51222 Federal Register / Vol. 71, No. 167 / Tuesday, August 29, 2006 / Notices 

NUCLEAR REGULATORY 

COMMISSION 

Southern Nuclear Operating Company; 

Notice of Receipt and Availability of an 

Application for an Early Site Permit for 

the Vogtle ESP Site 

On August 15, 2006, the Nuclear 

Regulatory Commission (NRC, the 

Commission) received an application 

from Southern Nuclear Operating 

Company filed pursuant to Section 103 

of the Atomic Energy Act and 10 CFR 

part 52, for an early site permit (ESP) for 

a location in eastern Georgia (near 

Waynesboro, Georgia) identified as the 

Vogtle ESP site. 

An applicant may seek an ESP in 

accordance with Subpart A of 10 CFR 

part 52 separate from the filing of an 

application for a construction permit 

(CP) or combined license (COL) for a 

nuclear power facility. The ESP process 

allows resolution of issues relating to 

siting. At any time during the period of 

an ESP (up to 20 years), the permit 

holder may reference the permit in an 

application for a CP or COL. 

Subsequent Federal Register notices 

will address the acceptability of the 

tendered ESP application for docketing 

and provisions for participation of the 

public and other parties in the ESP 

review process. 

A copy of the application is available 

for public inspection at the 

Commission’s Public Document Room 

(PDR), located at One White Flint North, 

11555 Rockville Pike (first floor), 

Rockville, Maryland and via the 

Agencywide Documents Access and 

Management System (ADAMS) Public 

Electronic Reading Room on the Internet 

at the NRC Web site, http:// 

www.nrc.gov/reading-rm/adams.html. 

The accession number for the 

application is ML062290246. 

Future publicly available documents 

related to the application will also be 

posted in ADAMS. Persons who do not 

have access to ADAMS or who 

encounter problems in accessing the 

documents located in ADAMS should 

contact the NRC Public Document Room 

staff by telephone at 1–800–397–4209 or 

301–415–4737, or by e-mail to 

pdr@nrc.gov. The application is also 

available to local residents at the Burke 

County Library, in Waynesboro, 

Georgia, and it will be available on the 

NRC Web page at http://www.nrc.gov/ 

reactors/new-licensing/esp.html. 

Dated at Rockville, Maryland, this 21st day 

of August, 2006. 

For the Nuclear Regulatory Commission. 

David B. Matthews, 

Director, Division of New Reactor Licensing, 

Office of Nuclear Reactor Regulation. 

[FR Doc. E6–14285 Filed 8–28–06; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

Sunshine Act Notice 

DATE: Weeks of August 28, September 4, 

11, 18, 25, October 2, 2006. 

PLACE: Commissioners’ Conference 

Room, 1155 Rockville Pike, Rockville, 

Maryland. 

STATUS: Public and Closed. 

MATTERS TO BE CONSIDERED: 

Week of August 28, 2006 

There are no meetings scheduled for 

the week of August 28, 2006. 

Week of September 4, 2006—Tentative 

Wednesday, September 6, 2006 

1:50 p.m. Affirmation Session (Public) 

(Tentative) 

a. Pacific Gas & Elec. Co. (Diablo 

Canyon ISFSI), Docket No. 72–26–

ISFSI ‘‘Motion by San Luis Obispo 

Mothers for Peace, Sierra Club, and 

Peg Pinard for Declaratory and 

Injunctive Relief with respect to 

Diablo Canyon ISFSI’’. (Tentative). 

b. AmerGen Energy Company, LLC 

(License Renewal for Oyster Creek 

Nuclear Generating Station) Docket 

No. 50–0219, Legal challenges to 

LBP–06–07 and LBP–06–11. 

(Tentative). 

c. Pa’ina Hawaii, LLC, LBP–06–4, 63 

NRC 99 (2006) and LBP–06–12, 63 

NRC 409 (2006). (Tentative). 

Week of September 11, 2006—Tentative 

Monday, September 11, 2006 

9:30 a.m. Discussion of Security Issues 

(Closed—Ex. 1). 

1:30 p.m. Discussion of Security Issues 

(Closed—Ex. 1& 3). 

Tuesday, September 12, 2006 

9:30 a.m. Meeting with Organization of 

Agreement States (OAS) and 

Conference of Radiation Control 

Program Directors (CRCPD) (Public 

Meeting) (Contact: Shawn Smith, 

301–415–2620). 

This meeting will be webcast live at 

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

1 p.m. Discussion of Security Issues 

(Closed—Ex. 1). 

Week of September 18, 2006—Tentative 

There are no meetings scheduled for 

the week of September 18, 2006. 

Week of September 25, 2006—Tentative 

There are no meetings scheduled for 

the week of September 25, 2006. 

Week of October 2, 2006—Tentative 

There are no meetings scheduled for 

the week of October 2, 2006. 

* * * * * 

*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: 

Michelle Schroll, (301) 415–1661. 

* * * * * 

The NRC Commission Meeting 

Schedule can be found on the Internet 

at: http://www.nrc.gov/what-we-do/ 

policy-making/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 the 

NRC’s Disability Program Coordinator, 

Deborah Chan, at 301–415–7041, TDD: 

301–415–2100, or by e-mail at 

DLC@nrc.gov. Determinations on 

requests for reasonable accommodation 

will be made on a case-by-case basis. 

* * * * * 

This notice is distributed by mail to 

several hundred 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). 

In addition, distribution of this meeting 

notice over the Internet system is 

available. If you are interested in 

receiving this Commission meeting 

schedule electronically, please send an 

electronic message to dkw@nrc.gov. 

Dated: August 24, 2006. 

R. Michelle Schroll, 

Office of the Secretary. 

[FR Doc. 06–7236 Filed 8–25–06; 9:49 am] 

BILLING CODE 7590–01–M 

NUCLEAR REGULATORY 

COMMISSION 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a. (2) of the 

Atomic Energy Act of 1954, as amended 

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


jlentini on PROD1PC65 with NOTICES
Federal Register / Vol. 71, No. 167 / Tuesday, August 29, 2006 / Notices 51223 

Commission (the Commission or NRC 

staff) is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued from August 4, 

2006 to August 17, 2006. The last 

biweekly notice was published on 

August 15, 2006 (71 FR 46929). 

