5614 Federal Register / Vol. 76, No. 21 / Tuesday, February 1, 2011 / Notices 

information collection, unless it is 

currently approved by the OMB under 

the PRA and displays a currently valid 

OMB Control Number. In addition, 

notwithstanding any other provisions of 

law, no person shall generally be subject 

to penalty for failing to comply with a 

collection of information if the 

collection of information does not 

display a currently valid OMB control 

number. See 5 CFR 1320.5(a) and 

1320.6. The DOL obtains OMB approval 

for this information collection under 

OMB Control Number 1218–0208. The 

current OMB approval is scheduled to 

expire on January 31, 2011; however, it 

should be noted that information 

collections submitted to the OMB 

receive a month-to-month extension 

while they undergo review. For 

additional information, see the related 

notice published in the Federal Register 

on November 18, 2010 (75 FR 70687). 

Interested parties are encouraged to 

send comments to the OMB, Office of 

Information and Regulatory Affairs at 

the address shown in the ADDRESSES

section within 30 days of publication of 

this notice in the Federal Register. In 

order to ensure appropriate 

consideration, comments should 

reference OMB Control Number 1218– 

0208. The OMB is particularly 

interested in comments that: 

• 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. 

Agency: Occupational Safety and 

Health Administration (OSHA). 

Title of Collection: Storage and 

Handling of Anhydrous Ammonia. 

OMB Control Number: 1218–0208. 

Affected Public: Private sector— 

businesses or other for-profits and 

farms. 

Total Estimated Number of 

Respondents: 2030. 

Total Estimated Number of 

Responses: 2030. 

Total Estimated Annual Burden 

Hours: 345. 

Total Estimated Annual Costs Burden: 

$0. 

Dated: January 27, 2010. 

Michel Smyth, 

Departmental Clearance Officer. 

[FR Doc. 2011–2155 Filed 1–31–11; 8:45 am] 

BILLING CODE 4510–26–P 

DEPARTMENT OF LABOR 

Public Availability of Department of 

Labor FY 2010 Service Contract 

Inventory 

AGENCY: Office of the Assistant 

Secretary for Administration and 

Management, Labor. 

ACTION: Notice of public availability of 

FY 2010 service contract inventories. 

SUMMARY: In accordance with Section 

743 of Division C of the Consolidated 

Appropriations Act of 2010 (Pub. L. 

111–117), the Department of Labor is 

publishing this notice to advise the 

public of the availability of the FY 2010 

Service Contract Inventory. This 

inventory provides information on 

service contract actions over $25,000 

made in FY 2010. The information is 

organized by function to show how 

contracted resources are distributed 

throughout the agency. The inventory 

has been developed in accordance with 

guidance issued on November 5, 2010, 

by the Office of Management and 

Budget’s Office of Federal Procurement 

Policy (OFPP). OFPP’s guidance is 

available at http://www.whitehouse.gov/ 

sites/default/files/omb/procurement/ 

memo/service-contract-inventoriesguidance-11052010.pdf. The 

Department of Labor has posted its 

inventory and a summary of the 

inventory on the agency’s Web site at 

the following link: http://www.dol.gov/ 

dol/aboutdol/main.htm. 

FOR FURTHER INFORMATION CONTACT: 

Questions regarding the service contract 

inventory should be directed to Brent 

Goe in the Office of Acquisition 

Management Services at (202) 693–7266 

or goe.brent2@dol.gov. 

Dated: January 27, 2011. 

Edward C. Hugler, 

Deputy Assistant Secretary for 

Administration and Management. 

[FR Doc. 2011–2211 Filed 1–27–11; 4:15 pm] 

BILLING CODE 4510–23–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2011–0021] 

Applications and Amendments to 

Facility Operating Licenses Involving 

Proposed No Significant Hazards 

Considerations and Containing 

Sensitive Unclassified Non-Safeguards 

Information and Order Imposing 

Procedures for Access to Sensitive 

Unclassified Non-Safeguards 

Information 

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, NRC, or 

NRC staff) is publishing this 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 notice includes notices of 

amendments containing sensitive 

unclassified non-safeguards information 

(SUNSI). 

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. 


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Federal Register / Vol. 76, No. 21 / Tuesday, February 1, 2011 / Notices 5615 

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 20852–2738. 

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 20852–2738, 

or at http://www.nrc.gov/reading-rm/ 

doc-collections/cfr/part002/part002- 

0309.html. 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.html. If a request for a 

hearing or petition for leave to intervene 

is filed within 60 days, 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 set forth 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, and the 

Commission has not made a final 

determination on the issue of no 

significant hazards consideration, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, then any hearing held 

would take place before the issuance of 

any amendment. 

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 

identification (ID) certificate, which 

allows the participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and (2) advise the 

Secretary that the participant will be 

submitting a request or petition for 

hearing (even in instances in which the 

participant, or its counsel or 

representative, already holds an NRCVerDate Mar<15>2010 15:05 Jan 31, 2011 Jkt 223001 PO 00000 Frm 00057 Fmt 4703 Sfmt 4703 E:\FR\FM\01FEN1.SGM 01FEN1

srobinson on DSKHWCL6B1PROD with NOTICES
5616 Federal Register / Vol. 76, No. 21 / Tuesday, February 1, 2011 / Notices 

issued 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 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 

e-mail 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 20852–2738, 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 firstclass 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:// 

ehd1.nrc.gov/EHD/, unless excluded 

pursuant to an order of the Commission, 

or the presiding officer. Participants are 

requested not to include personal 

privacy information, such as social 

security numbers, home addresses, or 

home phone numbers in their filings, 

unless an NRC regulation or other law 

requires submission of such 

information. 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 

amendment action, 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 20852– 

2738. Publicly available records will be 

accessible electronically from the 

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. 

Dominion Nuclear Connecticut Inc., et 

al., Docket Nos. 50–336 and 50–423, 

Millstone Power Station, Units 2 and 3, 

New London County, Connecticut 

Date of amendment request: July 12, 

2010, as supplemented by letter dated 

August 5, 2010. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). The licensee 

proposed an amendment to the Facility 

Operating Licenses for Millstone Power 

Station, Units 2 and 3 (MPS2 and MPS3, 

respectively). This amendment request 

pertains to the MPS2 and MPS3 Cyber 

Security Plans. In the same amendment 

request letter, sent under Dominion 

Resources Services, Inc. (DRC) 

letterhead, Kewaunee Power Station, 

Surry Power Station Units 1 and 2, and 

North Anna Power Station Units 1 and 

2, submitted amendment requests 

pertaining to their Cyber Security Plans. 

This notice only addresses the 

application as it pertains to MPS2 and 

MPS3. The licensee requested NRC 

approval of the MPS2 and MPS3 Cyber 

Security Plan, provided a proposed 

implementation schedule, and proposed 

to add a sentence to License Condition 


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Federal Register / Vol. 76, No. 21 / Tuesday, February 1, 2011 / Notices 5617 

2.C.4, ‘‘Physical Protection,’’ of MPS2, 

Facility Operating License (FOL) DPR– 

65 and to License Condition 2.E, of 

MPS3, FOL NPF–49, that would affirm 

when the licensee would fully 

implement and maintain in effect all 

provisions of the Cyber Security Plan. 

Basis for proposed no significant 

hazards consideration determination: 

As required by Title 10 of the Code of 

Federal Regulations (10 CFR) 50.91(a), 

the licensee has provided its analysis of 

the issue of no significant hazards 

consideration (NSHC). The NRC staff 

reviewed the licensee’s NSHC analysis 

against the standards of 10 CFR 50.92(c). 

The NRC staff’s review 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 Plan establishes the licensing basis for 

the Cyber Security Program for the sites. The 

Plan establishes how to achieve high 

assurance that specified nuclear power plant 

digital computer and communication 

systems, networks and functions are 

adequately protected against cyber attacks up 

to and including the design basis threat. 

Part one of the proposed change is 

designed to achieve high assurance that the 

systems are protected from cyber attacks. The 

Plan describes how plant modifications that 

involve digital computer systems are 

reviewed to provide high assurance of 

adequate protection against cyber attacks, up 

to and including the design basis threat. The 

proposed change does not alter accident 

analysis assumptions, add any initiators, or 

affect the function of plant systems or the 

manner in which systems are operated, 

maintained, modified, tested, or inspected. 

The first part of the proposed change is 

designed to achieve high assurance that the 

systems within the scope of the requirement 

are protected from cyber attacks and has no 

impact on the probability or consequences of 

an accident previously evaluated. The 

proposed change implements a Cyber 

Security Plan as a requirement not formally 

addressed previously. As such, the proposed 

Plan provides a significant enhancement to 

cyber security where no requirement existed 

before. 

The second part of the proposed change 

adds a sentence to the existing facility license 

conditions for Physical Protection. These 

changes are administrative and have no 

impact on the probability or consequences of 

an accident previously evaluated. 

Therefore, it is concluded that these 

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. 

