke Energy Carolinas, LLC, Docket 

Nos. 50–269, 50–270, and 50–287, 

Oconee Nuclear Station, Units 1, 2, and 

3, Oconee County, South Carolina 

Date of amendment request: 

November 15, 2010. 

Description of amendment request: 

The proposed amendments would 

approve changes to the updated final 

safety analysis report to allow operation 

of a reverse osmosis system during 

normal plant operation to remove silica 

from borated water storage tank and the 

spent fuel pool. 

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 change requests Nuclear 

Regulatory Commission (NRC) approval of 

design features and controls that will be used 

to ensure that periodic limited operation of 

a Reverse Osmosis (RO) System during Unit 

operation does not significantly impact the 

Borated Water Storage Tank (BWST) or Spent 

Fuel Pool (SFP) function or other plant 

equipment. Duke Energy evaluated the effect 

of potential failures, identified precautionary 

measures that must be taken before and 

during RO System operation, and required 

operator actions to protect affected 

structures, systems, and components (SSCs) 

important to safety. The new high energy 

piping and non seismic piping being 

installed for the RO System is non QA–1 and 

is postulated to fail and cause an Auxiliary 

Building flood. Duke Energy determined that 

adequate time is available to isolate the flood 

source (BWST or SFP) prior to affecting SSCs 

important to safety. 

The existing Auxiliary Building Flood 

evaluation postulates a single break in the 

nonseismic piping occurring in a seismic 

event. The addition of the RO System will 

not increase the frequency of a seismic event. 

This event does not consider the amount of 

non-seismic piping that is currently in the 

Auxiliary Building. The new piping is not 

more likely to fail as compared to the existing 

non-seismic piping. The existing postulated 

source of the pipe break in the Auxiliary 

Building is due to the piping not being 

seismically designed. The new RO System 

piping is considered a potential source of a 

single pipe break for the same reason. Since 

the accident itself is defined as the failure of 

non-seismic pipe, the new non-seismic 

piping does not increase the frequency of 

occurrence of an Auxiliary Building flood. 

The mitigation of an Auxiliary Building flood 

due to non seismic piping failure is by 

manual operator action. The same mitigation 

technique is used for the high energy line 

break. 

The RO System takes suction from the top 

of the SFP to protect SFP inventory. Plant 

procedures will prohibit the use of the RO 

System during the time period directly after 

an outage that requires the Unit 1 & 2 SFP 

level to be maintained higher than the 

Technical Specification (TS) Limiting 

Condition for Operation (LCO) 3.7.11 level 

requirement. The higher level is required to 

support TS LCO 3.10.1 requirements for 

Standby Shutdown Facility (SSF) Reactor 

Coolant (RC) Makeup System operability 

(due to the additional decay heat from the 

recently offloaded spent fuel). Plant 

procedures will also specify the siphon be 

broken during this time period so the SFP 

water above the RO suction point cannot be 

siphoned off if the RO piping breaks. The 

proposed change does not impact the fuel 

assemblies, the movement of fuel, or the 

movement of fuel shipping casks. The SFP 

boron concentration, level, and temperature 

limits will not be outside of required 

parameters due to restrictions/requirements 

on the system’s operation. 

The BWST is used for mitigation of Steam 

Generator Tube Rupture (SGTR), Main Steam 

Line Break (MSLB) and Loss of Coolant 

Accidents (LOCAs). The SGTR and MSLB are 

bounded by the [small-break] SBLOCA 

analyses with respect to the performance 

requirements for the [high pressure injection] 

HPI System. In the normal mode of Unit 

operation, the BWST is not an accident 

initiator. The SFP is assumed to maintain 

acceptable criticality margin for all abnormal 

and accident conditions including Fuel 

Handling Accidents (FHAs) and cask drop 

accidents. Both the BWST and SFP are 

specified by TS requirements to have 

minimum levels/volumes and boron 

concentrations. The BWST also has TS 

requirements for temperature. Prior to RO 

operation, procedures will require that 

minimum required initial boron 

concentration, and initial level/volume be 

adjusted and that the RO System be operated 

for a specified maximum time period before 

readjusting volume and boron concentration 

prior to another RO session to ensure that the 

TS specified boron concentration and level/ 

volume limits for both the SFP and the 

BWST are not exceeded during RO System 

operation. Thus, the design functions of the 

BWST and the SFP will continue to be met 

during RO System operation. 

An Auxiliary Building flood due to a nonseismic RO System pipe break does not 

increase the consequences of the flood since 

the new non-seismic pipe break is bounded 

by the Auxiliary Building flood caused by 

existing non-seismic pipe breaks. Although 

the RO System will return water with lower 

boron concentration, procedural controls will 

ensure that the TS boron concentration level 

does not go below the limit. Thus, no adverse 

effects from decreased boron concentration 

levels will occur. 

Since the BWST and SFP will still have TS 

required boron concentration and level/ 

volume, the mitigation of a LOCA or FHA 

does not result in an increase in dose 

consequence. 

Therefore, installation and operation of the 

RO System during Unit operation does not 

significantly increase the probability or 

consequences of any accident previously 

evaluated. 

(2) Does the proposed amendment create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. The RO System adds non-seismic 

piping in the Auxiliary Building. However, 

the break of a single non-seismic pipe in the 

Auxiliary Building has already been 

postulated as an event in the licensing basis. 

The RO System also does not create the 

possibility of a seismic event concurrent with 

a LOCA since a seismic event is a natural 

phenomena event. The RO System does not 

adversely affect the Reactor Coolant System 

pressure boundary. The suction to the RO 

System, when using the system for BWST 

purification, contains a normally closed 

manual seismic boundary valve so the 

seismic design criteria is met for separation 

of seismic/non-seismic piping boundaries. 

Duke Energy also evaluated potential 

releases of radioactive liquid to the 

environment due to RO System piping 

failures. Design features and administrative 

controls preclude release of radioactive 

materials outside the Auxiliary Building. 

Releases inside the Auxiliary Building are 

bounded by existing analyses. 

The SFP suction line is designed such that 

the SFP water level will not go below TS 

required levels, thus the fuel assemblies will 

have the TS required water level over them. 