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 

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. Within 60 days after the 

date of publication of this notice, the 

licensee may file a request for a hearing 

with respect to issuance of the 

amendment to the subject facility 

operating license and any person whose 

interest may be affected by this 

proceeding and who wishes to 

participate as a party in the proceeding 

must file a written request for a hearing 

and a petition for leave to intervene. 

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 and 

Directives Branch, 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 delivered to Room 6D22, Two 

White Flint North, 11545 Rockville 

Pike, Rockville, Maryland, from 7:30 

a.m. to 4:15 p.m. Federal workdays. 

Copies of written comments received 

may be examined at the Commission’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. The filing of 

requests for a hearing and petitions for 

leave to intervene is discussed below. 

Within 60 days after the date of 

publication of this notice, the licensee 

may file a request for a hearing with 

respect to issuance of the amendment to 

the subject facility operating license and 

any person whose interest may be 

affected by this proceeding and who 

wishes to participate as a party in the 

proceeding must file a written request 

for a hearing and a petition for leave to 

intervene. 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 persons should 

consult a current copy of 10 CFR 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

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

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

request for a hearing or petition for 

leave to intervene is filed 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 petitioner/ 

requestor 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 petitioner/requestor 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 petitioner/requestor 

intends to rely in proving the contention 

at the hearing. The petitioner/requestor 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the petitioner/requestor 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 petitioner/ 

requestor to relief. A petitioner/ 

requestor 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 


jlentini on PROD1PC65 with NOTICES
51224 Federal Register / Vol. 71, No. 167 / Tuesday, August 29, 2006 / Notices 

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, any hearing held would 

take place before the issuance of any 

amendment. 

A request for a hearing or a petition 

for leave to intervene must be filed 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; (2) courier, express 

mail, and expedited delivery services: 

Office of the Secretary, Sixteenth Floor, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland, 20852, 

Attention: Rulemaking and 

Adjudications Staff; (3) E-mail 

addressed to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

HearingDocket@nrc.gov; or (4) facsimile 

transmission addressed to the Office of 

the Secretary, U.S. Nuclear Regulatory 

Commission, Washington, DC, 

Attention: Rulemakings and 

Adjudications Staff at (301) 415–1101, 

verification number is (301) 415–1966. 

A copy of the request for hearing and 

petition for leave to intervene should 

also be sent to the Office of the General 

Counsel, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and it is requested that copies be 

transmitted either by means of facsimile 

transmission to (301) 415–3725 or by email to OGCMailCenter@nrc.gov. A copy 

of the request for hearing and petition 

for leave to intervene should also be 

sent to the attorney for the licensee. 

Nontimely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission or the presiding officer of 

the Atomic Safety and Licensing Board 

that the petition, request and/or the 

contentions should be granted based on 

a balancing of the factors specified in 10 

CFR 2.309(a)(1)(i)–(viii). 

For further details with respect to this 

action, see the application for 

amendment which is available for 

public inspection at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the ADAMS Public 

Electronic Reading Room on the Internet 

at the NRC Web site, http:// 

www.nrc.gov/reading-rm/adams.html. 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@nrc.gov. 

Carolina Power & Light Company, 

Docket No. 50–261, H. B. Robinson 

Steam Electric Plant, Unit No. 2, 

Darlington County, South Carolina 

Date of amendment request: April 11, 

2006. 

Description of amendment request: 

The proposed amendment would 

modify Technical Specification 5.6.5 

‘‘Core Operating Limits Report (COLR)’’

to add two U.S. Nuclear Regulatory 

Commission-approved topical reports to 

the COLR methodologies list. These 

topical reports allow the use of S–

RELAP5 thermal-hydraulic analysis 

code for accident safety analyses. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. The Proposed Change Does Not Involve 

a Significant Increase in the Probability or 

Consequences of an Accident Previously 

Evaluated. 

The two topical reports have been 

reviewed and approved by the NRC for use 

in determining core operating limits. The 

core operating limits to be developed using 

the new methodologies for HBRSEP [H. B. 

Robinson Steam Electric Plant], Unit No. 2, 

will be established in accordance with the 

applicable limitations as documented in the 

NRC Safety Evaluation Reports. In a May 11, 

2001, NRC Safety Evaluation Report, the NRC 

concluded that the S–RELAP5 code is 

capable of addressing the thermal-hydraulic 

response of the target non-LOCA [loss-ofcoolant accident] events in a conservative 

manner and is, therefore, an acceptable 

replacement for the ANF–RELAP code. In the 

May 19, 2004, Safety Evaluation Report for 

Revision 1 to EMF–2310(P)(A), the NRC 

concluded that the code remained acceptable 

for use for the non-LOCA events. In a March 

15, 2001, Safety Evaluation Report, the NRC 

concluded that the code was acceptable for 

use for small break LOCA analyses at 

Westinghouse pressurized water reactors. 

The proposed change, by itself, does not 

impact the current design bases. The 

proposed change enables the use of new 

methodologies to re-analyze certain events. 

Revised analyses may either result in 

continued conformance with design bases, or 

may change the design bases. If design basis 

changes result from a revised analysis, then 

the specific design changes will be evaluated 

in accordance with HBRSEP, Unit No. 2, 

design change procedures and 10 CFR 50.59. 

The proposed change does not involve 

physical changes to any plant structure, 

system, or component. Therefore, the 

probability of occurrence for a previously 

analyzed accident is not significantly 

increased. 