This proposed amendment provides 

assurance that safety-related structures, 

systems and components (SSCs) are 

protected from cyber attacks. Implementation 

of 10 CFR 73.54 and the inclusion of a plan 

in the FOL do not result in the need of any 

new or different design-basis accident 

analysis. It does not introduce new 

equipment that could create a new or 

different kind of accident, and no new 

equipment failure modes are created. As a 

result, no new accident scenarios, failure 

mechanisms, or limiting single failures are 

introduced as a result of this proposed 

amendment. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

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. The proposed amendment would not 

alter the way any safety-related SSC 

functions and would not alter the way the 

plant is operated. The amendment provides 

assurance that safety-related SSCs are 

protected from cyber attacks. The proposed 

amendment would not introduce any new 

uncertainties or change any existing 

uncertainties associated with any safety 

limit. The proposed amendment would have 

no impact on the structural integrity of the 

fuel cladding, reactor coolant pressure 

boundary, or containment structure. Based 

on the above considerations, the proposed 

amendment would not degrade the 

confidence in the ability of the fission 

product barriers to limit the level of radiation 

to the public. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

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: Lillian M. 

Cuoco, Senior Counsel, Dominion 

Resources Services, Inc., 

120 Tredegar Street, RS–2, Richmond, 

VA 23219. 

NRC Branch Chief: Harold K. 

Chernoff. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–456 and 50–457, 

Braidwood Station, Units 1 and 2, Will 

County, Illinois Docket Nos. STN 50– 

454 and 50–455, Byron Station, Units 1 

and 2, Ogle County, Illinois 

Date of amendment request: 

December 14, 2010. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). The amendment 

would revise Technical Specification 

(TS) 5.5.9, ‘‘Steam Generator (SG) 

Program,’’ to exclude portions of the 

tubes within the tubesheet from 

periodic SG inspections and plugging or 

repair. In addition, this amendment 

request proposes to revise TS 5.6.9, 

‘‘Steam Generator (SG) Tube Inspection 

Report,’’ to remove reference to previous 

interim alternate repair criteria and 

provide reporting requirements specific 

to the temporary alternate criteria. 

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 previously analyzed accidents are 

initiated by the failure of plant structures, 

systems, or components. The proposed 

change that alters the steam generator (SG) 

inspection and reporting criteria does not 

have a detrimental impact on the integrity of 

any plant structure, system, or component 

that initiates an analyzed event. The 

proposed change will not alter the operation 

of, or otherwise increase the failure 

probability of any plant equipment that 

initiates an analyzed accident. 

Of the various accidents previously 

evaluated, the proposed changes only affect 

the steam generator tube rupture (SGTR), 

postulated steam line break (SLB), feedwater 

line break (FLB), locked rotor and control rod 

ejection accident evaluations. Loss-of-coolant 

accident (LOCA) conditions cause a 

compressive axial load to act on the tube. 

Therefore, since the LOCA tends to force the 

tube into the tubesheet rather than pull it out, 

it is not a factor in this amendment request. 

Another faulted load consideration is a safe 

shutdown earthquake (SSE); however, the 

seismic analysis of Model D5 SGs has shown 

that axial loading of the tubes is negligible 

during an SSE. 

During the SGTR event, the required 

structural integrity margins of the SG tubes 

and the tube-to-tubesheet joint over the H* 

distance will be maintained. Tube rupture in 

tubes with cracks within the tubesheet is 

precluded by the constraint provided by the 

presence of the tubesheet and the tube-totubesheet joint. Tube burst cannot occur 

within the thickness of the tubesheet. The 

tube-to-tubesheet joint constraint results from 

the hydraulic expansion process, thermal 

expansion mismatch between the tube and 

tubesheet, and from the differential pressure 

between the primary and secondary side, and 

tubesheet rotation. Based on this design, the 

structural margins against burst, as discussed 

in draft Regulatory Guide (RG) 1.121, ‘‘Bases 

for Plugging Degraded PWR Steam Generator 

Tubes,’’ and TS 5.5.9, are maintained for both 

normal and postulated accident conditions. 

The proposed change has no impact on the 

structural or leakage integrity of the portion 

of the tube outside of the tubesheet. The 

proposed change maintains structural and 


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5618 Federal Register / Vol. 76, No. 21 / Tuesday, February 1, 2011 / Notices 

leakage integrity of the SG tubes consistent 

with the performance criteria of TS 5.5.9. 

Therefore, the proposed change results in no 

significant increase in the probability of the 

occurrence of a SGTR accident. 

At normal operating pressures, leakage 

from tube degradation below the proposed 

limited inspection depth is limited by the 

tube-to-tubesheet crevice. Consequently, 

negligible normal operating leakage is 

expected from degradation below the 

inspected depth within the tubesheet region. 

The consequences of an SGTR event are not 

affected by the primary-to-secondary leakage 

flow during the event as primary-tosecondary leakage flow through a postulated 

tube that has been pulled out of the tubesheet 

is essentially equivalent to a severed tube. 

Therefore, the proposed change does not 

result in a significant increase in the 

consequences of a SGTR. 

Primary-to-secondary leakage from tube 

degradation in the tubesheet area during 

operating and accident conditions is 

restricted due to contact of the tube with the 

tubesheet. The leakage is modeled as flow 

through a porous medium through the use of 

the Darcy equation. The leakage model is 

used to develop a relationship between 

operational leakage and leakage at accident 

conditions that is based on differential 

pressure across the tubesheet and the 

viscosity of the fluid. A leak rate ratio was 

developed to relate the leakage at operating 

conditions to leakage at accident conditions. 

Since the fluid viscosity is based on fluid 

temperature and it is shown that for the most 

limiting accident, the fluid temperature does 

not exceed the normal operating temperature 

and therefore the viscosity ratio is assumed 

to be 1.0. Therefore, the leak rate ratio is a 

function of the ratio of the accident 

differential pressure and the normal 

operating differential pressure. 

The leakage factor of 1.93 for Braidwood 

Station Unit 2 and Byron Station Unit 2, for 

a postulated SLB/FLB, has been calculated as 

shown in Table 9–7 of WCAP–17072–P. 

However, EGC Braidwood Station Unit 2 and 

Byron Station Unit 2 will apply a factor of 

3.11 as determined by Westinghouse 

evaluation LTR–SGMP–09–100 P– 

Attachment, Revision 1, to the normal 

operating leakage associated with the 

tubesheet expansion region in the condition 

monitoring (CM) and operational assessment 

(OA). The leakage factor of 3.11 applies 

specifically to Byron Unit 2 and Braidwood 

Unit 2, both hot and cold legs, in Table 

RAI24–2 of LTRSGMP–09–100 P– 

Attachment, Revision 1. Through application 

of the limited tubesheet inspection scope, the 

existing operating leakage limit provides 

assurance that excessive leakage (i.e., greater 

than accident analysis assumptions) will not 

occur. The assumed accident induced leak 

rate limit is 0.5 gallons per minute at room 

temperature (gpmRT) for the faulted SG and 

0.218 gpmRT for the unfaulted SGs for 

accidents that assume a faulted SG. These 

accidents are the SLB and the locked rotor 

with a stuck open PORV. The assumed 

accident induced leak rate limit for accidents 

that do not assume a faulted SG is 1.0 gpmRT 

for all SGs. These accidents are the locked 

rotor and control rod ejection. 

No leakage factor will be applied to the 

locked rotor or control rod ejection transients 

due to their short duration, since the 

calculated leak rate ratio is less than 1.0. 

The TS 3.4.13 operational leak rate limit is 

150 gallons per day (gpd) (0.104 gpmRT) 

through any one SG. Consequently, there is 

sufficient margin between accident leakage 

and allowable operational leakage. The 

maximum accident leak rate ratio for the 

Model D5 design SGs is 1.93 as indicated in 

WCAP–1 7072–P, Table 9–7. However, EGC 

will use the more conservative value of 3.11 

accident leak rate ratio for the most limiting 

SG model design identified in Table RA124– 

2 of LTR–SGMP–09–100 P–Attachment 

Revision 1. This results in significant margin 

between the conservatively estimated 

accident leakage and the allowable accident 

leakage (0.5 gpmRT). 

For the CM assessment, the component of 

leakage from the prior cycle from below the 

H* distance will be multiplied by a factor of 

3.11 and added to the total leakage from any 

other source and compared to the allowable 

accident induced leakage limit. For the OA, 

the difference in the leakage between the 

allowable leakage and the accident induced 

leakage from sources other than the tubesheet 

expansion region will be divided by 3.11 and 

compared to the observed operational 

leakage. 

Based on the above, the performance 

criteria of NEI–97–06, Revision 2, and draft 

RG 1.121 continue to be met and the 

proposed change does not involve a 

significant increase in the probability or 

consequences of the applicable accidents 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not introduce 

any changes or mechanisms that create the 

possibility of a new or different kind of 

accident. Tube bundle integrity is expected 

to be maintained for all plant conditions 

upon implementation of the permanent 

alternate repair criteria. The proposed change 

does not introduce any new equipment or 

any change to existing equipment. No new 

effects on existing equipment are created nor 

are any new malfunctions introduced. 