Procedural controls will restrict the use of 

the RO System and require breaking vacuum 

on the SFP suction line when the SSF 

conditions require the SFP level be raised to 

support SSF RC Makeup System operability. 

Thus, the SFP water level will not be reduced 

below required water levels for these 

conditions. RO System operating restrictions 

will prevent reducing the SFP boron 

concentration below TS limits. 

Therefore, operation of the RO System 

during Unit operation will not create the 

possibility of a new or different kind of 

accident from any kind of accident 

previously evaluated. 

(3) Does the proposed amendment involve a 

significant reduction in a margin of safety? 

No. The RO System adds non-seismic 

piping in the Auxiliary Building. Duke 


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Energy evaluated the impact of RO System 

operation on SSCs important to safety and 

determined that procedural controls will 

ensure that TS limits for SFP and BWST 

volume, temperature and boron 

concentration will continue to be met during 

RO operation. For the BWST, these controls 

will ensure the TS minimum BWST boron 

concentration and level are available to 

mitigate the consequences of a small break 

LOCA or a large break LOCA. For the SFP, 

these controls ensure the assumptions of the 

fuel handling and cask drop accident 

analyses are preserved. Additionally, the 

failure of non seismic RO System piping will 

not significantly impact SSCs important to 

safety. The BWST level may drop below the 

TS required level due to a rupture of the non 

seismic piping during a seismic event. 

However, due to the low probability of a 

seismic event coupled with the relatively 

short period of time the RO System will be 

aligned to the BWST, the possibility of 

dropping below the TS required level does 

not involve a significant reduction in the 

margin of safety. In addition, Oconee’s 

licensing basis does not assume a design 

basis event occurs simultaneously with a 

seismic event. The proposed change does not 

significantly impact the condition or 

performance of SSCs relied upon for accident 

mitigation. This change does not alter the 

existing TS allowable values or analytical 

limits. The existing operating margin 

between Unit conditions and actual Unit 

setpoints is not significantly reduced due to 

these changes. The assumptions and results 

in any safety analyses are not impacted. 

Therefore, operation of the RO System during 

Unit operation 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. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of amendment request: 

September 30, 2010. 

Description of amendment request: 

The proposed amendment would 

modify Technical Specification (TS) 

3.1.7, ‘‘Standby Liquid Control (SLC) 

System,’’ to add Surveillance 

Requirement (SR) 3.1.7.9 to verify 

sodium pentaborate enrichment prior to 

the addition to the SLC tank. The 

increase in boron-10 enrichment is 

needed to support future reloads of 

GE14 fuel by providing additional 

margin for preserving the shutdown 

objective of the SLC system. Reload 

analysis indicates that a core that is 

made up of a majority of GE14 fuel has 

a higher reactivity than previous 

Columbia Generating Station core 

designs warranting a corresponding 

increase in the shutdown capability of 

the SLC system. 

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 SLC system is designed to provide 

sufficient negative reactivity to bring the 

reactor from full power to a subcritical 

condition at any time in a fuel cycle, without 

taking credit for control rod movement. The 

proposed changes to the SLC sodium 

pentaborate solution requirements maintain 

the capability of the SLC to perform this 

reactivity control function, and assure 

continued compliance with the requirements 

of 10 CFR 50.62 for ATWS [automatic 

transient without scram]. The proposed 

changes do not impact the LOCA [loss-ofcoolant accident] suppression pool pH 

control function of SLC because single-pump 

minimum flow and sodium pentaborate 

solution concentration (weight percent) are 

not changed from the level credited in the 

LOCA analysis. The SLC is provided to 

mitigate ATWS events and LOCA and, as 

such, is not considered to be an initiator of 

the ATWS event, LOCA, or any other 

analyzed accident. The use of sodium 

pentaborate solution enriched with the 

boron-10 isotope, which is chemically and 

physically similar to the current solution, 

does not alter the design or operation of the 

SLC or increase the likelihood of a system 

malfunction that could increase the 

consequences of an accident. 

Based on the above discussion, it is 

concluded that the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Injection of sodium pentaborate solution 

into the reactor vessel has been considered in 

the plant design. The proposed changes 

revise the SLC boron solution requirements 

such that the capability of the SLC system to 

bring the reactor to a subcritical condition 

without taking credit for control rod 

movement is maintained, considering 

operation with an equilibrium core of GE14 

fuel. The use of sodium pentaborate solution 

enriched with the boron-10 isotope, which is 

chemically and physically similar to the 

current solution, does not alter the design, 

function, or operation of the SLC system. The 

correct boron-10 enrichment is assured by 

the proposed addition of an SR to the TS. 

The solution concentration and volume are 

not changed; thus, the existing minimum 

volume and solution and piping temperature 

specified in the TS will ensure that the boron 

remains in solution and does not precipitate 

out in the SLC storage tank or in the SLC 

pump suction piping. The minimum volume 

and concentration specified in the TS ensure 

that the LOCA suppression pool pH control 

function is not impacted. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes revise the SLC 

boron solution requirements to maintain the 

capability of the SLC system to bring the 

reactor to a subcritical condition without 

taking credit for control rod movement. 

These changes support operation with an 

equilibrium core of GE14 fuel and assure 

continued compliance with the requirements 

of 10 CFR 50.62. The minimum required 

average boron-10 concentration in the reactor 

core, resulting from the injection of sodium 

pentaborate solution by the SLC system, has 

been determined using approved analytical 

methods. The analysis demonstrates that 

sufficient shutdown margin is maintained in 

the reactor such that the reactivity control 

function of the SLC system is assured. The 

additional quantity of boron included to 

account for imperfect mixing and leakage is 

maintained at 25 percent. No change in the 

solution pH or volume is made. Thus, the 

safety margin is maintained to bring the 

reactor subcritical in the event of an ATWS 

and to control suppression pool pH in the 

event of a LOCA. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: William A. 

Horin, Esq., Winston & Strawn, 1700 K 

Street, NW., Washington, DC 20006– 

3817. 

NRC Branch Chief: Michael T. 

