Federal Register / Vol. 75, No. 229 / Tuesday, November 30, 2010 / Notices 74091 

19. National Archives and Records 

Administration, Office of the General 

Counsel (DAA–64–2010–5, 1 item, 1 

temporary item). Master files of an 

electronic information system used to 

track garnishment actions. 

20. National Mediation Board, 

Agency-wide (N1–13–11–1, 4 items, 1 

temporary item). Records of the 

Presidential Emergency Boards, 

including routine case notes. Proposed 

for permanent retention are official case 

files and reports. 

Dated: November 24, 2010. 

Michael J. Kurtz, 

Assistant Archivist for Records Services— 

Washington, DC. 

[FR Doc. 2010–30216 Filed 11–29–10; 8:45 am] 

BILLING CODE 7515–01–P 

NATIONAL ARCHIVES AND RECORDS 

ADMINISTRATION 

Senior Executive Service (SES) 

Performance Review Board; Members 

AGENCY: National Archives and Records 

Administration. 

ACTION: Notice; SES Performance 

Review Board. 

SUMMARY: Notice is hereby given of the 

appointment of members of the National 

Archives and Records Administration 

(NARA) Performance Review Board. 

DATES: Effective Date: This appointment 

is effective on November 30, 2010. 

FOR FURTHER INFORMATION CONTACT: 

Pamela S. Pope, Human Resources 

Services Division (NAH), National 

Archives and Records Administration, 

9700 Page Avenue, St. Louis, MO 63132, 

(314) 801–0882. 

SUPPLEMENTARY INFORMATION: Section 

4314(c) of Title 5, U.S.C., requires each 

agency to establish, in accordance with 

regulations prescribed by the Office of 

Personnel Management, one or more 

SES Performance Review Boards. The 

Board shall review the initial appraisal 

of a senior executive’s performance by 

the supervisor and recommend final 

action to the appointing authority 

regarding matters related to senior 

executive performance. 

The members of the Performance 

Review Board for the National Archives 

and Records Administration are: 

Adrienne C. Thomas, Deputy Archivist 

of the United States, Michael J. Kurtz, 

Assistant Archivist for Records 

Services—Washington, DC, Thomas E. 

Mills, Assistant Archivist for Regional 

Records Services, and Miriam M. 

Nisbet, Director, Office of Government 

Information Services. These 

appointments supersede all previous 

appointments. 

Dated: November 23, 2010. 

David S. Ferriero, 

Archivist of the United States. 

[FR Doc. 2010–30064 Filed 11–29–10; 8:45 am] 

BILLING CODE 7515–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0367] 

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 (NRC or the Commission) 

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 3, 

2010, to November 17, 2010. The last 

biweekly notice was published on 

November 16, 2010 (75 FR 70032). 

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), 50.92, this means 

that operation of the facility in 

accordance with the proposed 

amendment would not (1) Involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated; or (2) create the possibility of 

a new or different kind of accident from 

any accident previously evaluated; or 

(3) involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 

60-day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rules, 

Announcements and Directives Branch 

(RADB), TWB–05–B01M, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be faxed to the RADB at 301–492– 

3446. Documents may be examined, 

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

Public Document Room (PDR), located 

at One White Flint North, Room 

O1–F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland 20852. 

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 20854. 

Publicly available records will be 

accessible from the Agencywide 


jdjones on DSK8KYBLC1PROD with NOTICES
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Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed by the above 

date, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the requestor/ 

petitioner seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the requestor/petitioner shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the requestor/petitioner 

intends to rely in proving the contention 

at the hearing. The requestor/petitioner 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the requestor/petitioner intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the requestor/ 

petitioner to relief. A requestor/ 

petitioner who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

All documents filed in NRC 

adjudicatory proceedings, including a 

request for hearing, a petition for leave 

to intervene, any motion or other 

document filed in the proceeding prior 

to the submission of a request for 

hearing or petition to intervene, and 

documents filed by interested 

governmental entities participating 

under 10 CFR 2.315(c), must be filed in 

accordance with the NRC E-Filing rule 

(72 FR 49139, August 28, 2007). The EFiling process requires participants to 

submit and serve all adjudicatory 

documents over the internet, or in some 

cases to mail copies on electronic 

storage media. Participants may not 

submit paper copies of their filings 

unless they seek an exemption in 

accordance with the procedures 

described below. 

To comply with the procedural 

requirements of E-Filing, at least ten 

(10) days prior to the filing deadline, the 

participant should contact the Office of 

the Secretary by e-mail at 

hearing.docket@nrc.gov, or by telephone 

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

ID certificate, which allows the 

participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and (2) advise the 

Secretary that the participant will be 

submitting a request or petition for 

hearing (even in instances in which the 

participant, or its counsel or 

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

this information, the Secretary will 

establish an electronic docket for the 

hearing in this proceeding if the 

Secretary has not already established an 

electronic docket. 

Information about applying for a 

digital ID certificate is available on 

NRC’s public Web site at http:// 

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

apply-certificates.html. System 

requirements for accessing the 

E-Submittal 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 the 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 


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Federal Register / Vol. 75, No. 229 / Tuesday, November 30, 2010 / Notices 74093 

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, Public File Area 

O1F21, 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. 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of amendment request: 

September 24, 2010. 