The consequences of a previously analyzed 

accident are dependent on the initial 

conditions assumed for the analysis, the 

behavior of the fission product barriers 

during the analyzed accident, the availability 

and successful functioning of the equipment 

assumed to operate in response to the 

analyzed event, and the setpoints at which 

these actions are initiated. The proposed 

methodologies will ensure that the plant 

continues to meet applicable design and 

safety analyses acceptance criteria. The 

proposed change does not affect the 

performance of any equipment used to 

mitigate the consequences of an analyzed 

accident. As a result, no analysis 

assumptions are impacted and there are no 

adverse effects on the factors that contribute 

to offsite or onsite dose as a result of an 

accident. The proposed change does not 

affect setpoints that initiate protective or 

mitigative actions. The proposed change 

ensures that plant structures, systems, and 

components are maintained consistent with 

the safety analysis and licensing bases. Based 

on this evaluation, there is no significant 

increase in the consequences of a previously 

analyzed accident. 

Therefore, this change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. The Proposed Change Does Not Create 

the Possibility of a New or Different Kind of 

Accident From Any Previously Evaluated. 

The proposed change does not involve any 

physical alteration of plant systems, 

structures, or components, other than 

allowing for fuel design in accordance with 

NRC-approved methodologies. No new or 

different equipment is being installed. No 

installed equipment is being operated in a 

different manner. There is no change to the 

parameters within which the plant is 

normally operated or in the setpoints that 

initiate protective or mitigative actions. As a 

result, no new failure modes are being 

introduced. Therefore, the proposed change 

does not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

3. The Proposed Change Does Not Involve 

a Significant Reduction in the Margin of 

Safety. 

There is no impact on any margin of safety 

resulting from the incorporation of these new 

topical reports into the Technical 


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Federal Register / Vol. 71, No. 167 / Tuesday, August 29, 2006 / Notices 51225 

Specifications. If design basis changes result 

from a revised analysis that uses these new 

methodologies, the specific design changes 

will be evaluated in accordance with 

HBRSEP, Unit No. 2, design change 

procedures and 10 CFR 50.59. Any potential 

reduction in the margin of safety would be 

evaluated for that specific design change. 

Therefore, this change does not involve a 

significant reduction in the margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 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 T. 

Conley, Associate General Counsel II—

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, North Carolina 27602. 

NRC Branch Chief: L. Raghavan. 

Carolina Power & Light Company, 

Docket No. 50–261, H. B. Robinson 

Steam Electric Plant, Unit No. 2, 

Darlington County, South Carolina 

Date of amendment request: July 17, 

2006. 

Description of amendment request: 

The proposed amendment would revise 

the containment pressure requirements 

specified in Surveillance Requirements 

3.6.8 and 5.5.16 due to a revision in the 

Loss-of-Coolant Accident (LOCA) 

containment pressure analysis. 

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? 

No. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. The revised post-LOCA 

containment pressure and temperature 

analysis used more conservative assumptions 

and the increase in the calculated peak 

pressure was approximately 1 psig. The 

revised value of 41.49 psig remains less than 

the containment design pressure of 42 psig. 

The increase in the calculated peak 

temperature was approximately 2 °F, which 

was analyzed to have no impact on structures 

or equipment. Although there is an increase 

in the calculated pressure, the allowable 

containment leakage rate, as measured at the 

peak pressure, is not being changed. Since 

there is no increase in the allowable leakage, 

there is no increase in consequences. The 

proposed change is related to containment 

pressure analysis. There are no physical 

changes being made to the plant, or to the 

manner in which the plant is operated. 

Surveillance procedures for containment 

leakage have been conservatively testing at 

pressures in excess of 42 psig and 

surveillance procedures for the Isolation 

Valve Seal Water System have been 

conservatively testing at pressures in excess 

of 46.2 psig. The change can have no impact 

on the probability of an accident occurring. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

No. The proposed change does not create 

the possibility of a new or different kind of 

accident from any previously evaluated. 

There are no physical changes being made to 

the plant or to the manner in which the plant 

is operated. Surveillance procedures for 

containment leakage have been 

conservatively testing at pressures in excess 

of 42 psig and surveillance procedures for the 

Isolation Valve Seal Water System have been 

conservatively testing at pressures in excess 

of 46.2 psig. The revised containment 

analysis results in a calculated peak 

containment pressure that remains less than 

the containment design pressure. The 

increase in the calculated peak temperature 

was analyzed to have no impact on structures 

or equipment. Therefore, this change does 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in the margin of safety? 

No. The proposed change does not involve 

a significant reduction in the margin of 

safety. The proposed change imposes more 

conservative surveillance test requirements. 

The calculated increase in post-LOCA peak 

containment pressure is only 1 psig and the 

revised value of 41.49 psig remains less than 

the containment design pressure of 42 psig. 

The increase in the calculated peak 

temperature was approximately 2 °F, which 

was analyzed to have no impact on structures 

or equipment. Although there was an 

increase in the calculated pressure, the 

allowable containment leakage rate, as 

measured at the peak pressure, is not being 

changed. Therefore, this 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: David T. 

Conley, Associate General Counsel II—

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, North Carolina 27602. 

NRC Branch Chief: L. Raghavan. 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of amendment request: July 12, 

2006. 

Description of amendment request: 

The proposed amendment would 

modify Conditions, Required Actions 

and Completion Times associated with 

the inoperability of one or more 

emergency diesel generators (EDGs) in 

Technical Specification (TS) 3.8.1. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

This license amendment request proposes 

a change to extend the Technical 

Specification 3.8.1, ‘‘AC Sources-Operating,’’

Completion Time. This change allows a 

single EDG to be inoperable for 7 days more 

than Technical Specification 3.8.1 currently 

provides. The Required Actions for CTG 

[Combustion Turbine Generator] 11–1 are 

also removed from Condition A and TSTF 

[Technical Specification Task Force] -439 is 

implemented for TS 3.8.1, removing the 

second Completion Times. 

The EDGs are safety related components 

which provide backup electrical power 

supply to the onsite ESF [Engineered Safety 

Feature] power distribution system. CTG 11–

1 provides backup electrical power to the 

Division 1 power distribution system. 

Neither the EDGs nor CTG 11–1 are accident 

initiators, thus these changes do not increase 

the probability of a previously evaluated 

accident. 