Markley. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Nuclear 

Plant, Van Buren County, Michigan 

Date of amendment request: July 20, 

2010. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 3.8.3, 

‘‘Diesel Fuel, Lube Oil, and Starting 

Air,’’ by relocating the current stored 


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diesel fuel oil and lube oil numerical 

volume requirements from the TS to the 

TS Bases so that they may be modified 

under licensee control. The TS are 

modified so that the stored diesel fuel 

oil and lube oil inventory will require 

that a 7-day supply be available for 

either diesel generator. Condition A and 

Condition B in the Action table are 

revised and Surveillance Requirements 

(SR) 3.8.3.1 and 3.8.3.2 are revised to 

reflect the above change. 

The proposed changes also revise TS 

3.8.3 by reducing the Completion Time 

for Condition C. Condition C currently 

requires that an inoperable fuel transfer 

system associated with fuel oil transfer 

pump P–18A be restored to operable 

status within 15 hours. The proposed 

TS change reduces the Completion Time 

for this Required Action from 15 to 12 

hours. The Completion Time is reduced 

to reflect the amount of time that an 

emergency diesel generator fuel oil day 

tank can support emergency diesel 

generator operation under design 

conditions. 

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 relocates the volume 

of diesel fuel oil and lube oil required to 

support 7-day operation of the onsite 

emergency diesel generators, and the volume 

equivalent to a 6-day supply, to licensee 

control. The specific volume of fuel oil 

equivalent to a 7-day and 6-day supply is 

calculated using the NRC approved 

methodology described in Regulatory Guide 

1.137, Revision 1, ‘‘Fuel Oil Systems for 

Standby Emergency diesel generators’’ and 

ANSI N195–1976, ‘‘Fuel Oil Systems for 

Standby Diesel Generators.’’ The specific 

volume of lube oil equivalent to a 7-day and 

6-day supply is based on the emergency 

diesel generator manufacturer’s consumption 

values for the run time of the diesel 

generator. Because the requirement to 

maintain a 7-day supply of diesel fuel oil and 

lube oil is not changed and is consistent with 

the assumptions in the accident analyses, 

and the actions taken when the volume of 

fuel oil and lube oil are less than a 6-day 

supply have not changed, neither the 

probability or the consequences of any 

accident previously evaluated will be 

affected. 

The proposed change also reduces the 

Completion Time for TS 3.8.3 Condition C 

for an inoperable P–18A fuel transfer system 

from 15 hours to 12 hours. Reducing the 

Completion Time to 12 hours bounds the 

13.5-hour time duration that the emergency 

diesel generator day tank will support 

emergency diesel generator operation under 

accident loading conditions. The change in 

Completion Time does not affect required TS 

actions if the Completion Time is exceeded. 

The Completion Time change does not affect 

the probability or consequences of an 

accident previously evaluated. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed fuel oil and lube oil 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. The change does not alter 

assumptions made in the safety analysis but 

ensures that the emergency diesel generator 

operates as assumed in the accident analysis. 

The proposed change is consistent with the 

safety analysis assumptions. 

The proposed change also reduces the 

Completion Time for TS 3.8.3 Condition C 

for an inoperable P–18A fuel transfer system 

from 15 hours to 12 hours. This change does 

not involve a physical alteration of the plant 

(i.e., no new or different type of equipment 

will be installed). This change does not 

create a condition in which a new or 

different kind of accident can occur. It does 

not alter assumptions made in the safety 

analysis. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change relocates the volume 

of fuel oil and lube oil required to support 

7-day operation of either emergency diesel 

generator, and the volume equivalent to a 6- 

day supply, to licensee control. As the bases 

for the existing limits on diesel fuel oil and 

lube oil are not changed, no change is made 

to the accident analysis assumptions and no 

margin of safety is reduced as part of this 

change. 

The proposed change also reduces the 

Completion Time for TS 3.8.3 Condition C 

for an inoperable P–18A fuel transfer system 

from 15 hours to 12 hours. There are no 

adverse affects on margins of safety since a 

more stringent operability requirement will 

be applied to the P–18A fuel transfer system. 

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: Mr. William 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 440 

Hamilton Ave., White Plains, NY 10601. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Exelon Generation Company, LLC, and 

PSEG Nuclear, LLC, Docket Nos. 50–277 

and 50–278, Peach Bottom Atomic 

Power Station (PBAPS), Units 2 and 3, 

York and Lancaster Counties, 

Pennsylvania 

Date of amendment request: March 

24, 2010, as supplemented by letter 

dated July 23, 2010. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) Section 

3.1.7, ‘‘Standby Liquid Control (SLC) 

System,’’ to extend the completion time 

for Condition C (i.e., two SLC 

subsystems inoperable for reasons other 

than Condition A) from 8 hours to 72 

hours. 

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 (NSHC), 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 revises 

Technical Specification (TS) 3.1.7, ‘‘Standby 

Liquid Control (SLC) System,’’ to extend the 

completion time (CT) for Condition C (i.e., 

‘‘Two SLC subsystems inoperable for reasons 

other than Condition A.’’) from eight hours to 

72 hours. 

The proposed change is based on a riskinformed evaluation performed in 

accordance with Regulatory Guides (RG) 

1.174, ‘‘An Approach for Using Probabilistic 

Risk Assessment in Risk-Informed Decisions 

On Plant-Specific Changes to the Licensing 

Basis,’’ and RG 1.177, ‘‘An Approach for 

Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications.’’ 

The proposed amendment modifies an 

existing CT for a dual-train SLC System 

inoperability. The condition evaluated, the 

action requirements, and the associated CT 

do not impact any initiating conditions for 

any accident previously evaluated. 

The proposed amendment does not 

increase postulated frequencies or the 

analyzed consequences of an Anticipated 

Transient Without Scram (ATWS). 

Requirements associated with 10 CFR 50.62 

will continue to be met. In addition, the 

proposed amendment does not increase 

postulated frequencies or the analyzed 

consequences of a large-break loss-of-coolant 

accident for which the SLC System is used 

for pH control. The new action requirement 

provides appropriate remedial actions to be 

taken in response to a dual-train SLC System 


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inoperability while minimizing the risk 

associated with continued operation. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

(2) Does the proposed amendment create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment revises TS 3.1.7 

to extend the CT for Condition C from eight 

hours to 72 hours. The proposed amendment 

does not involve any change to plant 

equipment or system design functions. This 

proposed TS amendment does not change the 

design function of the SLC System and does 

not affect the system’s ability to perform its 

design function. The SLC System provides a 

method to bring the reactor, at any time in 

a fuel cycle, from full power and minimum 

control rod inventory to a subcritical 

condition with the reactor in the most 

reactive xenon free state without taking 

credit for control rod movement. Required 

actions and surveillance requirements are 

sufficient to ensure that the SLC System 

functions are maintained. No new accident 

initiators are introduced by this amendment. 