Description of amendment request: 

The proposed amendment would revise 

the Fermi 2 Radiological Emergency 

Response Preparedness (RERP) Plan to 

increase the staff augmentation times for 

Technical Support Center-related 

functions from 30 to 60 minutes and for 

Emergency Operations Facility-related 

functions from 60 to 90 minutes. 

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. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

The proposed extension of staff 

augmentation times has no effect on normal 

plant operation or on any accident initiator. 

The change affects the response to 

radiological emergencies under the Fermi 2 

Radiological Emergency Response 

Preparedness (RERP) Plan. The ability of the 

emergency response organization to respond 

adequately to radiological emergencies has 

been evaluated. Improvements have been 

made to equipment, procedures, and training 

since initial approval of the Fermi 2 

Emergency Plan that have resulted in a 

significant increase in the on-shift 

capabilities and knowledge such that there 

would be no degradation or loss of 

Emergency Plan function as a result of the 

proposed change. A functional analysis was 

also performed on the effect of the proposed 

change on the timeliness of performing major 

tasks for the major functional areas of the 

RERP Plan. The analysis concluded that 

extension of staff augmentation times would 

not significantly affect the ability to perform 

the required tasks. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed change affects the required 

response times for supplementing onsite 

personnel in response to a Radiological 

emergency. It has been evaluated and 

determined not to significantly affect the 

ability to perform that function. It has no 

effect on the plant design or on the normal 

operation of the plant and does not affect 

how the plant is physically operated under 

emergency conditions. The extension of staff 

augmentation times in the RERP Plan does 

not affect the plant Operating, Abnormal 

Operating, or Emergency Operating 

procedures which are performed by plant 

staff during all plant conditions. 

Therefore, since the proposed change does 

not affect the design or method of operation 

of the plant, it does not create the possibility 

of a new or different kind of accident from 

any accident previously evaluated. 

3. The proposed change does not involve 

a significant reduction in the margin of 

safety. 

The proposed change does not affect plant 

design or method of operation. 10 CFR 

50.47(b) and 10 CFR Part 50, Appendix E 

establish emergency planning standards that 

require adequate staffing, satisfactory 

performance of key functional areas and 

critical tasks; and timely augmentation of the 

response capability. Since the initial NRC 

approval of the Emergency Plan, there have 

been improvements in the technology used to 

support the RERP functions and in the 

capabilities of onsite personnel. A functional 

analysis was performed on the effect of the 

proposed change on the timeliness of 

performing major tasks for the functional 

areas of the RERP Plan. The analysis 

concluded that an increase in staff 

augmentation times would not significantly 

affect the ability to perform the required 

RERP tasks. Thus, the proposed change has 

been determined not to adversely affect the 

ability to meet the emergency planning 

standards as described in 10 CFR 50.47(b) 

and 10 CFR Part 50, Appendix E. 

Therefore, the proposed change will not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 


jdjones on DSK8KYBLC1PROD with NOTICES
74094 Federal Register / Vol. 75, No. 229 / Tuesday, November 30, 2010 / Notices 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David G. 

Pettinari, Attorney—Corporate Matters, 

688 WCB, Detroit Edison Company, One 

Energy Plaza, Detroit, Michigan 48226– 

1279. 

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: January 

6, 2010, as supplemented by letters 

dated August 20, 2010, and October 14, 

2010. 

Description of amendment request: 

The proposed amendment would enable 

PBAPS, Units 2 and 3, to possess 

byproduct and special nuclear material 

from Limerick Generating Station (LGS), 

Units 1 and 2. Specifically, the revised 

license paragraph would permit storage 

of low-level radioactive waste (LLRW) 

from LGS in the PBAPS LLRW Storage 

Facility. The PBAPS LLRW Storage 

Facility currently provides storage for 

LLRW generated at PBAPS. 

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 change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change is an amendment to 

the PBAPS FOLs [Facility Operating 

Licenses] that will enable PBAPS to receive 

and store Class B/C LLRW from LGS in the 

PBAPS LLRWSF [Low Level Radioactive 

Waste Storage Facility]. This proposed 

change does not impact any initiators or 

precursors of previously analyzed accidents. 

The storage of Class B/C LLRW from LGS 

does not impact the failure of any plant 

structures, systems, or components. The 

proposed change does not have a detrimental 

impact on the integrity of any plant structure, 

system, or component that initiates an 

analyzed event. The proposed change does 

not affect any active or passive failure 

mechanisms that could lead to an accident. 

The PBAPS LLRWSF is not safety related, 

and is not used for plant shutdown resulting 

from accident or nonstandard operational 

conditions. 

The proposed change does not significantly 

increase the consequences of postulated 

design basis events (i.e., seismic, flood, 

tornado, fire, and container drop events), in 

that the postulated impact of these events 

remains well below regulatory requirements 

(i.e., less than 10 percent of 10 CFR Part 100, 

‘‘Reactor Site Criteria’’ acceptance criteria). 

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 previously evaluated? 

Response: No. 