Therefore, based on the above evaluation, 

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 defines the safety 

significant portion of the SG tube that must 

be inspected and repaired. WCAP–17072–P 

as modified by WCAP–1 7330–P identifies 

the specific inspection depth below which 

any type tube degradation has no impact on 

the performance criteria in NEI 97–06, 

Revision 2, ‘‘Steam Generator Program 

Guidelines.’’ 

The proposed change that alters the SG 

inspection and reporting criteria maintains 

the required structural margins of the SG 

tubes for both normal and accident 

conditions. NEI 97–06, and draft RG 1.121 

are used as the bases in the development of 

the limited tubesheet inspection depth 

methodology for determining that SG tube 

integrity considerations are maintained 

within acceptable limits. Draft RG 1.121 

describes a method acceptable to the NRC for 

meeting General Design Criteria (GDC) 14, 

‘‘Reactor Coolant Pressure Boundary,’’ GDC 

15, ‘‘Reactor Coolant System Design,’’ GDC 

31, ‘‘Fracture Prevention of Reactor Coolant 

Pressure Boundary,’’ and GDC 32, ‘‘Inspection 

of Reactor Coolant Pressure Boundary,’’ by 

reducing the probability and consequences of 

a SGTR. Draft RG 1.121 concludes that by 

determining the limiting safe conditions for 

tube wall degradation, the probability and 

consequences of a SGTR are reduced. This 

draft RG uses safety factors on loads for tube 

burst that are consistent with the 

requirements of Section III of the American 

Society of Mechanical Engineers (ASME) 

Code. 

For axially oriented cracking located 

within the tubesheet, tube burst is precluded 

due to the presence of the tubesheet. For 

circumferentially oriented cracking, WCAP– 

1 7072–P as modified by WCAP–17330–P 

defines a length of degradation-free expanded 

tubing that provides the necessary resistance 

to tube pullout due to the pressure induced 

forces, with applicable safety factors applied. 

Application of the limited hot and cold leg 

tubesheet inspection criteria will preclude 

unacceptable primary-to-secondary leakage 

during all plant conditions. The methodology 

for determining leakage as described in 

WCAP–17072–P as modified by LTRSGMP– 

09–100 P–Attachment shows that significant 

margin exists between an acceptable level of 

leakage during normal operating conditions 

that ensures meeting the SLB accidentinduced leakage assumption and the TS 

leakage limit of 150 gpd. 

Based on the above, it is concluded that the 

proposed changes do not result in any 

reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Nuclear, 4300 Winfield Road, 

Warrenville, IL 60555. 

NRC Branch Chief: Robert D. Carlson. 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station (CPS), Unit 1, DeWitt County, 

Illinois 

Date of amendment request: 

September 23, 2010. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). The proposed 

amendment would modify the CPS 

Technical Specifications (TS) Limiting 


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Condition for Operation (LCO) 3.7.6, 

‘‘Main Turbine Bypass System,’’ by 

allowing revision of the reactor 

operational limits, as specified in the 

CPS Core Operating Limits Report 

(COLR), to compensate for the 

inoperability of the Main Turbine 

Bypass System (MTBS). The revised TS 

will require that either the MTBS be 

OPERABLE or that the reactor power, 

Minimum Critical Power Ratio (MCPR), 

and Linear Heat Generation Rate (LHGR) 

limits for an inoperable MTBS be placed 

in effect as specified in the COLR. 

Additionally, the amendment proposes 

modifying TS 5.6.5, ‘‘Core Operating 

Limits Report (COLR),’’ to add a 

requirement to establish cycle 

dependent reactor thermal power limits 

for an inoperable MTBS. 

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 MTBS functions to limit reactor 

pressure and power increases during certain 

transients postulated in the accident analysis. 

The MTBS is a mitigation function and not 

the initiator of any evaluated accident or 

transient. Operation with an inoperable 

MTBS while in compliance with the imposed 

reactor power limitation, and MCPR and 

LHGR limits will offset the impact of losing 

the MTBS function. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change will not create any 

new modes of plant or equipment operation. 

The proposed change allows the option to 

apply a reactor power penalty and an 

additional penalty factor to the MCPR and 

LHGR when the MTSS is inoperable. The 

imposed reactor power limitation and the 

revised set of MCPR and LHGR limits will 

offset the impact of losing the MTBS 

function, and maintain the margin to the 

MCPR safety limit and the thermal 

mechanical design limits. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

By establishing more restrictive reactor 

power and MCPR and LHGR operating limits, 

there are no changes to the plant design and 

safety analysis. There are no changes to the 

reactor core design instrument setpoints. The 

margin of safety assumed in the safety 

analysis is not affected. Applicable regulatory 

requirements will continue to be met and 

adequate defense-in-depth will be 

maintained. Sufficient safety margins will be 

maintained. 

The analytical methods used to determine 

the reactor power limitation and the revised 

core operating limits were reviewed and 

approved by the NRC and are described in 

Technical Specification 5.6.5, ‘‘Core 

Operating Limits Report (COLR).’’ 

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. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Nuclear, 4300 Winfield Road, 

Warrenville, IL 60555. 

NRC Branch Chief: Robert D. Carlson. 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station (DNPS), 

Units 2 and 3, Grundy County, Illinois 

Date of amendment request: October 

4, 2010. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). The proposed 

amendment would revise Technical 

Specification (TS) Table 3.3.1.1 to 

eliminate Functions 5 and 10 from TS 

Table 3.3.1.1–1, delete footnote (c) from 

that table, and rename the footnote (d) 

to (c). These revisions would eliminate 

the requirement for a reactor scram, if 

vessel pressure is greater than or equal 

to 600 pounds per square inch gage 

(psig), with the reactor mode switch in 

startup and the main steam isolation 

valves closed or with a main turbine 

condenser vacuum low 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: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes to the DNPS Units 

2 and 3 TS revise the applicability of two 

protective functions and delete the associated 

TS Action statement. TS requirements that 

govern operability or routine testing of plant 

instruments are not assumed to be initiators 

of any analyzed event because these 

instruments are intended to prevent, detect, 

or mitigate accidents. Specifically, the reactor 

scram associated with the main steam 

isolation valve (MSIV) closure and low 

condenser vacuum (i.e., Functions 5 and 10 

of TS 3.3.1.1) is in anticipation of the loss of 

the normal heat sink and subsequent 

overpressurization transient. The scram at 

high pressure in startup conditions when 

MSIVs close and/or main condenser vacuum 

is low does not impact the limiting accident 

or transient analyses. An analysis by General 

Electric Hitachi Nuclear Energy (GEH) 

demonstrated that the Mode 2 scram function 

for MSIV closure and low condenser vacuum 

can be eliminated without affecting safe plant 

operation. Elimination of these required 

scrams will not involve an increase in the 

probability of an accident previously 

evaluated. 

Additionally, these proposed changes will 

not increase the consequences of an accident 

previously evaluated because the proposed 

changes do not adversely impact structures, 

systems, or components. These changes will 

not alter the operation of equipment assumed 

to be available for the mitigation of accidents 

or transients by the plant safety analysis. 

Function 5 is currently required in Mode 

2 with reactor pressure greater than or equal 

to 600 psig to ensure that the reactor is shut 

down, thus helping to prevent an 

overpressurization transient due to closure of 

main steam isolation valves. Similarly, 

Function 10 is currently required in Mode 2 

with reactor pressure greater than or equal to 

600 psig to help prevent an 

overpressurization transient by anticipating 

the turbine stop valve closure scram on loss 

of condenser vacuum. 

The existing scram logic is the result of 

experience gained during startup of an early 

vintage bailing water reactor in 1966 when 

operators had difficulty controlling reactor 

power above approximately 600 psig without 

pressure control. Experience on later plant 

startups indicates that the early experience 

may not be inherent to later boiling water 

reactor designs. As such, GEH subsequently 

recommended elimination of the Mode 2 

scram requirement. 

In Mode 2, the heat generation rate is low 

enough so that the other diverse Reactor 

Protection System (RPS) functions provide 

sufficient protection from an 

overpressurization transient. During normal 

power ascension in Mode 2 with the MSIVs 

open, reactor pressure vessel (RPV) pressure 

is controlled by the pressure regulator with 

increasing pressure setpoints. The maximum 

pressure regulator setpoint, which would 

translate to 1000 psig at rated power, would 

only allow a maximum dome pressure of 

approximately 900 psig in the Mode 2 power 

range. The potential scenario in Mode 2 

whereby the MSIVs would close 

unexpectedly and cause the pressure to 

increase would lead to the Average Power 

Rate Monitors, Neutron Flux-High, Setdown 

scram (i.e., TS 3.3.1.1, Function 2.a), 

followed by the Reactor Vessel Steam Dome 

Pressure-High scram (i.e., TS 3.3.1.1, 

Function 3). 