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. 

The proposed amendment revises TS 3.1.7 

to extend the CT for Condition C from eight 

hours to 72 hours. The proposed amendment 

does not involve any change to plant 

equipment or system design functions. The 

margin of safety is established through the 

design of the plant structures, systems, and 

components, the parameters within which 

the plant is operated and the setpoints for the 

actuation of equipment relied upon to 

respond to an event. 

Safety margins applicable to the SLC 

System include pump capacity, boron 

concentration, boron enrichment, and system 

response timing. The proposed amendment 

does not modify these safety margins or the 

setpoints at which SLC is initiated, nor does 

it affect the system’s ability to perform its 

design function. In addition, the proposed 

change complies with the intent of the 

defense-in-depth philosophy and the 

principle that sufficient safety margins are 

maintained consistent with RG 1.177 

requirements (i.e., Section C, ‘‘Regulatory 

Position,’’ paragraph 2.2,‘‘Traditional 

Engineering Considerations’’). Therefore, the 

proposed amendment does not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves NSHC. 

Attorney for licensee: Mr. J. Bradley 

Fewell, Associate General Counsel, 

Exelon Generation Company LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

NextEra Energy Duane Arnold, LLC, 

Docket No. 50–331, Duane Arnold 

Energy Center, Linn County, Iowa 

Date of amendment request: August 

12, 2010. 

Description of amendment request: A 

change is proposed to the technical 

specifications to allow a delay time for 

entering a supported system technical 

specification (TS) when the 

inoperability is due solely to an 

unavailable barrier, if risk is assessed 

and managed consistent with the 

program in place for complying with the 

requirements of 10 CFR 50.65(a)(4). LCO 

3.0.9 will be added to individual TS 

providing this allowance. 

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 by affirming the 

applicability of the model analysis 

presented in the Federal Register notice 

dated October 3, 2006, starting on page 

71 FR 58452, which is presented below: 

Criterion 1: The Proposed Change Does Not 

Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated 

The proposed change allows a delay time 

for entering a supported system technical 

specification (TS) when the inoperability is 

due solely to an unavailable barrier if risk is 

assessed and managed. The postulated 

initiating events which may require a 

functional barrier are limited to those with 

low frequencies of occurrence, and the 

overall TS system safety function would still 

be available for the majority of anticipated 

challenges. Therefore, the probability of an 

accident previously evaluated is not 

significantly increased, if at all. The 

consequences of an accident while relying on 

the allowance provided by proposed LCO 

3.0.9 are no different than the consequences 

of an accident while relying on the TS 

required actions in effect without the 

allowance provided by proposed LCO 3.0.9. 

Therefore, the consequences of an accident 

previously evaluated are not significantly 

affected by this change. The addition of a 

requirement to assess and manage the risk 

introduced by this change will further 

minimize possible concerns. 

Therefore, this change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Criterion 2: The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Previously 

Evaluated 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed). 

Allowing delay times for entering supported 

system TS when inoperability is due solely 

to an unavailable barrier, if risk is assessed 

and managed, will not introduce new failure 

modes or effects and will not, in the absence 

of other unrelated failures, lead to an 

accident whose consequences exceed the 

consequences of accidents previously 

evaluated. The addition of a requirement to 

assess and manage the risk introduced by this 

change will further minimize possible 

concerns. 

Thus, this change does not create the 

possibility of a new or different kind of 

accident from an accident previously 

evaluated. 

Criterion 3: The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety 

The proposed change allows a delay time 

for entering a supported system TS when the 

inoperability is due solely to an unavailable 

barrier, if risk is assessed and managed. The 

postulated initiating events which may 

require a functional barrier are limited to 

those with low frequencies of occurrence, 

and the overall TS system safety function 

would still be available for the majority of 

anticipated challenges. The risk impact of the 

proposed TS changes was assessed following 

the three-tiered approach recommended in 

[Regulatory Guide] RG 1.177. A bounding 

risk assessment was performed to justify the 

proposed TS changes. This application of 

LCO 3.0.9 is predicated upon the licensee’s 

performance of a risk assessment and the 

management of plant risk. The net change to 

the margin of safety is insignificant as 

indicated by the anticipated low levels of 

associated risk (ICCDP and ICLERP) as shown 

in Table 1 of Section 3.1.1 in the [model] 

Safety Evaluation [on page 71 FR 58450 of 

the Federal Register dated October 3, 2006]. 

Therefore, this change does not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis, and based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. M. S. Ross, 

Florida Power & Light Company, P. O. 

Box 14000, Juno Beach, FL 33408–0420. 

NRC Branch Chief: Robert J. 

Pascarelli. 

South Carolina Electric and Gas 

Company (SCE and G), South Carolina 

Public Service Authority, Docket No. 

50–395, Virgil C. Summer Nuclear 

Station, Unit No. 1, Fairfield County, 

South Carolina 

Date of amendment request: 

November 11, 2010. 

Description of Amendment Request: 

The licensee proposes to amend the 

operating license for Virgil C. Summer 

Nuclear Station (VCSNS), by revising 


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the Technical Specifications (TS) and 

SCE&G proposes to provide surveillance 

enhancements that will improve 

operation and testing of the Emergency 

Diesel Generators (EDG). The changes 

will provide a more restrictive voltage 

and frequency band for operation when 

not connected in parallel with the 

offsite sources. 

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? 

No. 

The changes proposed by this license 

amendment will revise the Surveillance 

Requirements of Technical Specification 

3⁄4.8.1, AC SOURCES—OPERATING, to 

expand the continuous rated load 

specification to a range of 90% to 100% of 

the continuous rated load, specify an 

overload range of 105% to 110% of the 

continuous rated load, add a power factor 

limit while testing, allow gradual loading and 

unloading of the EDG, specify a maximum 

frequency for the overspeed limit, specify a 

maximum allowable overspeed voltage, and 

add a more restrictive voltage and frequency 

band for testing during steady state 

operation. 