The proposed change is an amendment to 

the PBAPS FOLs that will enable PBAPS to 

receive and store Class B/C LLRW from LGS 

in the PBAPS LLRWSF. The proposed 

amendment does not involve any change to 

the plant equipment or system design 

functions. EGC has verified that the storage 

of Class B/C LLRW from LGS in the PBAPS 

LLRWSF does not affect the ability of the 

PBAPS LLRWSF to perform its design 

function, including compliance with NRC 

regulatory requirements and guidance. 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 change involve a 

significant reduction in the margin of safety? 

Response: No. 

The proposed change is an amendment to 

the PBAPS FOLs that will enable PBAPS to 

receive and store Class B/C LLRW from LGS 

in the PBAPS LLRWSF. 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. The 

proposed amendment does not affect the 

PBAPS safety limits or setpoints at which 

protective actions are initiated. 

The proposed amendment does not 

significantly increase the dose rate at the 

exterior wall of the LLRWSF, the nearest 

restricted area boundary, and the nearest 

residence when the LLRWSF is filled to 

capacity with Class B/C LLRW. Therefore, 

these dose rates will remain within limits 

specified in 10 CFR Part 20 and 40 CFR Part 

190. 

Additionally, the potential radiological 

impact of a postulated design basis container 

drop accident is less than 10 percent of the 

10 CFR Part 100 acceptance criteria. 

Therefore the margin of safety is not 

reduced by the proposed change. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves 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. 

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: June 25, 

2010, as supplemented by letter dated 

August 16, 2010. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 

Surveillance Requirement (SR) 3.6.1.3, 

‘‘Primary Containment Isolation Valves 

(PCIVs),’’ and SR 3.6.1.5, ‘‘Reactor 

Building-to-Suppression Chamber 

Vacuum Breakers,’’ to modify the 

required level for the liquid nitrogen 

storage tank. 

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 TS changes to increase the 

level in the liquid nitrogen storage tank from 

≥ 16 inches water column to a level of ≥ 22 

inches water column, or equivalent volume 

of ≥ 124,000 scf [standard cubic feet] at 250 

psig, is necessary in order to correct a nonconservative TS value. Increasing the level is 

intended to ensure continued operability of 

the PCIVs (SR 3.6.1.3.1) and Reactor 

Building-to-Suppression Chamber Vacuum 

Breakers (SR 3.6.1.5.1) via the SGIG [safety 

grade instrument gas] system. The nonconservative TS condition was identified 

based on a re-analysis of the liquid nitrogen 

storage tank operation. The leakage 

allowance that was previously assumed was 

not based on a rigorous empirical value. The 

re-analysis of the leakage allowance assumes 

more reasonable system leakage based on 

operational data. Exelon determined that the 

current PBAPS, Units 2 and 3, TS SR value 

for the minimum level in the liquid nitrogen 

storage tank of ≥ 16 inches water column is 

non-conservative and that the guidance of 

Nuclear Regulatory Commission (NRC) 

Administrative Letter 98–10, ‘‘Dispositioning 

of Technical Specifications that are 

Insufficient to Assure Plant Safety,’’ applies. 

Exelon has implemented administrative 

controls to maintain the amount of nitrogen 

in the liquid nitrogen storage tank at a level 

of > 22 inches water column in support of 

SGIG system operation. 

Exelon is submitting this License 

Amendment Request to address this nonconservative condition. The proposed TS 


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Federal Register / Vol. 75, No. 229 / Tuesday, November 30, 2010 / Notices 74095 

changes do not introduce new equipment or 

new equipment operating modes, nor do the 

proposed changes alter existing system 

relationships. The proposed changes do not 

affect plant operation, design function or any 

analysis that verifies the capability of a 

system, structure or component (SSC) to 

perform a design function. Further, the 

proposed changes do not increase the 

likelihood of the malfunction of any SSC or 

impact any analyzed accident. Consequently, 

the probability or consequences of an 

accident previously evaluated are not 

affected. 

Therefore, the proposed amendments 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 TS change to increase the 

level in the liquid nitrogen storage tank from 

≥ 16 inches water column to a level of ≥ 22 

inches water column, or equivalent volume 

of ≥ 124,000 scf at 250 psig, for the PCIVs (SR 

3.6.1.3.1) and Reactor Building-toSuppression Chamber Vacuum Breakers (SR 

3.6.1.5.1) is needed to correct a nonconservative value based on a revised 

analysis. The proposed TS changes do not 

alter the design function or operation of any 

SSC. There is no new system component 

being installed, no construction of a new 

facility, and no performance of a new test or 

maintenance function. The proposed TS 

changes do not create the possibility of a new 

credible failure mechanism or malfunction. 

The proposed changes do not modify the 

design function or operation of any SSC. 

Further, the proposed changes do not 

introduce new accident initiators. 

Consequently, the proposed changes cannot 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

Therefore, the proposed amendments do 

not create the possibility of a new or different 

kind of accident from any accident 

previously analyzed. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed TS changes to increase the 

level in the liquid nitrogen storage tank from 

≥ 16 inches water column to a level of ≥ 22 

inches water column, or equivalent volume 

of ≥ 124,000 scf at 250 psig, for the PCIVs (SR 

3.6.1.3.1) and Reactor Building-toSuppression Chamber Vacuum Breakers (SR 

3.6.1.5.1) are necessary to correct an existing 

non-conservative TS value. The proposed TS 

changes are needed based on a revised 

analysis that utilizes empirical data for 

nitrogen system uses and losses. The 

proposed changes do not exceed or alter a 

design basis or a safety limit for a parameter 

established in the PBAPS, Units 2 and 3, 

Updated Final Safety Analysis Report 

(UFSAR) or the PBAPS, Units 2 and 3, 

Renewed Facility Operating License (FOL). 