The consequences of a previously analyzed 

event are dependent on the initial conditions 


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assumed in the analysis, the availability and 

successful functioning of equipment assumed 

to operate in response to the analyzed event, 

and the setpoints at which these actions are 

initiated. The consequences of a previously 

evaluated accident are not significantly 

increased by the proposed change. The 

proposed change does not affect the 

performance of any equipment credited to 

mitigate the radiological consequences of an 

accident. Furthermore, there will be no 

change in the types or significant increase in 

the amounts of any effluents released offsite. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes to the DNPS Units 

2 and 3 TS revise the applicability of two 

protective functions and delete the associated 

TS Action statement. The RPS functions are 

not an initiator of any accident. Rather, the 

RPS is designed to initiate a reactor scram 

when one or more monitored parameters 

exceed their specified limits to preserve the 

integrity of the fuel cladding and the reactor 

coolant pressure boundary and minimize the 

energy that must be absorbed following an 

accident. The proposed changes do not alter 

the applicability for RPS functions during 

plant conditions in which an 

overpressurization transient is assumed to 

occur. Specifically, no changes are being 

made to the required number of channels per 

trip system, surveillance requirements, or 

allowable values for these functions during 

Mode 1 operation. 

The proposed change does not affect the 

control parameters governing unit operation 

or the response of plant equipment to 

transient conditions. The proposed change 

does not change or introduce any new 

equipment, modes of system operation or 

failure mechanisms. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

Margins of safety are established in the 

design of components, the configuration of 

components to meet certain performance 

parameters, and in the establishment of 

setpoints to initiate alarms and actions. The 

proposed changes revise the applicability for 

Functions 5 and 10 of TS 3.3.1.1 and delete 

an associated TS Action Statement. The 

proposed changes do not alter the 

applicability for RPS functions during plant 

conditions in which an overpressurization 

transient is assumed to occur. 

In addition, the proposed changes do not 

affect the probability of failure or availability 

of the affected instrumentation. Furthermore, 

the proposed changes will reduce the 

probability of test-induced plant transients 

and equipment failures. 

The proposed changes to the applicability 

for Functions 5 and 10 of TS 3.3.1.1 have no 

impact on equipment design or fundamental 

operation. There are no changes being made 

to safety limits or safety system allowable 

values that would adversely affect plant 

safety. The performance of the systems 

important to safety is not significantly 

affected by the proposed changes. The 

proposed change does not affect safety 

analysis assumptions or initial conditions 

and therefore, the margin of safety in the 

original safety analyses is maintained. 

As documented above, 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. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Nuclear, 4300 Winfield Road, 

Warrenville, IL 60555. 

NRC Branch Chief: Robert. D. Carlson. 

Exelon Generation Company, LLC, 

Docket No. 50–353, Limerick 

Generating Station, Unit 2, Montgomery 

County, Pennsylvania 

Date of amendment request: 

December 15, 2010. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). The proposed 

changes revise the Technical 

Specification (TS) relating to the Safety 

Limit Minimum Critical Power Ratios 

(SLMCPRs). The changes result from a 

cycle-specific analysis performed to 

support the operation of Limerick 

Generating Station, Unit 2, in the 

upcoming Cycle 12. Specifically, the 

proposed TS changes will revise the 

SLMCPRs contained in TS 2.1 for two 

recirculation loop operation and single 

recirculation loop operation to reflect 

the changes in the cycle-specific 

analysis. The new SLMCPRs are 

calculated using Nuclear Regulatory 

Commission (NRC)-approved 

methodology described in NEDE 24011– 

P–A, ‘‘General Electric Standard 

Application for Reactor Fuel,’’ Revision 

17. 

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 derivation of the cycle specific Safety 

Limit Minimum Critical Power Ratios 

(SLMCPRs) for incorporation into the 

Technical Specifications (TS), and their use 

to determine cycle specific thermal limits, 

has been performed using the methodology 

discussed in NEDE–24011–P–A, ‘‘General 

Electric Standard Application for Reactor 

Fuel,’’ Revision 17. 

The basis of the SLMCPR calculation is to 

ensure that during normal operation and 

during abnormal operational transients, at 

least 99.9% of all fuel rods in the core do not 

experience transition boiling if the limit is 

not violated. The new SLMCPRs preserve the 

existing margin to transition boiling. 

The MCPR [minimum critical power ratio] 

safety limit is reevaluated for each reload 

using NRC-approved methodologies. The 

analyses for Limerick Generating Station 

(LGS), Unit 2, Cycle 12 have concluded that 

a two loop MCPR safety limit of ≥1.09, based 

on the application of Global Nuclear Fuel’s 

NRC-approved MCPR safety limit 

methodology, will ensure that this 

acceptance criterion is met. For single-loop 

operation, a MCPR safety limit of ≥1.12 also 

ensures that this acceptance criterion is met. 

The MCPR operating limits are presented and 

controlled in accordance with the LGS, Unit 

2 Core Operating Limits Report (COLR). 

The requested TS changes do not involve 

any plant modifications or operational 

changes that could affect system reliability or 

performance or that could affect the 

probability of operator error. The requested 

changes do not affect any postulated accident 

precursors, do not affect any accident 

mitigating systems, and do not introduce any 

new accident initiation mechanisms. 

Therefore, the proposed TS 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 SLMCPR is a TS numerical value, 

calculated to ensure that during normal 

operation and during abnormal operational 

transients, at least 99.9% of all fuel rods in 

the core do not experience transition boiling 

if the limit is not violated. The new 

SLMCPRs are calculated using NRCapproved methodology discussed in NEDE– 

24011–P–A, ‘‘General Electric Standard 

Application for Reactor Fuel,’’ Revision 17. 

The proposed changes do not involve any 

new modes of operation or any plant 

modifications. The proposed revised MCPR 

safety limits have been shown to be 

acceptable for Cycle 12 operation. The core 

operating limits will continue to be 

developed using NRC-approved methods. 

The proposed MCPR safety limits or methods 

for establishing the core operating limits do 

not result in the creation of any new 

precursors to an accident. 

Therefore, the proposed TS changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 


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Response: No. 

There is no significant reduction in the 

margin of safety previously approved by the 

NRC as a result of the proposed change to the 

SLMCPRs. The new SLMCPRs are calculated 

using methodology discussed in NEDE– 

24011–P–A, ‘‘General Electric Standard 

Application for Reactor Fuel,’’ Revision 17. 

The SLMCPRs ensure that during normal 

operation and during abnormal operational 

transients, at least 99.9% of all fuel rods in 

the core do not experience transition boiling 

if the limit is not violated, thereby preserving 

the fuel cladding integrity. 

Therefore, the proposed TS changes do not 

involve a significant reduction in the margin 

of safety previously approved by the NRC. 

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. Bradley 

Fewell, Esquire, Associate General 

Counsel, Exelon Generation Company, 

LLC, 4300 Winfield Road, Warrenville, 

IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–346, 

Davis-Besse Nuclear Power Station, 

Unit No. 1, Ottawa County, Ohio 

Date of amendment request: July 16, 

2010, as supplemented by letters dated 

September 28, and November 23, 2010. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). The proposed 

amendment to the Facility Operating 

License (FOL) includes: (1) The 

proposed Davis-Besse Nuclear Power 

Station, Unit No. 1 (DBNPS) Cyber 

Security Plan (the Plan), (2) an 

implementation schedule, and (3) revise 

the existing FOL Physical Protection 

license condition to require the 

FirstEnergy Nuclear Operating Company 

(FENOC, the licensee) to fully 

implement and maintain in effect all 

provisions of the Commission approved 

Cyber Security Plan as required by Title 

10 of the Code of Federal Regulations 

(10 CFR) 73.54. 

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: The proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

The proposed change is required by 10 

CFR 73.54 and includes three parts. The first 

part is the submittal of the Plan for NRC 

review and approval. The Plan provides a 

description of how the requirements of the 

rule will be implemented at the DBNPS. The 

Plan establishes the licensing basis for the 

FENOC cyber security program for the 

DBNPS. The Plan establishes how to achieve 

high assurance that nuclear power plant 

digital computer and communication systems 

and networks associated with the following 

are adequately protected against cyber attacks 

up to and including the design basis threat: 

1. Safety-related and important-to-safety 

functions, 

2. Security functions, 

3. Emergency preparedness functions 

including offsite communications, and 

4. Support systems and equipment which 

if compromised, would adversely impact 

safety, security, or emergency preparedness 

functions. 

Part one of the proposed change is 

designed to achieve high assurance that the 

systems are protected from cyber attacks. The 

Plan itself does not require any plant 

modifications. However, the Plan does 

describe how plant modifications which 

involve digital computer systems are 

reviewed to provide high assurance of 

adequate protection against cyber attacks, up 

to and including the design basis threat as 

defined in the rule. 

The proposed change does not alter the 

plant configuration, require new plant 

equipment to be installed, alter accident 

analysis assumptions, add any initiators, 

affect the function of plant systems, or affect 

the manner in which systems are operated. 

The first part of the proposed change is 

designed to achieve high assurance that the 

systems within the scope of the rule are 

protected from cyber attacks and has no 

impact on the probability or consequences of 

an accident previously evaluated. 

The second part of the proposed change is 

an implementation schedule. The third part 

adds a sentence to the existing FOL license 

condition 2.D for Physical Protection. Both of 

these changes are administrative and have no 

impact on the probability or consequences of 

an 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. 