The plant ESF power distribution systems 

consist of two divisions for 100% 

redundancy. Accident analyses demonstrate 

that only one division is required for 

accident mitigation. Thus, with one division 

inoperable the other division is capable of 

performing the required safety function. 

Design basis analyses are not required to be 

performed assuming extended loss of all 

power supplies to the plant ESF power 

distribution system. Thus, this change does 

not involve a significant increase in the 

consequences of a previously analyzed 

accident. 

The proposed change also eliminates the 

second Completion Time from TS 3.8.1. 

These second Completion Times are not an 

initiator to any accident previously 

evaluated. As a result, the probability of an 

accident previously evaluated is not affected. 

The consequences of an accident during the 

revised Completion Times are no different 

than the consequences of the same accident 

during the existing Completion Times. As a 

result, the consequences of an accident 

previously evaluated are not affected by this 

change. Therefore, the proposed change does 

not involve a significant increase in the 


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51226 Federal Register / Vol. 71, No. 167 / Tuesday, August 29, 2006 / Notices 

probability or consequences of an accident 

previously evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The changes do not involve a physical 

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

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. The changes do not alter any 

assumptions made in the safety analysis. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. The change does not involve a 

significant reduction in the margin of safety. 

This license amendment request proposes 

Technical Specification changes to extend 

the Technical Specification 3.8.1, ‘‘AC 

Sources-Operating,’’ Completion Time for an 

inoperable EDG to 14 days. These changes 

allow an emergency diesel generator to be 

inoperable for 7 days more than TS 3.8.1 

currently provides. 

Deterministic and probabilistic risk 

assessments evaluated the effect of the 

proposed TS changes on the availability of an 

electrical power supply to the plant 

emergency safeguards features systems. 

These assessments concluded that the 

proposed TS changes do not involve a 

significant increase in the risk of power 

supply unavailability. 

This license amendment request proposes 

TS changes to remove the Required Actions 

for CTG 11–1 from TS 3.8.1 Condition A. If 

CTG 11–1 is inoperable at the same time that 

any single EDG is inoperable for the entire 

proposed 14 day period with no other 

equipment in maintenance, the risk remains 

within RG [Regulatory Guide] 1.174 

(Reference 2 [in the application]) thresholds 

for a ‘‘very small’’ classification. 

The proposed change to delete the second 

Completion Time does not alter the manner 

in which safety limits, limiting safety system 

settings, or limiting conditions for operation 

are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed changes will not result 

in plant operation in a configuration outside 

of the design basis. 

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 G. 

Pettinari, Legal Department, 688 WCB, 

Detroit Edison Company, 2000 2nd 

Avenue, Detroit, Michigan 48226–1279. 

NRC Acting Branch Chief: Martin 

Murphy. 

Dominion Nuclear Connecticut, Inc., 

Docket No. 50–336, Millstone Power 

Station, Unit No. 2, (MPS2) New London 

County, Connecticut 

Date of amendment request: June 13, 

2006. 

Description of amendment request: 

The proposed amendment would allow 

changes to the Technical Specifications 

based on the radiological dose analysis 

margins obtained by using an alternate 

source term consistent with 10 CFR 

50.67. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Involve a significant increase in the 

probability or consequences of an accident 

previously evaluated? 

No. 

The proposed amendment does not involve 

a significant increase in the probability or 

consequence of an accident previously 

analyzed. The MPS2 control room emergency 

ventilation system only functions following 

the initiation of a design basis radiological 

accident. Therefore, the change to the value 

used for in-leakage rate test acceptance 

criteria following a design basis accident will 

not increase the probability of any previously 

analyzed accident. The proposed 200 cfm 

control room habitability envelope inleakage 

surveillance acceptance criteria has no 

adverse impact on control room habitability 

analyses for postulated toxic chemical release 

events. These habitability analyses do not 

credit automatic or manual isolation of the 

control room fresh air ventilation flow during 

a toxic chemical release event. The control 

room’s forced ventilation fresh air exchange 

rate (e.g., 800 cfm) is much greater than the 

proposed 200 cfm envelope inleakage rate 

acceptance criteria. The MPS2 containment 

purge valve isolation signal is not credited in 

the accident analyses. The requirements 

contained in this specification do not meet 

any of 10 CFR 50.36(c)(2)(ii) criteria on items 

for which technical specifications must be 

established. Deletion of this technical 

specification will not increase the probability 

of any previously analyzed accident. The 

MPS2 containment and the containment 

systems function to prevent or control the 

release of radioactive fission products 

following a postulated accident. Therefore, 

the change to the value used for primary to 

secondary leak rate acceptance criteria, and 

for all secondary containment bypass leakage 

paths following a design basis accident, will 

not increase the probability of any previously 

analyzed accident. 

These systems are not initiators of any 

design bases accident. Revised dose 

calculations, which take into account the 

changes proposed by this amendment and 

the use of the alternative source term, have 

been performed for the MPS2 design basis 

radiological accidents. The results of these 

revised calculations indicate that public and 

control room doses will not exceed the limits 

specified in 10 CFR 50.67 and Regulatory 

Guide 1.183. There is not a significant 

increase in predicted dose consequences for 

any of the analyzed accidents. Therefore, the 

proposed changes do not involve a 

significant increase in the consequences of 

any previously analyzed accident. 

2. Create the possibility of a new or 

different kind of accident from any accident 

previously evaluated? 

No. 

The implementation of the proposed 

changes does not create the possibility of an 

accident of a different type than was 

previously evaluated in the UFSAR. 

Although the proposed changes could affect 

the operation of the control room emergency 

ventilation system, and the containment and 

containment systems following a design basis 

radiological accident, none of these changes 

can initiate a new or different kind of 

accident since they are only related to system 

capabilities that provide protection from 

accidents that have already occurred. These 

changes do not alter the nature of events 

postulated in the UFSAR nor do they 

introduce any unique precursor mechanisms. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from those previously 

analyzed. 

3. Involve a significant reduction in the 

margin of safety? 

No. 

The implementation of the proposed 

changes does not reduce the margin of safety. 