Consequently, the proposed changes do not 

result in a reduction in the margin of safety. 

Therefore, the proposed amendments 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 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. 

Exelon Generation Company, LLC, 

Docket No. 50–289, Three Mile Island 

Nuclear Station, Unit 1, Dauphin 

County, Pennsylvania 

Date of amendment request: 

September 22, 2010. 

Description of amendment request: 

The proposed amendment would 

relocate the list of pumps, fans, and 

valves in Technical Specification (TS) 

4.5.1.1b, Sequence and Power Transfer 

Test, to the Three Mile Island, Unit 1 

(TMI–1) Updated Final Safety Analysis 

Report. In addition, TS 4.5.1.2b, TS 

4.5.2.2a, and TS 4.5.2.2b refer to this 

test and are proposed for revision to 

reflect the proposed change to TS 

4.5.1.1b. 

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 

with an NRC edit in brackets: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of any accident previously 

evaluated? 

Response: No. 

The proposed license amendment does not 

add, delete or modify plant equipment. The 

proposed changes are administrative in 

nature. The proposed amendment would 

relocate the list of pumps, fans and valves in 

Technical Specification (TS) 4.5.1.1b, 

Sequence and Power Transfer Test, to the 

TMI–1 Updated Final Safety Analysis Report 

(UFSAR) Section 8.3, Tests and Inspections. 

The proposed changes relocate 

surveillance requirement details that are not 

required by 10 CFR 50.36, and are [partially] 

consistent with standard technical 

specifications, NUREG–1430, ‘‘Standard 

Technical Specifications Babcock and Wilcox 

Plants.’’ The proposed changes do not change 

current surveillance requirements. The 

subject list of pumps, fans and valves that 

will be relocated to the UFSAR Section 8.3 

will continue to be administratively 

controlled and future changes will be 

controlled under 10 CFR 50.59. 

The probability of an accident is not 

increased by these proposed changes because 

the Sequence and Power Transfer Test is not 

an initiator of any design basis event. 

Additionally, the proposed changes do not 

involve any physical changes to plant 

structures, systems, or components (SSCs), or 

the manner in which these SSCs are 

operated, maintained, or controlled. The 

consequences of an accident will not be 

increased because the proposed 

administrative changes to the Sequence and 

Power Transfer Test and Sequence Test will 

continue to provide a high degree of 

assurance that the Electric Power System will 

meet its safety related function. 

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 previously evaluated? 

Response: No. 

The proposed changes do not alter the 

physical design, safety limits, safety analyses 

assumptions, or the manner in which the 

plant is operated or tested. The proposed 

changes are administrative in nature and the 

surveillance requirements remain the same. 

Accordingly, the proposed changes do not 

introduce any new accident initiators, nor do 

they reduce or adversely affect the 

capabilities of any plant SSC in the 

performance of their safety function. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in the margin of 

safety? 

Response: No. 

The margin of safety is associated with the 

confidence in the ability of the fission 

product barriers (i.e., fuel cladding, reactor 

coolant pressure boundary, and containment 

structure) to limit the level of radiation to the 

public. There are no physical changes to 

SSCs or operating and testing procedures 

associated with the proposed amendment. 

The proposed changes do not impact the 

assumptions of any design basis accident, 

and do not alter assumptions relative to the 

mitigation of an accident or transient event. 

The proposed changes are administrative in 

nature and the surveillance requirements 

remain the same. 

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, with the NRC edit noted above 

incorporated, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: J. Bradley 

Fewell, Esquire, Associate General 

Counsel, Exelon Generation Company, 

LLC, 4300 Winfield Road, Warrenville, 

IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 


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Exelon Generation Company, LLC, 

Docket No. 50–289, Three Mile Island 

Nuclear Station, Unit 1, Dauphin 

County, Pennsylvania 

Date of amendment request: 

September 24, 2010. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 3.4.1.2.3, 

to allow up to two Main Steam Safety 

Valves (MSSVs) per steam generator to 

be inoperable with no required 

reduction in power level. It would also 

revise the required maximum 

overpower trip setpoints for any 

additional inoperable MSSVs consistent 

with the plant transient analysis. The 

proposed change requires that with less 

than four MSSVs associated with either 

steam generator operable, the plant 

would be required to be brought to the 

hot shutdown condition. 

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, with NRC edits in brackets: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of any accident previously 

evaluated? 

Response: No. 