Criterion 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 is required by 10 

CFR 73.54 and includes three parts. The first 

part is the submittal of the Plan for NRC 

review and approval. The Plan provides a 

description of how the requirements of the 

rule will be implemented at the DBNPS. The 

Plan establishes the licensing basis for the 

FENOC cyber security program for the 

DBNPS. The Plan establishes how to achieve 

high assurance that nuclear power plant 

digital computer and communication systems 

and networks associated with the following 

are adequately protected against cyber attacks 

up to and including the design basis threat: 

1. Safety-related and important-to-safety 

functions, 

2. Security functions, 

3. Emergency preparedness functions 

including offsite communications, and 

4. Support systems and equipment which 

if compromised, would adversely impact 

safety, security, or emergency preparedness 

functions. 

Part one of the proposed change is 

designed to achieve high assurance that the 

systems within the scope of the rule are 

protected from cyber attacks. The Plan itself 

does not require any plant modifications. 

However, the Plan does describe how plant 

modifications which involve digital 

computer systems are reviewed to provide 

high assurance of adequate protection against 

cyber attacks, up to and including the design 

basis threat defined in the rule. 

The proposed change does not alter the 

plant configuration, require new plant 

equipment to be installed, alter accident 

analysis assumptions, add any initiators, 

affect the function of plant systems, or affect 

the manner in which systems are operated. 

The first part of the proposed change is 

designed to achieve high assurance that the 

systems within the scope of the rule are 

protected from cyber attacks and does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

The second part of the proposed change is 

an implementation schedule. The third part 

adds a sentence to the existing FOL license 

condition 2.D for Physical Protection. Both of 

these changes are administrative and do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

Criterion 3: The proposed change does not 

involve a significant reduction in a margin of 

safety. 

The proposed change is required by 10 

CFR 73.54 and includes three parts. The first 

part is the submittal of the Plan for NRC 

review and approval. The Plan provides a 

description of how the requirements of the 

rule will be implemented at the DBNPS. The 

Plan establishes the licensing basis for the 

FENOC cyber security program for the 

DBNPS. The Plan establishes how to achieve 

high assurance that nuclear power plant 

digital computer and communication systems 

and networks associated with the following 

are adequately protected against cyber attacks 

up to and including the design basis threat: 

1. Safety-related and important-to-safety 

functions, 

2. Security functions, 

3. Emergency preparedness functions 

including offsite communications, and 

4. Support systems and equipment which 

if compromised, would adversely impact 

safety, security, or emergency preparedness 

functions. 

Part one of the proposed change is 

designed to achieve high assurance that the 

systems within the scope of the rule are 

protected from cyber attacks. Plant safety 

margins are established through Limiting 

Conditions for Operation, Limiting Safety 

System Settings and Safety limits specified in 


srobinson on DSKHWCL6B1PROD with NOTICES
5622 Federal Register / Vol. 76, No. 21 / Tuesday, February 1, 2011 / Notices 

the Technical Specifications, methods of 

evaluation that establish design basis or 

change Updated Final Safety Analysis. 

Because there is no change to these 

established safety margins, the proposed 

change does not involve a significant 

reduction in a margin of safety. 

The second part of the proposed change is 

an implementation schedule. The third part 

adds a sentence to the existing FOL license 

condition 2.D for Physical Protection. Both of 

these changes are administrative and do not 

involve a significant 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: David W. 

Jenkins, FirstEnergy Nuclear Operating 

Company, FirstEnergy Corporation, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: Robert. D. Carlson. 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Nuclear Power Plant, 

Units 1 and 2, Somervell County, Texas 

Date of amendment request: 

December 1, 2010. 

Brief description of amendments: This 

amendment request contains sensitive 

unclassified non-safeguards information 

(SUNSI). The proposed amendment 

would revise Technical Specification 

(TS) 5.5.9, ‘‘Unit 1 Model D76 and Unit 

2 Model D5 Steam Generator (SG) 

Program,’’ to exclude portions of the 

Unit 2 Model D5 steam generator (SG) 

tubes below the top of the SG tubesheet 

from periodic SG tube inspections 

during Comanche Peak Nuclear Power 

Plant (CPNPP), Unit 2 Refueling Outage 

12 and the subsequent operating cycle. 

In addition, the proposed amendment 

would revise TS 5.6.9, ‘‘Unit 1 Model 

D76 and Unit 2 Model D5 Steam 

Generator Tube Inspection Report,’’ to 

provide reporting requirements specific 

to CPNPP, Unit 2 for the temporary 

alternate repair criteria. 

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. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

Of the accidents previously evaluated, the 

limiting transients with consideration to the 

proposed change to the SG tube inspection 

and repair criteria are the steam generator 

tube rupture (SGTR) event, the steam line 

break (SLB), and the feed line break (FLB) 

postulated accidents. 

The required structural integrity margins of 

the SG tubes and the tube-to-tubesheet joint 

over the H* distance will be maintained. 

Tube rupture in tubes with cracks within the 

tubesheet is precluded by the constraint 

provided by the presence of the tubesheet 

and the tube-to-tubesheet joint. Tube burst 

cannot occur within the thickness of the 

tubesheet. The tube-to-tubesheet joint 

constraint results from the hydraulic 

expansion process, thermal expansion 

mismatch between the tube and tubesheet, 

differential pressure between the primary 

and secondary side, and tubesheet rotation. 

Based on this design, the structural margins 

against burst, as discussed in [NRC] 

Regulatory Guide (RG) 1.121, ‘‘Bases for 

Plugging Degraded PWR [Pressurized-Water 

Reactor] Steam Generator Tubes,’’ and TS 

5.5.9 are maintained for both normal and 

postulated accident conditions. 

The proposed change has no impact on the 

structural or leakage integrity of the portion 

of the tube outside of the tubesheet. The 

proposed change maintains structural and 

leakage integrity of the SG tubes consistent 

with the performance criteria in TS 5.5.9. 

Therefore, the proposed change results in no 

significant increase in the probability of the 

occurrence of a[n] SGTR accident. 

At normal operating pressures, leakage 

from tube degradation below the proposed 

limited inspection depth is limited by the 

tube-to-tubesheet crevice. Consequently, 

negligible normal operating leakage is 

expected from degradation below the 

inspected depth within the tubesheet region. 

The consequences of an SGTR event are not 

affected by the primary-to-secondary leakage 

flow during the event as primary-tosecondary leakage flow through a postulated 

tube that has been pulled out of the tubesheet 

is essentially equivalent to a severed tube. 

Therefore, the proposed change does not 

result in a significant increase in the 

consequences of a[n] SGTR. 

The probability of a[n] SLB is unaffected 

by the potential failure of a steam generator 

tube as the failure of tube is not an initiator 

for a[n] SLB event. 

The leakage factor of 3.16 for CPNPP Unit 

2, for a postulated SLB/FLB, has been 

calculated as described in Reference 8.29 

[Westinghouse Letter LTR–SGMP–09–100P– 

Attachment, Revision 1, dated September 7, 

2010] and is shown in Revised Table 9–7 of 

this same reference. Specifically, for the 

condition monitoring (CM) assessment, the 

component of leakage from the prior cycle 

from below the H* distance will be 

multiplied by a factor of 3.16 and added to 

the total leakage from any other source and 

compared to the allowable accident induced 

leakage limit. For the operational assessment 

(OA), the difference in the leakage between 

the allowable leakage and the accident 

induced leakage from sources other than the 

tubesheet expansion region will be divided 

by 3.16 and compared to the observed 

operational leakage. The accident-induced 

leak rate limit for CPNPP Unit 2 is 1.0 gpm 

[gallons per minute]. The TS operational leak 

rate limit through any one steam generator is 

150 gpd [gallons per day] (0.1 gpm). 

Consequently, there is significant margin 

between accident leakage and allowable 

operational leakage. The SLB/FLB overall 

leakage factor is 3.16 resulting in significant 

margin between the conservatively estimated 

accident induced leakage and the allowable 

accident leakage. 

No leakage factor was applied to the locked 

rotor or control rod ejection transients due to 

their short duration. 

The previously analyzed accidents are 

initiated by the failure of plant structures, 

systems, or components. The proposed 

change that alters the SG inspection and 

reporting criteria does not have a detrimental 

impact on the integrity of any plant structure, 

system, or component that initiates an 

analyzed event. The proposed change will 

not alter the operation of, or otherwise 

increase the failure probability of any plant 

equipment that initiates an analyzed 

accident. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

The proposed change that alters the steam 

generator inspection and reporting criteria 

does not introduce any new equipment, 

create new failure modes for existing 

equipment, or create any new limiting single 

failures. Plant operation will not be altered, 

and all safety functions will continue to 

perform as previously assumed in accident 

analyses. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in the margin of safety? 

Response: No. 