The majority of these changes are being 

proposed in order to implement 

recommendations contained in [Institute of 

Nuclear Power Operations] INPO Significant 

Operating Experience Report (SOER) 03–01, 

Emergency Power Reliability, 

Recommendation Number 5, which 

recommends that the utility review testing 

practices for emergency power systems to 

verify that the practices are representative of 

actual demand conditions and appropriately 

exercise equipment that is expected to 

respond in an actual demand condition. 

These changes are based on the guidance 

provided by Regulatory Guide 1.9, Revision 

3, Selection, Design, Qualification, and 

Testing of Emergency Diesel Generator Units 

Used as Class 1E Onsite Electric Power 

Systems at Nuclear Power Plant. 

The more restrictive voltage and frequency 

band for testing during steady state operation 

is proposed to ease the impact of EDG voltage 

and frequency that are being incorporated 

into the Charging Pump performance 

requirements. The allowable voltage and 

frequency uncertainty limits for steady state 

operation are being reduced. This will ensure 

that the Charging Pumps continue to operate 

within their analyzed range. 

These changes do not affect the probability 

or consequences of an accident previously 

evaluated because the proposed changes do 

not make a change to any accident initiator, 

initiating condition, or assumption. The 

proposed changes do not involve a 

significant change to the plant design or 

operation. These changes do not invalidate 

assumptions used in evaluating the 

radiological consequences of an accident, do 

not alter the source term or containment 

isolation, and do not provide a new radiation 

release path or alter a potential radiological 

release. Therefore, the proposed change does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. 

These changes do not create the possibility 

of a new or different kind of accident from 

any accident previously evaluated because 

the proposed changes do not introduce a new 

or different accident initiator or introduce a 

new or different equipment failure mode or 

mechanism. 

No changes are being made in equipment 

hardware or software, operational 

philosophy, testing frequency, or how the 

system actually operates. Therefore, the 

proposed amendment will 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? 

No. 

These changes do not involve a significant 

reduction in a margin of safety because the 

proposed changes do not reduce the margin 

of safety that exists in the present Technical 

Specifications or Updated Final Safety 

Analysis Report. The operability 

requirements of the Technical Specifications 

are consistent with the initial condition 

assumptions of the safety analyses. The 

proposed changes do not affect the Action 

statement requirements for the various levels 

of degradation in the EDG. Therefore, the 

proposed change does not involve a 

significant reduction in a margin of safety. 

Based on the above, SCE&G concludes that 

the proposed amendment presents no 

significant hazards consideration under the 

standards set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: J. Hagood 

Hamilton, Jr., South Carolina Electric & 

Gas Company, Post Office Box 764, 

Columbia, South Carolina 29218. 

NRC Branch Chief: Gloria Kulesa. 

Southern Nuclear Operating Company, 

Inc. (SNC), Docket Nos. 50–348 and 50– 

364, Joseph M. Farley Nuclear Plant 

(FNP), Units 1 and 2, Houston County, 

Alabama 

Date of amendment request: October 

29, 2010. 

Description of amendment request: 

The proposed amendments request the 

adoption of an approved change to the 

standard technical specifications for 

Westinghouse Plants (NUREG–1431), to 

allow relocation of specific Technical 

Specifications (TS) surveillance 

frequencies to a licensee-controlled 

program. The proposed change is 

described in Technical Specification 

Task Force (TSTF) Traveler, TSTF–425, 

Revision 3, ‘‘Relocate Surveillance 

Frequencies to Licensee Control— 

RITSTF Initiative 5b’’ (Agencywide 

Documents Access and Management 

System (ADAMS) Accession No. 

ML080280275), and was described in 

the Notice of Availability published in 

the Federal Register (FR) on July 6, 

2009 (74 FR 31996). The proposed 

changes are consistent with NRCapproved TSTF–425, Revision 3. The 

proposed change relocates surveillance 

frequencies to a licensee-controlled 

program, the surveillance frequency 

control program. This change is 

applicable to licensees using 

probabilistic risk guidelines contained 

in NRC-approved [Nuclear Energy 

Institute] NEI 04–10, ‘‘Risk-Informed 

Technical Specifications Initiative 5b, 

Risk-Informed Method for Control of 

Surveillance Frequencies,’’ (ADAMS 

Accession No. 071360456). 

The licensee affirmed the 

applicability to the FNP of the model no 

significant hazards consideration 

determination provided in the FR on 

July 6, 2009 (74 FR 31996), in its 

application dated October 29, 2010. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

analysis of the issue of no significant 

hazards consideration is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change relocates the 

specified frequencies for periodic 

surveillance requirements to licensee control 

under a new Surveillance Frequency Control 

Program [SFCP]. Surveillance frequencies are 

not an initiator to any accident previously 

evaluated. As a result, the probability of any 

accident previously evaluated is not 

significantly increased. The systems and 

components required by the Technical 

Specifications for which the surveillance 

frequencies are relocated are still required to 

be operable, meet the acceptance criteria for 

the surveillance requirements, and be 

capable of performing any mitigation 

function assumed in the accident analysis. 

As a result, the consequences of any accident 

previously evaluated are not significantly 

increased. 


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Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

No new or different accidents result from 

utilizing the proposed change. The changes 

do not involve a physical alteration of the 

plant (i.e., no new or different type of 

equipment will be installed) or a change in 

the methods governing normal plant 

operation. In addition, the changes do not 

impose any new or different requirements. 

The changes do not alter assumptions made 

in the safety analysis. The proposed changes 

are consistent with the safety analysis 

assumptions and current plant operating 

practice. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in the margin of safety? 

Response: No. 