The proposed amendment is not a change 

to the plant structures, systems, or 

components. There is no increase to the 

likelihood of Main Steam Safety Valve 

(MSSV) related failures. The MSSVs are 

relied upon to mitigate the effects of Updated 

Final Safety Analysis Report (UFSAR) 

Chapter 14 design basis events including the 

loss of load (turbine trip), which is the 

limiting event for secondary system 

overpressure. Analyses, performed in 

accordance with NRC approved methods, 

have demonstrated that with reduced MSSV 

availability and following the specified 

power level restrictions, the MSSVs will 

continue to limit the secondary system 

pressure to less than 110 percent of the 

design pressure of the Once Through Steam 

Generators (OTSGs) and the Main Steam 

(MS) System as required by [the American 

Society of Mechanical Engineers] ASME 

code. 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 amendment is not a change 

to the plant structures, systems, or 

components (SSCs). Furthermore, within the 

current licensing basis, the MSSVs are 

accident mitigation SSCs. The current 

licensing basis does not [explicitly] include 

consideration of a MSSV failure as an event 

initiator [and a failed open MSSV has been 

shown to be bounded by the larger maximum 

break size analysis presented in the TMI–1 

UFSAR]. The proposed amendment will not 

fundamentally alter or create any new 

operator actions. 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 the margin of 

safety? 

Response: No. 

The limiting event for secondary system 

overpressure is a loss of load event (turbine 

trip). The event has been analyzed for 

varying MSSVs out of service, using NRC 

approved methods. The results of the 

analysis demonstrate that the existing design 

acceptance criteria (i.e., MS and OTSG 

pressure remain less than 110 percent of the 

design pressure) are met for all combinations 

of inoperable MSSVs and initial power levels 

described in the proposed change. Therefore, 

the proposed change does not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis, and based on this 

review, including the edits listed above, 

it appears that the three standards of 10 

CFR 50.92(c) are satisfied. Therefore, the 

NRC staff proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: J. Bradley 

Fewell, Esquire, Associate General 

Counsel, Exelon Generation Company, 

LLC, 4300 Winfield Road, Warrenville, 

IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

FPL Energy, Point Beach, LLC, Docket 

Nos. 50–266 and 50–301, Point Beach 

Nuclear Plant, Units 1 and 2, Town of 

Two Creeks, Manitowoc County, 

Wisconsin 

Date of amendment request: January 

27, 2010. 

Description of amendment request: 

The proposed changes would amend 

Renewed Facility Operating Licenses 

DPR–24 and DPR–27 for the Point Beach 

Nuclear Plant, Units 1 and 2, 

respectively. The proposed amendment 

consists of changes to Technical 

Specification 3.8.3, ‘‘Diesel Fuel Oil and 

Starting Air.’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration which is presented below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This proposed change increases the 

minimum required amount of stored diesel 

fuel in the associated common fuel oil 

storage tank [FOST] for two standby 

emergency power sources to start, load to 

their respective loading limits and to operate 

continuously up to a maximum of 48 hours. 

An increase in the minimum required fuel oil 

volume required in the fuel oil storage tanks 

does not increase the probability or 

consequences of an accident previously 

evaluated. 

[Limiting Condition for Operation] LCO 

3.8.3 Condition A, currently requires that one 

or more standby emergency power sources 

have ≥ 11,000 gallons of fuel when the 

associated [emergency diesel generator] EDG 

is declared operable. The proposed change 

increases the amount of stored fuel to 

≥ 24,000 gallons for two standby EDGs. It 

further adds new Required Action A.2 if the 

FOST stored capacity falls below the 

minimum required values. The proposed 

change also accounts for instrument indicator 

loop uncertainty values for unusable volume. 

New LCO [3.8.3] Condition B, addresses 

the case of one EDG operating in either Train 

‘‘A’’ or Train ‘‘B.’’ The new condition specifies 

that the minimum volume of diesel fuel 

required to support continued operation of a 

single EDG for 48 hours at rated load is 

≥ 13,000 gallons. This proposed change also 

accounts for instrument indicator loop 

uncertainty values for unusable volume. 

[Surveillance Requirement] SR 3.8.3.1 is 

revised to reflect the increased amount of 

diesel fuel required to be maintained to 

support operation of the EDGs following 

recalculation of required values. 

Following implementation of this proposed 

change, there will be no change in the ability 

of the EDGs to supply maximum postaccident load demands for 48 hours. The 

proposed minimum volume of fuel, ≥ 24,000 

gallons for two EDGs and ≥ 13,000 gallons for 

one EDG per train, ensures that a 48-hour 

supply of fuel is available when the 

associated standby emergency power source 

is required to be operable. 

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 EDGs and the associated support 

systems, such as the fuel oil storage and 

transfer systems, are designed to mitigate 

accidents and are not accident initiators. 

Following this change, the EDGs will 

continue to supply the required maximum 

post-accident load demand. The current 48- 

hour fuel supply requirements will be 

maintained following this change. The new 

required fuel oil volumes are within the 

capacities of the fuel oil storage tanks. 

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. 

There are two underground fuel oil storage 

tanks on site. Each tank has a capacity of 

approximately 35,000 gallons and each 

common fuel tank supports one EDG train. 


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Fuel can be manually transferred from one 

tank to another via a cross-connect valve. 

Sufficient fuel is maintained between the two 

tanks to allow one EDG to operate 

continuously at the required load for seven 

(7) days. At the proposed minimum required 

level, which is ≥ 24,000 gallons in the 

common fuel oil storage tanks for two 

standby emergency power sources, one tank 

could provide enough fuel for two EDGs in 

either Train A or Train B to continue 

operation for great than 48 hours. At the 

proposed minimum required level, which is 

≥ 13,000 gallons in each fuel oil storage tanks, 

one tank could provide enough fuel for one 

EDG in Train A and Train B to continue 

operation for greater than 48 hours. 