The proposed changes for the control room 

ventilation system, and the containment and 

containment systems do not affect the ability 

of these systems to perform their intended 

safety functions to maintain dose less than 

the required limits during design basis 

radiological events. The radiological analysis 

results, when compared with the revised 

TEDE [total effective dose equivalent] 

acceptance criteria, meet the applicable 

limits. These acceptance criteria have been 

developed for application to analyses 

performed with alternative source terms. 

These acceptance criteria have been 

developed for the purpose of use in design 

basis accident analyses such that meeting the 

stated limits demonstrates adequate 

protection of public health and safety. It is 

thus concluded that the margin of safety will 

not be reduced by the implementation of the 

changes. 

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: Lillian M. 

Cuoco, Senior Nuclear Counsel, 

Dominion Nuclear Connecticut, Inc., 

Rope Ferry Road, Waterford, CT 06385. 

NRC Acting Branch Chief: Brooke D. 

Poole. 


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Dominion Nuclear Connecticut, Inc., 

Docket No. 50–423, Millstone Power 

Station, Unit No. 3, New London 

County, Connecticut 

Date of amendment request: July 19, 

2006. 

Description of amendment request: 

The proposed amendment will revise 

reactor core safety limits Technical 

Specifications (TSs) and relocate the 

reactor core safety limit figure to the 

Core Operating Limits Report (COLR) in 

the Millstone Power Station, Unit 3, 

Technical Requirements Manual. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

Criterion 1: 

Does the proposed amendment involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The relocation of cycle-specific core 

operating limits from the TS to the COLR and 

the addition of the RCS [reactor coolant 

system] total flow rate to the COLR has no 

influence or impact on the probability or 

consequences of a design basis accident. 

Adherence to the COLR and methodologies 

acceptable for establishing COLR parameters 

continues to be controlled by TS. The 

proposed amendment still requires exactly 

the same actions to be taken when or if limits 

are exceeded. Each accident analysis 

addressed in the final safety analysis report 

(FSAR) will be examined with respect to 

changes in cycle-dependent parameters, 

which are obtained from application of the 

NRC-approved reload design methodologies, 

to ensure that the transient evaluation of new 

core designs are bounded by previously 

accepted analyses. This examination, which 

will be performed in accordance with the 

requirements of 10 CFR 50.59, ensures that 

future core designs will not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Criterion 2: 

Does the proposed amendment create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The relocation and addition of the cyclespecific variables to the COLR and adding 

new document references in TS Section 

6.9.1.6.b does not influence or impact, nor 

does it contribute in any way to the 

probability or consequences of an accident. 

No safety-related equipment, safety function, 

or plant operations will be altered as a result 

of these proposed changes. The cycle-specific 

variables are calculated using NRC-approved 

methods and submitted to the NRC to allow 

the staff to continue to trend the values of 

these limits. The TS will continue to require 

operation within the required core operating 

limits and appropriate actions will be taken 

when or if limits are exceeded. Therefore the 

proposed amendment does not in any way 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

Criterion 3: 

Does the proposed amendment involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes have no impact on 

plant equipment operation. The proposed 

changes do not revise any setpoints or 

acceptance criteria assumed in the accident 

analyses. Therefore, the proposed changes 

will not result in a reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the 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 Nuclear Counsel, 

Dominion Nuclear Connecticut, Inc., 

Rope Ferry Road, Waterford, CT 06385. 

NRC Acting Branch Chief: Brooke D. 

Poole. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit No. 2 (IP2), 

Westchester County, New York 

Date of amendment request: July 10, 

2006. 

Description of amendment request: 

Energy Nuclear Operations, Inc., is 

planning to operate an Independent 

Spent Fuel Storage Installation (ISFSI) 

facility at IP2 using the HOLTEC HISTORM 100 Cask System. To support 

this activity, the proposed amendment 

adds Spent Fuel Cask loading 

requirements to IP2 Technical 

Specifications (TSs). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change will revise the Indian 

Point Unit 2 TSs associated with the SFP 

[spent fuel pool] to assure that the regulatory 

requirements related to criticality in the SFP 

and applied to the Holtec HI-STORM 100 

Multi-Purpose Canister MPC–32 when in the 

SFP are reflected in the IP2 TS. The proposed 

change does not require any physical changes 

to Part 50 structures, systems, or 

components, nor will their performance 

requirements be altered. Therefore, the 

response of the plant to previously analyzed 

accidents and related radiological releases 

will not be adversely impacted, and will 

bound those postulated during cask loading 

activities in the cask storage area. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Existing fuel handling procedures and 

associated administrative controls remain 

applicable for cask loading operations within 

the SFP. Additionally, the soluble boron 

concentration required to maintain keff ≤ 

0.95 for postulated criticality accidents 

associated with cask loading operations was 

also evaluated. The results of the analyses, 

using a methodology previously approved by 

the NRC [Nuclear Regulatory Commission], 

demonstrate that the amount of soluble boron 

required to compensate for the positive 

reactivity associated with these postulated 

accidents (371 ppm [parts per million]) 

remains well below the existing spent fuel pit 

minimum boron concentration limit of 2000 

ppm. Accordingly, the same limit has been 

proposed for cask loading operations in the 

cask storage area. 

Therefore, the possibility of a new or 

different kind of accident from any accident 

previously evaluated is not created. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

An NRC approved methodology was used 

to perform the criticality analysis which 

provides the basis to incorporate a new 

family of burnup versus enrichment curves, 

for various cooling times, into the plant 

Technical Specifications to ensure criticality 

requirements are met during spent fuel cask 

loading. Accordingly, the existing minimum 

boron concentration limit for the spent fuel 

pit of 2000 ppm will continue to remain 

bounding during cask loading operations. 

This determination accounts for uncertainties 

at a 95 percent probability, 95 percent 

confidence level. Should it be postulated that 

a boron dilution event does occur during this 

time period, keff will remain less than 1.0 

should the cask storage area become fully 

flooded with unborated water. 