The design, operation, testing methods, 

and acceptance criteria for systems, 

structures, and components (SSCs), specified 

in applicable codes and standards (or 

alternatives approved for use by the NRC) 

will continue to be met as described in the 

plant licensing basis (including the final 

safety analysis report and bases to TS), since 

these are not affected by changes to the 

surveillance frequencies. Similarly, there is 

no impact to safety analysis acceptance 

criteria as described in the plant licensing 

basis. To evaluate a change in the relocated 

surveillance frequency, the licensee will 

perform a probabilistic risk evaluation using 

the guidance contained in NRC approved NEI 

04–10, Rev. 1, in accordance with the TS 

SFCP. NEI 04–10, Rev. 1, methodology 

provides reasonable acceptance guidelines 

and methods for evaluating the risk increase 

of proposed changes to surveillance 

frequencies consistent with Regulatory Guide 

1.177. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

Based upon the reasoning presented above, 

licensee concludes that the requested change 

does not involve a significant hazards 

consideration as set forth in 10 CFR 50.92(c), 

Issuance of Amendment. 

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: M. Stanford 

Blanton, Esq., Balch and Bingham, Post 

Office Box 306, 1710 Sixth Avenue 

North, Birmingham, Alabama 35201. 

NRC Branch Chief: Gloria J. Kulesa. 

Southern Nuclear Operating Company, 

Inc., Georgia Power Company, 

Oglethorpe Power Corporation, 

Municipal Electric Authority of Georgia, 

City of Dalton, Georgia, Docket Nos. 50– 

321 and 50–366, Edwin I. Hatch Nuclear 

Plant (HNP), Units 1 and 2, Appling 

County, Georgia 

Date of amendment request: October 

29, 2010. 

Description of amendment request: 

The proposed amendments request the 

adoption of an approved change to the 

standard technical specifications for 

General Electric Plants, BWR/4 

(NUREG–1433), to allow relocation of 

specific Technical Specification (TS) 

surveillance frequencies to a licenseecontrolled program. The proposed 

change is described in Technical 

Specification Task Force (TSTF) 

Traveler, TSTF–425, Revision 3, 

‘‘Relocate Surveillance Frequencies to 

Licensee Control—RITSTF Initiative 

5b.’’ (Agencywide Documents Access 

and Management System (ADAMS) 

Accession No. ML080280275), and was 

described in the Notice of Availability 

published in the Federal Register (FR) 

on July 6, 2009 (74 FR 31996). The 

proposed changes are consistent with 

NRC-approved TSTF–425, Revision 3. 

The proposed change relocates 

surveillance frequencies to a licenseecontrolled program, the surveillance 

frequency control program. This change 

is applicable to licensees using 

probabilistic risk guidelines contained 

in NRC-approved [Nuclear Energy 

Institute] NEI 04–10, ‘‘Risk-Informed 

Technical Specifications Initiative 5b, 

Risk-Informed Method for Control of 

Surveillance Frequencies,’’ (ADAMS 

Accession No. 071360456). The licensee 

affirmed the applicability to the HNP of 

the model no significant hazards 

consideration determination provided 

in the FR on July 6, 2009 (74 FR 31996) 

in its application dated October 29, 

2010. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

analysis of the issue of no significant 

hazards consideration is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change relocates the 

specified frequencies for periodic 

surveillance requirements to licensee control 

under a new Surveillance Frequency Control 

Program [SFCP]. Surveillance frequencies are 

not an initiator to any accident previously 

evaluated. As a result, the probability of any 

accident previously evaluated is not 

significantly increased. The systems and 

components required by the Technical 

Specifications for which the surveillance 

frequencies are relocated are still required to 

be operable, meet the acceptance criteria for 

the surveillance requirements, and be 

capable of performing any mitigation 

function assumed in the accident analysis. 

As a result, the consequences of any accident 

previously evaluated are not significantly 

increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

No new or different accidents result from 

utilizing the proposed change. The changes 

do not involve a physical alteration of the 

plant (i.e., no new or different type of 

equipment will be installed) or a change in 

the methods governing normal plant 

operation. In addition, the changes do not 

impose any new or different requirements. 

The changes do not alter assumptions made 

in the safety analysis. The proposed changes 

are consistent with the safety analysis 

assumptions and current plant operating 

practice. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in the margin of safety? 

Response: No. 

The design, operation, testing methods, 

and acceptance criteria for systems, 

structures, and components (SSCs), specified 

in applicable codes and standards (or 

alternatives approved for use by the NRC) 

will continue to be met as described in the 

plant licensing basis (including the final 

safety analysis report and bases to TS), since 

these are not affected by changes to the 

surveillance frequencies. Similarly, there is 

no impact to safety analysis acceptance 

criteria as described in the plant licensing 

basis. To evaluate a change in the relocated 

surveillance frequency, SNC will perform a 

probabilistic risk evaluation using the 

guidance contained in NRC approved NEI 

04–10, Rev. 1, in accordance with the TS 

SFCP. NEI 04–10, Rev.1, methodology 

provides reasonable acceptance guidelines 

and methods for evaluating the risk increase 

of proposed changes to surveillance 

frequencies consistent with Regulatory Guide 

1.177. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

Based upon the reasoning presented above, 

licensee concludes that the requested change 

does not involve a significant hazards 

consideration as set forth in 10 CFR 50.92(c), 

Issuance of Amendment. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 


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satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: M. Stanford 

Blanton, Esq., Balch and Bingham, Post 

Office Box 306, 1710 Sixth Avenue 

North, Birmingham, Alabama 35201. 

NRC Branch Chief: Gloria J. Kulesa. 

Tennessee Valley Authority, Docket 

Nos. 50–259, 50–260 and 50–296, 

Browns Ferry Nuclear Plant, Units 1, 2 

and 3, Limestone County, Alabama 

Date of amendment request: February 

18, 2010, as supplemented on November 

12, 2010 (TS–468). 

Description of amendment request: 

The proposed amendment would 

modify Technical Specification 3.8.1 to 

extend the completion time (CT) for the 

return of an inoperable emergency 

diesel generator (DGs) to operable status 

from 7 days to 14 days, based on the 

availability of two non-safety related 

temporary diesel generators (TDGs). 

Commensurate changes to the maximum 

completion times were also proposed, 

extending the times from 14 to 21 days 

in Required Actions A.3 and B.4. The 

change also eliminates a historical 

footnote for a previous CT for Unit 3 

only that is no longer needed. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes do not affect the 

design of the DGs, the operational 

characteristics or function of the DGs, the 

interfaces between the DGs and other plant 

systems, or the reliability of the DGs. 