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: William Blair, 

Senior Attorney, NextEra Energy Point 

Beach, LLC, P.O. Box 14000, Juno 

Beach, FL 33408–0420. 

NRC Branch Chief: Robert J. 

Pascarelli. 

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: August 

27, 2010. 

Description of amendment request: 

The proposed amendment would add a 

new Action to Technical Specification 

(TS) 3.7.3, ‘‘Control Room Emergency 

Ventilation (CREV) System,’’ to permit 

one or more CREV subsystems to be 

inoperable for up to 90 days when the 

inoperability is due to inoperable CREV 

System High Efficiency Particulate Air 

(HEPA) filter and/or charcoal absorbers. 

The proposed TS changes also include 

an administrative change to correct 

errors in Unit 2 TS page header 

information that occurred during 

issuance of TS pages for a previous 

amendment. 

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. 

UFSAR [Updated Finale Safety Analysis 

Report] Chapter 14, ‘‘Plant Safety Analysis,’’ 

evaluates operational transients and 

accidents that result in radiological releases 

that affect control room occupants. UFSAR 

section 14.6, ‘‘Analysis of Design Basis 

Accidents—Uprated,’’ evaluates accidents 

that release fission products to the 

environment. The CREV System is not an 

accident initiator for any of the accidents 

described. The CREV System processes 

outside air needed to provide ventilation and 

pressurization for control room habitability 

to limit the control room dose during 

accidents evaluated in the UFSAR. Without 

crediting the performance of the HEPA filter 

or charcoal adsorbers, the analyses results 

concludes that the 30[-]day integrated postaccident doses in the control room are within 

the limits of 5 rem TEDE [total effective dose 

equivalent], as specified in 10 CFR 50.67 and 

GDC [General Design Criterion]-19. The 

control room dose increase is less than 10 

percent; leaving more than 60 percent 

remaining margin to the regulatory limit. 

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 CREV System is a ventilation system 

that filters outside air used to pressurize the 

control rooms to provide a protected 

environment from which operators can 

control the unit during airborne challenges 

from radioactivity during accident 

conditions. The CREV System does not 

initiate accidents. The proposed amendment 

allows the CREV HEPA filters and charcoal 

adsorbers to be repaired or replaced without 

shutting down the operating unit(s). No new 

modes of operation are introduced. 

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. 

Analyses associated with the prior 

approval of Alternate Source Term 

methodology for design basis accident dose 

consequences previously did not credit the 

CREV System charcoal adsorbers. Recent 

analyses have been performed to assess the 

post-accident 30-day control room dose 

removing credit for the CREV System HEPA 

filter. The results indicate a minimal increase 

in dose consequences (9.5 percent increase) 

due to removing credit for the CREV System 

HEPA filter. Even with no credit for either 

the CREV System HEPA filter or CREV 

System charcoal filter, the resultant control 

room dose maintains more than 60 percent 

margin to the regulatory limit of 5 rem TEDE. 

As such there is no reduction in a margin of 

safety for any duration of inoperability of the 

CREV System HEPA filter or charcoal 

adsorbers. While the HEPA filter and 

charcoal adsorbers are not credited for 

accident mitigation, they remain required by 

the BFN TS for compliance with the LCO 

3.7.3, ‘‘Control Room Emergency Ventilation 

(CREV) System,’’ further minimizing any 

potential reduction in a margin of safety. 

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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, 6A West 

Tower, Knoxville, Tennessee 37902. 

NRC Branch Chief: Douglas A. 

Broaddus. 

Virginia Electric and Power Company, 

Docket Nos. 50–338 and 50–339, North 

Anna Power Station, Units No. 1 and 

No. 2, Louisa County, Virginia 

Date of amendment request: October 

21, 2010. 

Description of amendment request: 

The proposed action involves the 

inclusion of the Westinghouse BestEstimate (BE) Large Break Loss-ofCoolant Accident (LBLOCA) analysis 

methodology using the Automated 

Statistical Treatment of Uncertainty 

Method (ASTRUM) for the analysis of 

LBLOCA to the list of methodologies 

approved for reference in the Core 

Operating Limits Report (COLR) in 

Technical Specification (TS) 5.6.5.b. 

This action also removes four obsolete 

COLR references that supported North 

Anna Improved Fuel (NAIF) product, 

Westinghouse Vantage 5, since this 

product is not planned to be used in 

future North Anna cores. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

[Criterion 1] 

Does the change involve a significant 

increase in the probability or consequences 

of an accident previously evaluated? 

Response: No. 