The proposed change that alters the steam 

generator inspection and reporting criteria 

maintains the required structural margins of 

the SG tubes for both normal and accident 

conditions. Nuclear Energy Institute 97–06, 

Rev. 2, ‘‘Steam Generator Program 

Guidelines,’’ and NRC Regulatory Guide (RG) 

1.121, ‘‘Bases for Plugging Degraded PWR 

Steam Generator Tubes,’’ are used as the 

bases in the development of the limited 

tubesheet inspection depth methodology for 

determining that SG tube integrity 

considerations are maintained within 

acceptable limits. RG 1.121 describes a 

method acceptable to the NRC for meeting 

General Design Criteria (GDC) 14, ‘‘Reactor 

Coolant Pressure Boundary,’’ GDC 15, 

‘‘Reactor Coolant System Design,’’ GDC 31, 

‘‘Fracture Prevention of Reactor Coolant 

Pressure Boundary,’’ and GDC 32, ‘‘Inspection 

of Reactor Coolant Pressure Boundary,’’ by 

reducing the probability and consequences of 

a[n] SGTR. RG 1.121 concludes that by 

determining the limiting safe conditions for 

tube wall degradation, the probability and 


srobinson on DSKHWCL6B1PROD with NOTICES
Federal Register / Vol. 76, No. 21 / Tuesday, February 1, 2011 / Notices 5623 

consequences of a[n] SGTR are reduced. RG 

1.121 uses safety factors on loads for tube 

burst that are consistent with the 

requirements of Section III of the American 

Society of Mechanical Engineers (ASME) 

[Boiler and Pressure Vessel] Code. 

For axially oriented cracking located 

within the tubesheet, tube burst is precluded 

due to the presence of the tubesheet. For 

circumferentially oriented cracking, the H* 

Analysis documented in Section 4.1 

[Attachment 1 to letter dated December 1, 

2010] defines a length of degradation-free 

expanded tubing that provides the necessary 

resistance to tube pullout due to the pressure 

induced forces, with applicable safety factors 

applied. Application of the limited hot and 

cold leg tubesheet inspection criteria will 

preclude unacceptable primary-to-secondary 

leakage during all plant conditions. The 

methodology for determining leakage 

provides for large margins between 

calculated and actual leakage values in the 

proposed limited tubesheet inspection depth 

criteria. 

Therefore, the proposed change does not 

involve a significant reduction in any 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: Timothy P. 

Matthews, Esq., Morgan, Lewis and 

Bockius, 1800 M Street, NW., 

Washington, DC 20036. 

NRC Branch Chief: Michael T. 

Markley. 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, Wolf 

Creek Generating Station, Coffey 

County, Kansas 

Date of amendment request: 

November 30, 2010. 

Description of amendment request: 

This amendment request contains 

sensitive unclassified non-safeguards 

information (SUNSI). The proposed 

amendment would revise the Wolf 

Creek Generating Station’s (WCGS’s) 

Technical Specification (TS) 5.5.9, 

‘‘Steam Generator (SG) Program,’’ to 

exclude portions of the tube below the 

top of the steam generator tubesheet 

from periodic steam generator tube 

inspections during Refueling Outage 18 

and the subsequent operating cycle. In 

addition, the proposed amendment 

would revise TS 5.6.10, ‘‘Steam 

Generator Tube Inspection Report,’’ to 

remove references to previous interim 

alternate repair criteria and provide 

reporting requirements specific to the 

temporary alternate repair criteria. 

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 previously analyzed accidents are 

initiated by the failure of plant structures, 

systems, or components. The proposed 

change that alters the steam generator 

inspection criteria does not have a 

detrimental impact on the integrity of any 

plant structure, system, or component that 

initiates an analyzed event. The proposed 

change will not alter the operation of, or 

otherwise increase the failure probability of 

any plant equipment that initiates an 

analyzed accident. 

Of the applicable accidents previously 

evaluated, the limiting transients with 

consideration to the proposed change to the 

steam generator tube inspection and repair 

criteria are the steam generator tube rupture 

(SGTR) event and the feedline break (FLB) 

postulated accidents. 

During the SGTR event, the required 

structural integrity margins of the steam 

generator tubes and the tube-to-tubesheet 

joint over the H* distance will be 

maintained. Tube rupture in tubes with 

cracks within the tubesheet is precluded by 

the presence of the tubesheet and constraint 

provided by the tube-to-tubesheet joint. Tube 

burst cannot occur within the thickness of 

the tubesheet. The tube-to-tubesheet joint 

constraint results from the hydraulic 

expansion process, thermal expansion 

mismatch between the tube and tubesheet, 

from the differential pressure between the 

primary and secondary side, and tubesheet 

deflection. Based on this design, the 

structural margins against burst, as discussed 

in Regulatory Guide (RG) 1.121, ‘‘Bases for 

Plugging Degraded PWR [Pressurized-Water 

Reactor] Steam Generator Tubes,’’ and TS 

5.5.9 are maintained for both normal and 

postulated accident conditions. 

The proposed change has no impact on the 

structural or leakage integrity of the portion 

of the tube outside of the tubesheet. The 

proposed change maintains structural and 

leakage integrity of the steam generator tubes 

consistent with the performance criteria in 

TS 5.5.9. Therefore, the proposed change 

results in no significant increase in the 

probability of the occurrence of a[n] SGTR 

accident. 

At normal operating pressures, leakage 

from tube degradation below the proposed 

limited inspection depth is limited by the 

tube-to-tubesheet joint. Consequently, 

negligible normal operating leakage is 

expected from degradation below the 

inspected depth within the tubesheet region. 

The consequences of an SGTR event are not 

affected by the primary to secondary leakage 

flow during the event as primary to 

secondary leakage flow through a postulated 

tube that has been pulled out of the tubesheet 

is essentially equivalent to a severed tube. 

Therefore, the proposed changes do not 

result in a significant increase in the 

consequences of a[n] SGTR. 

The consequences of a steam line break 

(SLB) are also not significantly affected by 

the proposed changes. During a[n] SLB 

accident, the reduction in pressure above the 

tubesheet on the shell side of the steam 

generator creates an axially uniformly 

distributed load on the tubesheet due to the 

reactor coolant system pressure on the 

underside of the tubesheet. The resulting 

bending action constrains the tubes in the 

tubesheet thereby restricting primary-tosecondary leakage below the midplane. 

Primary-to-secondary leakage from tube 

degradation in the tubesheet area during the 

limiting accident (i.e., an SLB) is limited by 

flow restrictions. These restrictions result 

from the crack and tube-to-tubesheet contact 

pressures that provide a restricted leakage 

path above the indications and also limit the 

degree of potential crack face opening as 

compared to free span indications. 

The leakage factor of 2.50 for WCGS, for a 

postulated SLB/FLB, has been calculated as 

shown in Revised Table 9–7 of Reference 15 

[Westinghouse Letter LTR–SGMP–09–100, 

dated August 12, 2009]. Specifically, for the 

condition monitoring (CM) assessment, the 

component of leakage from the prior cycle 

from below the H* distance will be 

multiplied by a factor of 2.50 and added to 

the total leakage from any other source and 

compared to the allowable accident induced 

leakage limit. For the operational assessment 

(OA), the difference in the leakage between 

the allowable leakage and the accident 

induced leakage from sources other than the 

tubesheet expansion region will be divided 

by 2.50 and compared to the observed 

operational leakage. 

The probability of an SLB is unaffected by 

the potential failure of a steam generator tube 

as the failure of the tube is not an initiator 

for an SLB event. SLB leakage is limited by 

leakage flow restrictions resulting from the 

leakage path above potential cracks through 

the tube-to-tubesheet crevice. The leak rate 

during postulated accident conditions 

(including locked rotor) has been shown to 

remain within the accident analysis 

assumptions for all axial and or 

circumferentially orientated cracks occurring 

15.2 inches below the top of the tubesheet. 

The accident induced leak rate limit for 

WCGS is 1.0 gpm [gallon per minute]. The TS 

3.4.13, ‘‘RCS [Reactor Coolant System] 

Operational LEAKAGE,’’ operational leak rate 

limit is 150 gpd [gallons per day] (0.1 gpm) 

through anyone steam generator. 

Consequently, accident leakage is 

approximately 10 times the allowable 

leakage, if only one steam generator is 

leaking. Using an SLB/FLB overall leakage 

factor of 2.50, accident induced leakage is 

approximately 0.5 gpm, if all 4 steam 

generators are leaking at 150 gpd at the 

beginning of the accident. Therefore, 

significant margin exists between the 

conservatively estimated accident induced 

leakage and the allowable accident leakage 

(1.0 gpm). 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 


srobinson on DSKHWCL6B1PROD with NOTICES
5624 Federal Register / Vol. 76, No. 21 / Tuesday, February 1, 2011 / Notices 

1While a request for hearing or petition to 

intervene in this proceeding must comply with the 

filing requirements of the NRC’s ‘‘E-Filing Rule,’’ the 

initial request to access SUNSI under these 

procedures should be submitted as described in this 

paragraph. 

2. Does the change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

Response: No. 

The proposed change alters the steam 

generator inspection and reporting criteria. It 

does not introduce any new equipment, 

create new failure modes for existing 

equipment, or create any new limiting single 

failures. Plant operation will not be altered, 

and safety functions will continue to perform 

as previously assumed in accident analyses. 