Therefore, there will not be 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. John Fulton, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc., 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Branch Chief: Richard J. Laufer. 


jlentini on PROD1PC65 with NOTICES
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Exelon Generation Company, LLC, 

Docket No. 50–249, Dresden Nuclear 

Power Station (DNPS), Unit 3, Grundy 

County, Illinois 

Date of amendment request: July 21, 

2006. 

Description of amendment request: 

The proposed amendment would revise 

the values of the safety limit minimum 

critical power ratio (SLMCPR) in 

Technical Specification Section 2.1.1, 

‘‘Reactor Core SLs [Safety Limits].’’

Specifically, the proposed change 

would require that for Unit 3, the 

minimum critical power ratio (MCPR) 

for Global Nuclear Fuel fuel shall be ≥ 

1.10 for two recirculation loop 

operation, or ≥ 1.11 for single 

recirculation loop operation. 

Additionally, the proposed change 

would require that MCPR for 

Westinghouse fuel shall be ≥ 1.12 for 

two recirculation loop operation, or ≥ 

1.14 for single recirculation loop 

operation. 

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 probability of an evaluated accident is 

derived from the probabilities of the 

individual precursors to that accident. The 

consequences of an evaluated accident are 

determined by the operability of plant 

systems designed to mitigate those 

consequences. Limits have been established 

consistent with NRC [Nuclear Regulatory 

Commission]-approved methods to ensure 

that fuel performance during normal, 

transient, and accident conditions is 

acceptable. The proposed change 

conservatively establishes the SLMCPR for 

DNPS, Unit 3, Cycle 20 such that the fuel is 

protected during normal operation and 

during plant transients or anticipated 

operational occurrences (AOOs). 

Changing the SLMCPR does not increase 

the probability of an evaluated accident. The 

change does not require any physical plant 

modifications, physically affect any plant 

components, or entail changes in plant 

operation. Therefore, no individual 

precursors of an accident are affected. 

The proposed change revises the SLMCPR 

to protect the fuel during normal operation 

as well as during plant transients or AOOs. 

Operational limits will be established based 

on the proposed SLMCPR to ensure that the 

SLMCPR is not violated. This will ensure 

that the fuel design safety criterion (i.e., that 

at least 99.9% of the fuel rods do not 

experience transition boiling during normal 

operation and AOOs) is met. Since the 

proposed change does not affect operability 

of plant systems designed to mitigate any 

consequences of accidents, the consequences 

of an accident previously evaluated are not 

expected to increase. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Creation of the possibility of a new or 

different kind of accident requires creating 

one or more new accident precursors. New 

accident precursors may be created by 

modifications of plant configuration, 

including changes in allowable modes of 

operation. The proposed change does not 

involve any plant configuration 

modifications or changes to allowable modes 

of operation. The proposed change to the 

SLMCPR assures that safety criteria are 

maintained for DNPS, Unit 3, Cycle 20. 

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 SLMCPR provides a margin of safety 

by ensuring that at least 99.9% of the fuel 

rods do not experience transition boiling 

during normal operation and AOOs if the 

MCPR limit is not violated. The proposed 

change will ensure the current level of fuel 

protection is maintained by continuing to 

ensure that at least 99.9% of the fuel rods do 

not experience transition boiling during 

normal operation and AOOs if the MCPR 

limit is not violated. The proposed SLMCPR 

values were developed using NRC-approved 

methods. Additionally, operational limits 

will be established based on the proposed 

SLMCPR to ensure that the SLMCPR is not 

violated. This will ensure that the fuel design 

safety criterion (i.e., that no more than 0.1% 

of the rods are expected to be in boiling 

transition if the MCPR limit is not violated) 

is met. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Assistant General Counsel, 

Exelon Generation Company, LLC, 200 

Exelon Way, Kennett Square, PA 19348. 

NRC Branch Chief: Daniel S. Collins. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–440, 

Perry Nuclear Power Plant, Unit 1 

(PNPP), Lake County, Ohio 

Date of amendment request: June 6, 

2006. 

Description of amendment request: 

The proposed amendment would revise 

the Ventilation Filter Test Program 

(VFTP) in Technical Specification (TS) 

5.5.7. The license amendment is a 

corrective action to revise the flow rate 

units specified in the VFTP from 

standard cubic feet per minute to cubic 

feet per minute. This amendment will 

ensure the PNPP TS are consistent with 

plant design documentation, testing 

criteria, and the industry. 

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: 

The ESF [Engineered Safety Feature] 

Ventilation systems reduce the concentration 

of airborne radioactive contaminants 

following a design basis accident and 

therefore are not initiators of design bases 

accidents. The proposed amendment does 

not change the manner in which the ESF 

ventilation systems are operated or tested. 

Implementation of the proposed amendment 

will ensure the ESF ventilation systems 

perform their function when called upon and 

does not affect the plant operations, design 

function or analysis that verifies the 

capability of a [plant] structures, systems or 

components. 

The proposed amendment does not affect 

the design of the ESF ventilation systems, the 

operational characteristics of the ESF 

ventilation systems, the interfaces between 

the ESF ventilation systems and those plant 

systems they support, or the reliability of the 

ESF ventilation systems. 

Therefore, the ESF ventilation systems will 

be capable of performing their accident 

mitigation function and there is no increase 

in the probability or consequences of an 

accident already evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: 

The proposed amendment introduces no 

new mode of plant operation and does not 

involve a physical modification to the plant. 

New equipment is not installed with the 

proposed amendment, nor does the proposed 

amendment cause existing equipment to be 

operated in a new or different manner. 

Since the proposed changes do not involve 

a change to the plant design or operation, no 

new system interactions are created by this 

change. The proposed amendment does not 

produce any parameters or conditions that 


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could contribute to the initiation of accidents 

different from those already evaluated in the 

Updated Safety Analysis Report. 

The changes to the VFTP do not affect the 

assumed accident performance of the ESF 

Ventilation systems, nor [sic] any plant 

structure, system or component previously 

evaluated. 