No physical plant changes are being made 

as a result of using the Westinghouse Best 

Estimate Large Break LOCA (BE–LBLOCA) 

analysis methodology. The proposed TS 

change simply involves updating the 

references in TS 5.6.5.b, Core Operating 

Limits Report (COLR), to reference the 

Westinghouse BE–LBLOCA analysis 

methodology, which is an NRC approved 

methodology, and to delete unnecessary 

references. Therefore, the probability of 

LOCA occurrence is not affected by the 

change. Further, the consequences of a LOCA 

are not increased, since the BE–LBLOCA 

analysis has demonstrated that the 

performance of the Emergency Core Cooling 


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System (ECCS) continues to conform to the 

criteria contained in 10 CFR 50.46, 

‘‘Acceptance Criteria for Emergency Core 

Cooling Systems for Light-Water Nuclear 

Power Reactors.’’ No other accident 

consequence is potentially affected by this 

change. 

Systems will continue to be operated in 

accordance with current design requirements 

under the new analysis, therefore no new 

components or system interactions have been 

identified that could lead to an increase in 

the probability of any accident previously 

evaluated in the Updated Final Safety 

Analysis Report (UFSAR). No changes were 

required to the Reactor Protection System 

(RPS) or Engineering Safety Features (ESF) 

setpoints because of the new analysis 

methodology. 

An analysis of the LBLOCA accident for 

North Anna Units 1 and 2 has been 

performed with the Westinghouse BE– 

LBLOCA analysis methodology using 

ASTRUM. The analysis was performed in 

compliance with the NRC conditions and 

limitations as identified in WCAP–1 6009– 

P–A. Based on the analysis results, it is 

concluded that the North Anna Units 1 and 

2 continue to satisfy the limits prescribed by 

10 CFR 50.46. 

There are no changes to assumptions of the 

radiological dose calculations. Hence, there 

is no increase in the predicted radiological 

consequences of accidents postulated in the 

UFSAR. 

Therefore, neither the probability of 

occurrence nor the consequences of an 

accident previously evaluated is significantly 

increased. 

[Criterion 2] 

Does the change create the possibility of a 

new or different kind of accident from any 

accident previously evaluated? 

Response: No. 

The use of the Westinghouse BE–LBLOCA 

analysis methodology with ASTRUM does 

not impact any of the applicable design 

criteria and pertinent licensing basis criteria 

continue to be met. Demonstrated adherence 

to the criteria in 10 CFR 50.46 precludes new 

challenges to components and systems that 

could introduce a new type of accident. 

Safety analysis evaluations have 

demonstrated that the use of Westinghouse 

BE–LBLOCA analysis methodology with 

ASTRUM is acceptable. Design and 

performance criteria continue to be met and 

no new single failure mechanisms have been 

created. The use of the Westinghouse BE– 

LBLOCA analysis methodology with 

ASTRUM does not involve any alteration to 

plant equipment or procedures that would 

introduce any new or unique operational 

modes or accident precursors. Furthermore, 

no changes have been made to any RPS or 

ESF actuation setpoints. Based on this 

review, it is concluded that no new accident 

scenarios, failure mechanisms, or limiting 

single failures are introduced as a result of 

the proposed changes. 

Therefore, the possibility for a new or 

different kind of accident from any accident 

previously evaluated is not created. 

[Criterion 3] 

Does this change involve a significant 

reduction in a margin of safety? 

Response: No. 

It has been demonstrated that the 

analytical technique used in the 

Westinghouse BE–LBLOCA analysis 

methodology using ASTRUM realistically 

describes the expected behavior of the reactor 

system during a postulated LOCA. 

Uncertainties have been accounted for as 

required by 10 CFR 50.46. A sufficient 

number of LOCAs with different break sizes, 

different locations, and other variations in 

properties have been considered to provide 

assurance that the most severe postulated 

LOCAs have been evaluated. The analysis 

has demonstrated that the acceptance criteria 

contained in 10 CFR 50.46 continue to be 

satisfied. 

Therefore, it is concluded that this change 

does not involve a significant reduction in 

the margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: Lillian M. 

Cuoco, Senior Counsel, Dominion 

Resources Services, Inc., 120 Tredegar 

Street, RS–2, Richmond, VA 23219. 

NRC Branch Chief: Gloria Kulesa. 

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 

20854. 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: 

April 29, 2010. 

Brief description of amendment: The 

amendments adopted Nuclear 

Regulatory Commission (NRC)-approved 

TS Task Force (TSTF) Standard 

Technical Specification change traveler 

TSTF–491, Revision 2, ‘‘Removal of 

Main Steam and Main Feedwater Valve 

Isolation Times from Technical 

Specifications.’’ The isolation times will 

be located outside of the TSs in a 

document subject to control by the 

10 CFR 50.59 process. 

Date of issuance: November 5, 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—181; Unit 

2—181; Unit 3—181. 

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 the Federal 

Register: July 27, 2010 (75 FR 44024). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated November 5, 

2010. 

No significant hazards consideration 

comments received: No. 


jdjones on DSK8KYBLC1PROD with NOTICES
Federal Register / Vol. 75, No. 229 / Tuesday, November 30, 2010 / Notices 74099 

Entergy Operations, Inc., Docket No. 50– 

368, Arkansas Nuclear One, Unit No. 2, 

Pope County, Arkansas 

Date of application for amendment: 

June 23, 2010. 