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 change involve a significant 

reduction in a margin of safety? 

Response: No. 

The proposed change alters the steam 

generator inspection and reporting criteria. It 

maintains the required structural margins of 

the steam generator tubes for both normal 

and accident conditions. NEI [Nuclear Energy 

Institute] 97–06, Revision 2, and RG 1.121, 

are used as the bases in the development of 

the limited tubesheet inspection depth 

methodology for determining that steam 

generator tube integrity considerations are 

maintained within acceptable limits. RG 

1.121 describes a method acceptable to the 

NRC for meeting GDC [General Design 

Criterion] 14, ‘‘Reactor Coolant Pressure 

Boundary,’’ GDC 15, ‘‘Reactor Coolant System 

Design,’’ GDC 31, ‘‘Fracture Prevention of 

Reactor Coolant Pressure Boundary,’’ and 

GDC 32, ‘‘Inspection of Reactor Coolant 

Pressure Boundary,’’ by reducing the 

probability and consequences of a[n] SGTR. 

RG 1.121 concludes that by determining the 

limiting safe conditions for tube wall 

degradation, the probability and 

consequences of a[n] SGTR are reduced. This 

RG uses safety factors on loads for tube burst 

that are consistent with the requirements of 

Section III of the American Society of 

Mechanical Engineers (ASME) [Boiler and 

Pressure Vessel] Code. For axially-oriented 

cracking located within the tubesheet, tube 

burst is precluded due to the presence of the 

tubesheet. For circumferentially-oriented 

cracking, the H* Analysis documented in 

Section 3 [of letter dated November 30, 

2010], defines a length of degradation-free 

expanded tubing that provides the necessary 

resistance to tube pullout due to the pressure 

induced forces, with applicable safety factors 

applied. Application of the limited hot and 

cold leg tubesheet inspection criteria will 

preclude unacceptable primary to secondary 

leakage during all plant conditions. The 

methodology for determining leakage 

provides for large margins between 

calculated and actual leakage values in the 

proposed limited tubesheet inspection depth 

criteria. 

Therefore, the proposed change does not 

involve a significant reduction in any 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. 

Order Imposing Procedures for Access 

to Sensitive Unclassified NonSafeguards Information for Contention 

Preparation 

Dominion Nuclear Connecticut Inc., et 

al., Docket Nos. 50–336 and 50–423, 

Millstone Power Station, Unit 2 and 

3, New London County, 

Connecticut 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station (CPS), Unit 1, DeWitt 

County, Illinois 

Exelon Generation Company, LLC, 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station 

(DNPS), Units 2 and 3, Grundy 

County, Illinois 

Exelon Generation Company, LLC, 

Docket No. 50–353, Limerick 

Generating Station, Unit 2, 

Montgomery County, Pennsylvania 

Exelon Generation Company, LLC 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–346, 

Davis-Besse Nuclear Power Station, 

Unit No. 1, Ottawa County, Ohio 

Luminant Generation Company LLC, 

Docket Nos. 50–445 and 50–446, 

Comanche Peak Nuclear Power 

Plant, Units 1 and 2, Somervell 

County, Texas 

Wolf Creek Nuclear Operating 

Corporation, Docket No. 50–482, 

Wolf Creek Generating Station, 

Coffey County, Kansas 

A. This Order contains instructions 

regarding how potential parties to this 

proceeding may request access to 

documents containing Sensitive 

Unclassified Non-Safeguards 

Information (SUNSI). 

B. Within 10 days after publication of 

this notice of hearing and opportunity to 

petition for leave to intervene, any 

potential party who believes access to 

SUNSI is necessary to respond to this 

notice may request such access. A 

‘‘potential party’’ is any person who 

intends to participate as a party by 

demonstrating standing and filing an 

admissible contention under 10 CFR 

2.309. Requests for access to SUNSI 

submitted later than 10 days after 

publication will not be considered 

absent a showing of good cause for the 

late filing, addressing why the request 

could not have been filed earlier. 

C. The requestor shall submit a letter 

requesting permission to access SUNSI 

to the Office of the Secretary, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, Attention: 

Rulemakings and Adjudications Staff, 

and provide a copy to the Associate 

General Counsel for Hearings, 

Enforcement and Administration, Office 

of the General Counsel, Washington, DC 

20555–0001. The expedited delivery or 

courier mail address for both offices is: 

U.S. Nuclear Regulatory Commission, 

11555 Rockville Pike, Rockville, 

Maryland 20852. The e-mail address for 

the Office of the Secretary and the 

Office of the General Counsel are 

Hearing.Docket@nrc.gov and 

OGCmailcenter@nrc.gov, respectively.1

The request must include the following 

information: 

(1) A description of the licensing 

action with a citation to this Federal 

Register notice; 

(2) The name and address of the 

potential party and a description of the 

potential party’s particularized interest 

that could be harmed by the action 

identified in C.(1); 

(3) The identity of the individual or 

entity requesting access to SUNSI and 

the requestor’s basis for the need for the 

information in order to meaningfully 

participate in this adjudicatory 

proceeding. In particular, the request 

must explain why publicly-available 

versions of the information requested 

would not be sufficient to provide the 

basis and specificity for a proffered 

contention; 

D. Based on an evaluation of the 

information submitted under paragraph 

C.(3) the NRC staff will determine 

within 10 days of receipt of the request 

whether: 

(1) There is a reasonable basis to 

believe the petitioner is likely to 

establish standing to participate in this 

NRC proceeding; and 

(2) The requestor has established a 

legitimate need for access to SUNSI. 

E. If the NRC staff determines that the 

requestor satisfies both D.(1) and D.(2) 

above, the NRC staff will notify the 

requestor in writing that access to 

SUNSI has been granted. The written 

notification will contain instructions on 

how the requestor may obtain copies of 

the requested documents, and any other 

conditions that may apply to access 

those documents. These conditions may 

include, but are not limited to, the 

signing of a Non-Disclosure Agreement 


srobinson on DSKHWCL6B1PROD with NOTICES
Federal Register / Vol. 76, No. 21 / Tuesday, February 1, 2011 / Notices 5625 

2Any motion for Protective Order or draft NonDisclosure Affidavit or Agreement for SUNSI must 

be filed with the presiding officer or the Chief 

Administrative Judge if the presiding officer has not 

yet been designated, within 30 days of the deadline 

for the receipt of the written access request. 

3Requestors should note that the filing 

requirements of the NRC’s E-Filing Rule (72 FR 

49139; August 28, 2007) apply to appeals of NRC 

staff determinations (because they must be served 

on a presiding officer or the Commission, as 

applicable), but not to the initial SUNSI request 

submitted to the NRC staff under these procedures. 

or Affidavit, or Protective Order 2 setting 

forth terms and conditions to prevent 

the unauthorized or inadvertent 

disclosure of SUNSI by each individual 

who will be granted access to SUNSI. 

F. Filing of Contentions. Any 

contentions in these proceedings that 

are based upon the information received 

as a result of the request made for 

SUNSI must be filed by the requestor no 

later than 25 days after the requestor is 

granted access to that information. 

However, if more than 25 days remain 

between the date the petitioner is 

granted access to the information and 

the deadline for filing all other 

contentions (as established in the notice 

of hearing or opportunity for hearing), 

the petitioner may file its SUNSI 

contentions by that later deadline. 

G. Review of Denials of Access. 

(1) If the request for access to SUNSI 

is denied by the NRC staff either after 

a determination on standing and need 

for access, or after a determination on 

trustworthiness and reliability, the NRC 

staff shall immediately notify the 

requestor in writing, briefly stating the 

reason or reasons for the denial. 

(2) The requestor may challenge the 

NRC staff’s adverse determination by 

filing a challenge within 5 days of 

receipt of that determination with: (a) 

the presiding officer designated in this 

proceeding; (b) if no presiding officer 

has been appointed, the Chief 

Administrative Judge, or if he or she is 

unavailable, another administrative 

judge, or an administrative law judge 

with jurisdiction pursuant to 10 CFR 

2.318(a); or (c) if another officer has 

been designated to rule on information 

access issues, with that officer. 

H. Review of Grants of Access. A 

party other than the requestor may 

challenge an NRC staff determination 

granting access to SUNSI whose release 

would harm that party’s interest 

independent of the proceeding. Such a 

challenge must be filed with the Chief 

Administrative Judge within 5 days of 

the notification by the NRC staff of its 

grant of access. 

If challenges to the NRC staff 

determinations are filed, these 

procedures give way to the normal 

process for litigating disputes 

concerning access to information. The 

availability of interlocutory review by 

the Commission of orders ruling on 

such NRC staff determinations (whether 

granting or denying access) is governed 

by 10 CFR 2.311.3

I. The Commission expects that the 

NRC staff and presiding officers (and 

any other reviewing officers) will 

consider and resolve requests for access 

to SUNSI, and motions for protective 

orders, in a timely fashion in order to 

minimize any unnecessary delays in 

identifying those petitioners who have 

standing and who have propounded 

contentions meeting the specificity and 

basis requirements in 10 CFR Part 2. 