Therefore, the proposed amendment does 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: 

The proposed amendment does not impact 

the ESF ventilation systems performance, 

including the capability for each ESF 

ventilation system to attain and maintain 

required air flow assumed in the plant safety 

analysis. 

The proposed amendment does not involve 

a significant reduction in a margin of safety 

since the operability of the ESF ventilations 

[sic] systems continues to be determined as 

required to support the capability of the ESF 

ventilations [sic] systems to provide the 

required ventilation, filtration and 

temperature control to mitigate the 

consequences of an accident. 

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, Attorney, FirstEnergy 

Corporation, 76 South Main Street, 

Akron, OH 44308. 

NRC Branch Chief: Daniel S. Collins. 

Florida Power Corporation, et al., 

Docket No. 50–302, Crystal River Unit 3 

Nuclear Generating Plant, Citrus 

County, Florida 

Date of amendment request: May 25, 

2006. 

Description of amendment request: 

The proposed license amendment 

revises the requirements in the Crystal 

River Unit 3 Improved Technical 

Specification related to steam generator 

tube integrity. The licensee states that 

the changes are consistent with NRCapproved Technical Specification (TS) 

Task Force (TSTF) Change Traveler, 

TSTF–449, ‘‘Steam Generator Tube 

Integrity,’’ Revision 4. The availability 

of this technical specification 

improvement was announced in the 

Federal Register on May 6, 2005, as part 

of the consolidated line item 

improvement process. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of no significant 

hazards consideration 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 requires a SG [Steam 

Generator] Program that includes 

performance criteria that will provide 

reasonable assurance that the SG tubing will 

retain integrity over the full range of 

operating conditions (including startup, 

operation in the power range, hot standby, 

cooldown and all anticipated transients 

included in the design specification). The SG 

performance criteria are based on tube 

structural integrity, accident induced 

leakage, and operational LEAKAGE. 

A SGTR [steam generarator tube rupture] 

event is one of the design basis accidents that 

are analyzed as part of a plant’s licensing 

basis. In the analysis of a SGTR event, a 

bounding primary to secondary LEAKAGE 

rate equal to the operational LEAKAGE rate 

limits in the licensing basis plus the 

LEAKAGE rate associated with a doubleended rupture of a single tube is assumed. 

For other design basis accidents such as 

MSLB [main steamline break], rod ejection, 

and reactor coolant pump locked rotor the 

tubes are assumed to retain their structural 

integrity (i.e., they are assumed not to 

rupture). These analyses typically assume 

that primary to secondary LEAKAGE for all 

SGs is 1 gallon per minute or increases to 1 

gallon per minute as a result of accident 

induced stresses. The accident induced 

leakage criterion introduced by the proposed 

changes accounts for tubes that may leak 

during design basis accidents. The accident 

induced leakage criterion limits this leakage 

to no more than the value assumed in the 

accident analysis. 

The SG performance criteria proposed 

change to the TS identify the standards 

against which tube integrity is to be 

measured. Meeting the performance criteria 

provides reasonable assurance that the SG 

tubing will remain capable of fulfilling its 

specific safety function of maintaining 

reactor coolant pressure boundary integrity 

throughout each operating cycle and in the 

unlikely event of a design basis accident. The 

performance criteria are only a part of the SG 

Program required by the proposed change to 

the TS. The program, defined by NEI 

[Nuclear Energy Institute] 97–06, Steam 

Generator Program Guidelines, includes a 

framework that incorporates a balance of 

prevention, inspection, evaluation, repair, 

and leakage monitoring. The proposed 

changes do not, therefore, significantly 

increase the probability of an accident 

previously evaluated. 

The consequences of design basis accidents 

are, in part, functions of the DOSE 

EQUIVALENT 1–131 in the primary coolant 

and the primary to secondary LEAKAGE 

rates resulting from an accident. Therefore, 

limits are included in the plant technical 

specifications for operational leakage and for 

DOSE EQUIVALENT 1–131 in primary 

coolant to ensure the plant is operated within 

its analyzed condition. The typical analysis 

of the limiting design basis accident assumes 

that primary to secondary leak rate after the 

accident is 1 gallon per minute with no more 

than [500 gallons per day or 720 gallons per 

day] in any one SG, and that the reactor 

coolant activity levels of DOSE 

EQUIVALENT 1–131 are at the TS values 

before the accident. 

The proposed change does not affect the 

design of the SGs, their method of operation, 

or primary coolant chemistry controls. The 

proposed approach updates the current TSs 

and enhances the requirements for SG 

inspections. The proposed change does not 

adversely impact any other previously 

evaluated design basis accident and is an 

improvement over the current TSs. 

Therefore, the proposed change does not 

affect the consequences of a SGTR accident 

and the probability of such an accident is 

reduced. In addition, the proposed changes 

do not affect the consequences of an MSLB, 

rod ejection, or a reactor coolant pump 

locked rotor event, or other previously 

evaluated accident. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident from any Previously 

Evaluated 

The proposed performance based 

requirements are an improvement over the 

requirements imposed by the current 

technical specifications. Implementation of 

the proposed SG Program will not introduce 

any adverse changes to the plant design basis 

or postulated accidents resulting from 

potential tube degradation. The result of the 

implementation of the SG Program will be an 

enhancement of SG tube performance. 

Primary to secondary LEAKAGE that may be 

experienced during all plant conditions will 

be monitored to ensure it remains within 

current accident analysis assumptions. 

The proposed change does not affect the 

design of the SGs, their method of operation, 

or primary or secondary coolant chemistry 

controls. In addition, the proposed change 

does not impact any other plant system or 

component. The change enhances SG 

inspection requirements. 

Therefore, the proposed change does not 

create the possibility of a new or different 

type of accident from any accident 

previously evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in the Margin 

of Safety 

The SG tubes in pressurized water reactors 

are an integral part of the reactor coolant 

pressure boundary and, as such, are relied 

upon to maintain the primary system’s 

pressure and inventory. As part of the reactor 

coolant pressure boundary, the SG tubes are 

unique in that they are also relied upon as 

a heat transfer surface between the primary 

and secondary systems such that residual 

heat can be removed from the primary 

system. In addition, the SG tubes isolate the 

radioactive fission products in the primary 

coolant from the secondary system. In 

summary, the safety function of an SG is 

maintained by ensuring the integrity of its 

tubes. 