Brief description of amendment: 

Current Technical Specification (TS) 

6.5.8, ‘‘Inservice Testing Program,’’ 

contains references to the American 

Society of Mechanical Engineers 

(ASME) Boiler and Pressure Vessel 

Code, Section XI as the source of 

requirements for the inservice testing 

(IST) of ASME Code Class 1, 2, and 3 

pumps and valves. The amendment 

deleted the references to Section XI of 

the Code and incorporated references to 

the ASME Code for Operation and 

Maintenance of Nuclear Power Plants 

(ASME OM Code). The amendment also 

indicates that there may be some 

nonstandard frequencies utilized in the 

IST Program in which the provisions of 

Surveillance Requirement (SR) 3.0.2 are 

applicable. The changes are consistent 

with Technical Specification Task Force 

(TSTF) Technical Change Travelers 

TSTF–479–A, ‘‘Changes to Reflect 

Revision of 10 CFR 50.55a,’’ and TSTF– 

497–A, ‘‘Limit Inservice Testing Program 

SR 3.0.2 Application to Frequencies of 

2 Years or Less.’’ 

Date of issuance: November 5, 2010. 

Effective Date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment No.: 291. 

Renewed Facility Operating License 

No. NPF–6: Amendment revised the 

Technical Specifications/license. 

Date of initial notice in the Federal 

Register: August 10, 2010 (75 FR 

48375). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated November 5, 

2010. 

No significant hazards consideration 

comments received: No. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Nuclear Power Plant, 

Unit Nos. 1 and 2, Somervell County, 

Texas 

Date of amendment request: May 27, 

2010, as supplemented by letter dated 

August 26, 2010. 

Brief description of amendments: The 

amendments revised Technical 

Specification (TS) 3.8.3, ‘‘Diesel Fuel 

Oil, Lube Oil, and Starting Air,’’ by 

relocating the current stored diesel fuel 

oil and lube oil numerical volume and 

level requirements from the TSs to the 

TS Bases so that it may be modified 

under licensee control. The TSs have 

been modified so that the stored diesel 

fuel oil and lube oil inventory will 

require that a 7-day supply be available 

for each diesel generator. Condition A 

and Condition B in the Action table and 

Surveillance Requirements (SRs) 3.8.3.1 

and 3.8.3.2 are also revised to reflect the 

above change. The changes are 

consistent with NRC-approved Revision 

1 to Technical Specification Task Force 

(TSTF) Improved Standard Technical 

Specification Change Traveler TSTF– 

501, ‘‘Relocate Stored Fuel Oil and Lube 

Oil Volume Values to Licensee Control.’’ 

The availability of the TS improvement 

was announced in the Federal Register 

on May 26, 2010, as part of the 

consolidated line item improvement 

process. 

Date of issuance: November 4, 2010. 

Effective Date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: Unit 1—153; 

Unit 2—153. 

Facility Operating License Nos. NPF– 

87 and NPF–89: The amendments 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in the Federal 

Register: August 10, 2010 (75 FR 

48376). The supplemental letter dated 

August 26, 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 4, 

2010. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Georgia Power Company, 

Oglethorpe Power Corporation, 

Municipal Electric Authority of Georgia, 

City of Dalton, Georgia, Docket Nos. 50– 

321 and 50–366, Edwin I. Hatch Nuclear 

Plant, Units 1 and 2, Appling County, 

Georgia 

Date of application for amendments: 

December 17, 2009. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications (TSs) for Limiting 

Condition for Operations 3.1.2 

‘‘Reactivity Anomalies’’ changing 

Surveillance Requirement 3.1.2.1 

methodology. 

Date of issuance: November 4, 2010. 

Effective Date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment Nos.: 263 and 207. 

Renewed Facility Operating License 

Nos. DPR–57 and NPF–5: Amendments 

revised the licenses and the TSs. 

Date of initial notice in Federal 

Register: February 23, 2010. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated November 4, 

2010. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 19th day 

of November 2010. 

For the Nuclear Regulatory Commission. 

Joseph G. Giitter, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2010–29941 Filed 11–29–10; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket No. 70–143; License No. SNM–124; 

EA–10–076; NRC–2010–0365] 

In the Matter of Nuclear Fuel Services, 

Inc., Erwin, TN; Confirmatory Order 

Modifying License (Effective 

Immediately) 

I 

Nuclear Fuel Services, Inc. (NFS or 

Licensee) is the holder of Special 

Nuclear Materials License No. SNM–124 

issued by the Nuclear Regulatory 

Commission (NRC or Commission) 

pursuant to Title 10 of the Code of 

Federal Regulations (10 CFR) part 70 on 

July 2, 1999. The license authorizes the 

operation of the NFS facility in 

accordance with the conditions 

specified therein. The facility is located 

on the Licensee’s site in Erwin, 

Tennessee. 

This Confirmatory Order is the result 

of an agreement reached during an 

Alternative Dispute Resolution (ADR) 

session conducted on October 4, 2010. 

II 

On October 7, 2009, the NRC Office of 

Investigations (OI) began an 

investigation (OI Case No. 2–2010–001) 

at NFS. Based on the evidence 

developed during its investigation, OI 

substantiated that a former Industrial 

Safety Specialist at NFS willfully 

provided the NRC incomplete and 

inaccurate information concerning fire 

damper inspections at NFS on two 

separate occasions. The results of the 

investigation, completed on February 

26, 2010, were sent to NFS in a letter 


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