Required Actions and their associated CTs 

are not considered initiating conditions for 

any UFSAR [updated final safety analysis 

report] accident previously evaluated, nor are 

the DGs considered initiators of any 

previously evaluated accidents. The DGs are 

provided to mitigate the consequences of 

previously evaluated accidents, including a 

loss of off-site power. 

The consequences of previously evaluated 

accidents will not be significantly affected by 

the extended DG CT, because a sufficient 

number of onsite Alternating Current [AC] 

power sources will continue to remain 

available to perform the accident mitigation 

functions associated with the DGs, as 

assumed in the accident analyses. In 

addition, as a risk mitigation and defense-indepth action, an independent AC power 

source, via two available TDGs, will be 

available to support the ESF [engineered 

safety feature] bus with the inoperable DG 

during a SBO [station blackout]. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

change in the permanent design, 

configuration, or method of operation of the 

plant. The proposed changes will not alter 

the manner in which equipment operation is 

initiated, nor will the functional demands on 

credited equipment be changed. The 

proposed changes allow operation of the unit 

to continue while a DG is repaired and 

retested with the TDGs in standby to mitigate 

a SBO event. The proposed extensions do not 

affect the interaction of a DG with any system 

whose failure or malfunction can initiate an 

accident. As such, no new failure modes are 

being introduced. Therefore, the proposed 

changes do not create the possibility of a new 

or different kind of accident from any 

accident previously evaluated. 

3. Does the proposed amendment involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes do not alter the 

permanent plant design, including 

instrument set points, nor does it change the 

assumptions contained in the safety analyses. 

The standby TDG alternate AC system is 

designed with sufficient redundancy such 

that a DG may be removed from service for 

maintenance or testing. The remaining seven 

DGs are capable of carrying sufficient 

electrical loads to satisfy the UFSAR 

requirements for accident Mitigation or unit 

safe shutdown. The proposed changes do not 

impact the redundancy or availability 

requirements of offsite power supplies or 

change the ability of the plant to cope with 

station blackout events. Therefore, the 

proposed changes do not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, 6A West 

Tower, Knoxville, Tennessee 37902. 

NRC Branch Chief: Douglas A. 

Broaddus. 

mitigation or unit safe shutdown. The 

proposed changes do not impact the 

redundancy or availability requirements 

of offsite power supplies or change the 

ability of the plant to cope with station 

blackout events. Therefore, the 

proposed changes do not involve a 

significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, 6A West 

Tower, Knoxville, Tennessee 37902. 

NRC Branch Chief: Douglas A. 

Broaddus. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: October 

21, 2010. 

Description of amendment request: 

The proposed amendment would 

correct a typographical error in Section 

5, Administrative Controls, of the 

Technical Specifications (TSs). The 

current TSs, on page 5.0–31, has two 

paragraphs numbered as 5.7.2d.3. The 

amendment proposes to renumber the 

second paragraph as 5.7.2d.4. The 

typographical error was introduced in 

Amendment No. 123 issued on March 

31, 1999. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change is administrative in 

nature. The change involves correcting a 

typographical error. This change does not 

affect possible initiating events for accidents 

previously evaluated or alter the 

configuration or operation of the facility. The 

Limiting Safety System Settings and Safety 

Limits specified in the TS remain unchanged. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

The proposed change is administrative in 

nature. The safety analysis of the facility 

remains complete and accurate. There are no 

physical changes to the facility and the plant 

conditions for which the design basis 

accidents have been evaluated are still valid. 

The operating procedures and emergency 

procedures are unaffected. Consequently no 

new failure modes are introduced as a result 

of the proposed change. 


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77918 Federal Register / Vol. 75, No. 239 / Tuesday, December 14, 2010 / Notices 

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 is 

administrative in nature. Since there 

[are] no changes to the operation of the 

facility or the physical design, the 

Updated Safety Analysis Report (USAR) 

design basis, accident assumptions, or 

TS Bases are not affected. 

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: Jay Silberg, Esq., 

Pillsbury Winthrop Shaw Pittman LLP, 

2300 N Street, NW., Washington, DC 

20037. 

NRC Branch Chief: Michael T. 

Markley. 

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. 

Indiana Michigan Power Company 

(IandM), Docket Nos. 50–315 and 50– 

316, Donald C. Cook Nuclear Plant, 

Units 1 and 2, Berrien County, Michigan 

Date of application for amendment: 

September 8, 2010. 

Brief description of amendment: The 

licensee proposed to delete the 

Technical Specification requirements 

related to the containment hydrogen 

recombiners and the hydrogen monitors, 

in accordance with Nuclear Energy 

Institute Technical Specification Task 

Force (TSTF) initiative designated as 

TSTF–447. 

Date of publication of individual 

notice in Federal Register: October 14, 

2010 (75 FR 63209). 

Expiration date of individual notice: 

December 13, 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 

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, Room O1–F21, 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: 

November 30, 2009, as supplemented by 

letter dated July 22, 2010. 

Brief description of amendment: The 

amendments revised Table 3.3.5–1 of 

Technical Specification (TS) 3.3.5, 

‘‘Engineered Safety Features Actuation 

System (ESFAS) Instrumentation,’’ to 

raise the refueling water tank (RWT) low 

level allowable values for the 

recirculation actuation signal; raised the 

minimum required RWT volume shown 

in TS Figure 3.5.5–1 of TS 3.5.5, 

‘‘Refueling Water Tank (RWT)’’; and 

implemented a time-critical operator 

action to close the RWT isolation valves, 

including consideration of a potentially 

more limiting single failure of a lowpressure safety injection pump to 

automatically stop, as designed, on a 

recirculation actuation signal. 

Date of issuance: November 24, 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—182; Unit 

2—182; Unit 3—182. 

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: April 20, 2010 (75 FR 20629). 

The supplemental letter dated 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 amendments is contained in a 

Safety Evaluation dated November 24, 

2010. 

No significant hazards consideration 

comments received: No. 

Calvert Cliffs Nuclear Power Plant, LLC, 

Docket Nos. 50–317 and 50–318, Calvert 

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of application for amendments: 

April 5, 2010. 