Attachment 1 to this Order summarizes 

the general target schedule for 

processing and resolving requests under 

these procedures. 

It Is So Ordered. 

Dated at Rockville, Maryland, this 25th day 

of January 2011. 

For the Commission. 

Annette L. Vietti-Cook, 

Secretary of the Commission. 

ATTACHMENT 1—General Target 

Schedule for Processing and Resolving 

Requests for Access to Sensitive 

Unclassified Non-Safeguards 

Information in this Proceeding 

Day Event/Activity 

0 ......................... Publication of Federal Register notice of hearing and opportunity to petition for leave to intervene, including order with instructions for access requests. 

10 ....................... Deadline for submitting requests for access to Sensitive Unclassified Non-Safeguards Information (SUNSI) with information: 

Supporting the standing of a potential party identified by name and address; describing the need for the information in 

order for the potential party to participate meaningfully in an adjudicatory proceeding. 

60 ....................... Deadline for submitting petition for intervention containing: (i) Demonstration of standing; (ii) all contentions whose formulation does not require access to SUNSI (+25 Answers to petition for intervention; +7 requestor/petitioner reply). 

20 ....................... Nuclear Regulatory Commission (NRC) staff informs the requestor of the staff’s determination whether the request for access 

provides a reasonable basis to believe standing can be established and shows need for SUNSI. (NRC staff also informs 

any party to the proceeding whose interest independent of the proceeding would be harmed by the release of the information.) If NRC staff makes the finding of need for SUNSI and likelihood of standing, NRC staff begins document processing 

(preparation of redactions or review of redacted documents). 

25 ....................... If NRC staff finds no ‘‘need’’ or no likelihood of standing, the deadline for requestor/petitioner to file a motion seeking a ruling 

to reverse the NRC staff’s denial of access; NRC staff files copy of access determination with the presiding officer (or Chief 

Administrative Judge or other designated officer, as appropriate). If NRC staff finds ‘‘need’’ for SUNSI, the deadline for any 

party to the proceeding whose interest independent of the proceeding would be harmed by the release of the information 

to file a motion seeking a ruling to reverse the NRC staff’s grant of access. 

30 ....................... Deadline for NRC staff reply to motions to reverse NRC staff determination(s). 

40 ....................... (Receipt +30) If NRC staff finds standing and need for SUNSI, deadline for NRC staff to complete information processing and 

file motion for Protective Order and draft Non-Disclosure Affidavit. Deadline for applicant/licensee to file Non-Disclosure 

Agreement for SUNSI. 

A ........................ If access granted: Issuance of presiding officer or other designated officer decision on motion for protective order for access 

to sensitive information (including schedule for providing access and submission of contentions) or decision reversing a 

final adverse determination by the NRC staff. 

A + 3 .................. Deadline for filing executed Non-Disclosure Affidavits. Access provided to SUNSI consistent with decision issuing the protective order. 

A + 28 ................ Deadline for submission of contentions whose development depends upon access to SUNSI. However, if more than 25 days 

remain between the petitioner’s receipt of (or access to) the information and the deadline for filing all other contentions (as

established in the notice of hearing or opportunity for hearing), the petitioner may file its SUNSI contentions by that later 

deadline. 

A + 53 ................ (Contention receipt +25) Answers to contentions whose development depends upon access to SUNSI. 

A + 60 ................ (Answer receipt +7) Petitioner/Intervenor reply to answers. 


srobinson on DSKHWCL6B1PROD with NOTICES
5626 Federal Register / Vol. 76, No. 21 / Tuesday, February 1, 2011 / Notices 

Day Event/Activity 

>A + 60 .............. Decision on contention admission. 

[FR Doc. 2011–2027 Filed 1–26–11; 4:15 pm] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2011–0006] 

Sunshine Federal Register Notice 

AGENCY HOLDING THE MEETINGS: Nuclear 

Regulatory Commission. 

DATE: Weeks of January 31, February 7, 

14, 21, 28, March 7, 2011. 

PLACE: Commissioners’ Conference 

Room, 11555 Rockville Pike, Rockville, 

Maryland. 

STATUS: Public and Closed. 

Week of January 31, 2011 

Tuesday, February 1, 2011 

9 a.m. 

Briefing on Digital Instrumentation 

and Controls (Public Meeting). 

(Contact: Steven Arndt, 301–415– 

6502). 

This meeting will be Webcast live at 

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

Week of February 7, 2011—Tentative 

Tuesday, February 8, 2011 

9 a.m. 

Briefing on Implementation of Part 26 

(Public Meeting). (Contact: Shana 

Helton, 301–415–7198). 

This meeting will be Webcast live at 

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

Week of February 14, 2011—Tentative 

There are no meetings scheduled for 

the week of February 14, 2011. 

Week of February 21, 2011—Tentative 

Thursday, February 24, 2011 

9 a.m. 

Briefing on Groundwater Task Force 

(Public Meeting). (Contact: Margie 

Kotzalas, 301–415–1727). 

This meeting will be Webcast live at 

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

Week of February 28, 2011—Tentative 

Tuesday, March 1, 2011 

9 a.m. 

Briefing on Reactor Materials Aging 

Management Issues (Public 

Meeting). (Contact: Allen Hiser, 

301–415–5650). 

This meeting will be Webcast live at 

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

Week of March 7, 2011—Tentative 

There are no meetings scheduled for 

the week of March 7, 2011. 

* * * * * 

*The schedule for Commission 

meetings is subject to change on short 

notice. To verify the status of meetings, 

call (recording)—(301) 415–1292. 

Contact person for more information: 

Rochelle Bavol, (301) 415–1651. 

* * * * * 

The NRC Commission Meeting 

Schedule can be found on the Internet 

at: http://www.nrc.gov/about-nrc/policymaking/schedule.html. 

* * * * * 

The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings, or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g. 

braille, large print), please notify Angela 

Bolduc, Chief, Employee/Labor 

Relations and Work Life Branch, at 301– 

492–2230, TDD: 301–415–2100, or by email at angela.bolduc@nrc.gov. 

Determinations on requests for 

reasonable accommodation will be 

made on a case-by-case basis. 

* * * * * 

This notice is distributed 

electronically to subscribers. If you no 

longer wish to receive it, or would like 

to be added to the distribution, please 

contact the Office of the Secretary, 

Washington, DC 20555 (301–415–1969), 

or send an e-mail to 

darlene.wright@nrc.gov. 

Dated: January 27, 2011. 

Rochelle C. Bavol, 

Policy Coordinator, Office of the Secretary. 

[FR Doc. 2011–2258 Filed 1–28–11; 11:15 am] 

BILLING CODE 7590–01–P 

SECURITIES AND EXCHANGE 

COMMISSION 

Sunshine Act Meeting 

Notice is hereby given, pursuant to 

the provisions of the Government in the 

Sunshine Act, Public Law 94–409, that 

the Securities and Exchange 

Commission will hold a Closed Meeting 

on Thursday, February 3, 2011 at 10 

a.m. 

Commissioners, Counsel to the 

Commissioners, the Secretary to the 

Commission, and recording secretaries 

will attend the Closed Meeting. Certain 

staff members who have an interest in 

the matters also may be present. 

The General Counsel of the 

Commission, or his designee, has 

certified that, in his opinion, one or 

more of the exemptions set forth in 5 

U.S.C. 552b(c)(3), (5), (7), 9(B) and (10) 

and 17 CFR 200.402(a)(3), (5), (7), 9(ii) 

and (10), permit consideration of the 

scheduled matters at the Closed 

Meeting. 

Commissioner Aguilar, as duty 

officer, voted to consider the items 

listed for the Closed Meeting in a closed 

session. 

The subject matter of the Closed 

Meeting scheduled for Thursday, 

February 3, 2011 will be: 

Consideration of amicus participation; 

Institution and settlement of injunctive 

actions; 

Institution and settlement of 

administrative proceedings; and 

Other matters relating to enforcement 

proceedings. 

At times, changes in Commission 

priorities require alterations in the 

scheduling of meeting items. 

For further information and to 

ascertain what, if any, matters have been 

added, deleted or postponed, please 

contact: 

The Office of the Secretary at (202) 

551–5400. 

Dated: January 27, 2011. 

Elizabeth M. Murphy, 

Secretary. 

[FR Doc. 2011–2228 Filed 1–28–11; 11:15 am] 

BILLING CODE 8011–01–P 

SECURITIES AND EXCHANGE 

COMMISSION 

[Release No. 34–63776; File No. 0–49764] 

Notice and Opportunity for Hearing: 

SinoFresh Healthcare, Inc. 

January 26, 2011 

Notice is hereby given that on 

November 1, 2010, SinoFresh 

Healthcare, Inc. (Applicant) filed with 

the Securities and Exchange 

Commission a Form 15 certification 

(Certification) pursuant to Section 12(g) 

of the Securities Exchange Act of 1934 

(Exchange Act) for termination of the 

registration of the Applicant’s common 

shares (no par value) under Section 

12(g) of the Exchange Act. The 


srobinson on DSKHWCL6B1PROD with NOTICES