Brief description of amendments: The 

amendment made title changes and 

corrections within Technical 

Specification (TS) 5.0, ‘‘Administrative 


srobinson on DSKHWCL6B1PROD with NOTICES
Federal Register / Vol. 75, No. 239 / Tuesday, December 14, 2010 / Notices 77919 

Controls.’’ Specifically, the changes 

included: 

(1) Replacement of the use of plant 

specific titles to generic titles consistent 

with TS Task Force (TSTF) Traveler 

TSTF–65, Revision 1, ‘‘Use of Generic 

Titles for Utility Positions,’’ 

(2) Changes made to more closely 

align selected TSs with the Improved 

Standard TSs, and 

(3) Administrative changes to 

specified TSs. 

Date of issuance: November 22, 2010. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment Nos.: 296 for Unit 1 and 

272 for Unit 2. 

Renewed Facility Operating License 

Nos. DPR–53 and DPR–69: Amendments 

revised the License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: June 1, 2010 (75 FR 30443). 

The Commission’s related evaluation 

of these amendments is contained in a 

Safety Evaluation dated November 22, 

2010. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 2nd day 

of December 2010. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2010–31086 Filed 12–13–10; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket No. 50–400; NRC–2010–0020] 

Carolina Power & Light Company 

Shearon Harris Nuclear Power Plant, 

Unit 1; Environmental Assessment and 

Finding of No Significant Impact 

The U.S. Nuclear Regulatory 

Commission (NRC, the Commission) is 

considering issuance of an exemption, 

pursuant to title 10 of the Code of 

Federal Regulations (10 CFR) section 

73.5, ‘‘Specific exemptions,’’ from the 

implementation date for certain 

requirements of 10 CFR part 73, 

‘‘Physical protection of plants and 

materials,’’ for Renewed Facility 

Operating License No. NPF–63, issued 

to Carolina Power & Light Company (the 

licensee), now doing business as 

Progress Energy Carolinas, Inc., for 

operation of the Shearon Harris Nuclear 

Power Plant (HNP), Unit 1, located in 

New Hill, North Carolina. In accordance 

with 10 CFR 51.21, the NRC staff 

prepared an environmental assessment 

documenting its finding. The NRC staff 

concluded that the proposed actions 

will have no significant environmental 

impact. 

Environmental Assessment 

Identification of the Proposed Action 

The proposed action would exempt 

the licensee from the required 

implementation date of March 31, 2010, 

for one specific requirement of 10 CFR 

part 73. Specifically, HNP, Unit 1 would 

be granted a second exemption, further 

extending the date for compliance with 

one remaining item of the requirements 

contained in 10 CFR 73.55, from 

December 15, 2010, (the date specified 

in a prior exemption granted by NRC on 

February 24, 2010), until November 30, 

2011. The proposed action, an extension 

of the schedule for completion of certain 

actions required by the revised 10 CFR 

part 73, does not result in any 

additional physical changes to the 

reactor, fuel, plant structures, support 

structures, water, or land at the HNP, 

Unit 1 site. 

The proposed action is in accordance 

with the licensee’s application dated 

September 20, 2010. 

The Need for the Proposed Action 

The proposed exemption is needed to 

provide the licensee with additional 

time, beyond the date granted by the 

NRC letter dated February 24, 2010, to 

implement one remaining item of the 

three requirements in the previous 

exemption that involves important 

physical modifications to the HNP, Unit 

1 security system. There are several 

issues which have delayed the work to 

this point, and/or impacted the 

projected schedule, such as the 

existence of safety-related conduit and 

dedicated safe shut down (SSD) 

equipment of HNP, Unit 1 within the 

room in which some important security 

modifications are planned. A direct 

outside access route to the physical 

construction area has not been available 

due to design basis tornado and missile 

considerations for the safety related 

conduits and SSD equipment. These 

issues were revealed as the design 

evolved from the conceptual state to a 

detailed design state. Presently, the 

licensee is pursuing a design solution 

that will allow both temporary and 

ultimately permanent direct outside 

access to the area. Additional time, 

beyond that previously approved, is 

needed due the extensive redesign and 

review effort that was unforeseen at the 

conceptual design stage. 

Environmental Impacts of the Proposed 

Action 

The NRC staff has completed its 

environmental assessment of the 

proposed exemption. The NRC staff has 

concluded that the proposed action to 

further extend the implementation 

deadline for one item would not 

significantly affect plant safety and 

would not have a significant adverse 

effect on the probability of an accident 

occurring. 

The proposed action would not result 

in an increased radiological hazard 

beyond those hazards previously 

analyzed in the environmental 

assessment and final finding of no 

significant impact made by the 

Commission in promulgating its 

revisions to 10 CFR part 73 as discussed 

in a Federal Register notice dated 

March 27, 2009 (74 FR 13926). There 

will be no change to radioactive 

effluents that affect radiation exposures 

to plant workers and members of the 

public. Therefore, no changes or 

different types of radiological impacts 

are expected as a result of the proposed 

exemption. 

The proposed action does not result 

in changes to land use or water use, or 

result in changes to the quality or 

quantity of non-radiological effluents. 

No changes to the National Pollution 

Discharge Elimination System permit 

are needed. No effects on the aquatic or 

terrestrial habitat in the vicinity of the 

plant, or to threatened, endangered, or 

protected species under the Endangered 

Species Act, or impacts to essential fish 

habitat covered by the MagnusonStevens Act are expected. There are no 

impacts to the air or ambient air quality. 

There are no impacts to historical and 

cultural resources. There would be no 

impact to socioeconomic resources. 

Therefore, no changes to or different 

types of non-radiological environmental 

impacts are expected as a result of the 

proposed exemption. 

Accordingly, the NRC concludes that 

there are no significant environmental 

impacts associated with the proposed 

action. 

With its request to extend the 

implementation deadline, the licensee 

currently maintains a security system 

acceptable to the NRC and that will 

continue to provide acceptable physical 

protection of HNP, Unit 1 in lieu of the 

new requirements in 10 CFR part 73. 

Therefore, the extension of the 

implementation date for one element of 

the new requirements of 10 CFR part 73 

to November 30, 2011, would not have 

any significant environmental impacts. 

The NRC staff’s safety evaluation will 

be provided in the exemption that will 


srobinson on DSKHWCL6B1PROD with NOTICES