Steam generator tube integrity is a function 

of the design, environment, and the physical 

condition of the tube. The proposed change 

does not affect tube design or operating 

environment. The proposed change is 

expected to result in an improvement in the 


jlentini on PROD1PC65 with NOTICES
51230 Federal Register / Vol. 71, No. 167 / Tuesday, August 29, 2006 / Notices 

tube integrity by implementing the SG 

Program to manage SG tube inspection, 

assessment, repair, and plugging. The 

requirements established by the SG Program 

are consistent with those in the applicable 

design codes and standards and are an 

improvement over the requirements in the 

current TSs. 

For the above reasons, the margin of safety 

is not changed and overall plant safety will 

be enhanced by the proposed change to the 

TS. 

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David T. 

Conley, Associate General Counsel II—

Legal Department, Progress Energy 

Service Company, LLC, Post Office Box 

1551, Raleigh, North Carolina 27602. 

NRC Branch Chief: L. Raghavan. 

Nuclear Management Company, LLC, 

Docket No. 50–266, Point Beach Nuclear 

Plant (PBNP), Unit 1, Town of Two 

Creeks, Manitowoc County, Wisconsin 

Date of amendment request: July 11, 

2006. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification (TS) 5.5.8, 

‘‘Steam Generator (SG) Program,’’ to 

exclude the portion of the tube below 17 

inches from the top of the tubesheet 

from the SG tube inspection 

requirements for Unit 1 on a one-time 

basis for a single operating cycle. In 

addition, administrative changes are 

proposed to correct a page number in 

the TS table of contents and delete two 

blank pages in TS Section 5. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration which is presented below. 

1. Operation of the Point Beach Nuclear 

Plant in accordance with the proposed 

amendments does not result in a significant 

increase in the probability or consequences 

of any accident previously evaluated. 

The proposed change revises Technical 

Specification (TS) 5.5.8, ‘‘Steam Generator 

(SG) Program’’ to redefine the PBNP Unit 1 

primary pressure boundary for purposes of 

the SG tube inspection requirements on a 

one-time basis for Unit 1 Refueling Outage 30 

and the subsequent operating cycle. The 

redefined primary pressure boundary is 

relocated from the seal weld at the bottom of 

the SG tube to the tube-to-tubesheet 

mechanical interface. The required structural 

integrity margins of the SG tubes in this area 

are unaffected by this change and will be 

maintained by the SG tubesheet. SG tubes are 

hydraulically expanded into the tubesheet. 

Steam generator tube rupture is constrained 

by the tubesheet for tubes with cracks in the 

tubesheet. This constraint results from the 

hydraulic expansion process which restricts 

further expansion of the tube, thermal 

expansion mismatch between the tube and 

tubesheet and from the differential pressure 

between the primary and secondary side. 

Thermal expansion and differential pressure 

also restrain the tube axially. For 

conservatism, hydraulic preload was not 

factored into the analysis. 

The proposed change continues to require 

that the SG Program include performance 

criteria that will provide reasonable 

assurance that the SG tubing will retain 

integrity over the full range of operating 

conditions (including startup, operation in 

the power range, hot standby, cooldown and 

all anticipated transients included in the 

design specification). 

The SG performance criteria are based on 

tube structural integrity, accident induced 

leakage, and operational LEAKAGE. The 

analysis shows that structural integrity 

retains acceptable safety factors against burst 

under normal steady state full power 

operation primary-to-secondary pressure 

differential and against burst applied to the 

design basis accident primary-to-secondary 

pressure differentials. The analysis also 

shows that accident induced leakage is 

bound by twice the normal operating leakage 

and well below the accident analysis 

assumption for each stream generator. The 

primary to secondary operational LEAKAGE 

limit is not changed. 

The planned inspection and supporting 

analysis provide reasonable assurance that 

the SG tubing will remain capable of 

fulfilling its specific safety function of 

maintaining reactor coolant pressure 

boundary integrity throughout the operating 

cycle and in the unlikely event of a design 

basis accident. The proposed change does 

not, therefore, significantly increase the 

probability of an accident previously 

evaluated. 

The consequences of design basis accidents 

are, in part, functions of the DOSE 

EQUIVALENT I–131 in the primary coolant 

and the primary to secondary LEAKAGE 

rates resulting from an accident. The plant 

technical specification limits for operational 

LEAKAGE and for DOSE EQUIVALENT I–

131 in primary coolant, which ensure the 

plant is operated within its analyzed 

condition, are unaffected by the proposed 

change. Therefore, the proposed change does 

not significantly increase the consequences 

of any accident previously evaluated. 

The proposed change does not significantly 

affect the probability of any event initiators. 

There will be no change to normal plant 

operating parameters, engineered safety 

feature actuation setpoints, accident 

mitigation capabilities, or accident analysis 

assumptions or inputs. 

Therefore, the probability or consequences 

of any accident previously evaluated will not 

be significantly increased as a result of the 

proposed change. 

2. Operation of the Point Beach Nuclear 

Plant in accordance with the proposed 

amendments does not result in a new or 

different kind of accident from any accident 

previously evaluated. 

Implementation of the proposed change 

will not introduce any adverse changes to the 

plant design basis or postulated accidents 

resulting from potential tube degradation. 

Primary to secondary leakage that may be 

experienced during all plant conditions will 

continue to be monitored to ensure it remains 

within current accident analysis 

assumptions. The proposed change does not 

affect the method of operation of the SGs, or 

primary or secondary coolant chemistry 

controls. In addition, the proposed change 

does not impact any other plant system or 

component. 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures are introduced as a result of 

the proposed change. Equipment important 

to safety will continue to operate as designed. 

The changes do not result in any event 

previously deemed incredible being made 

credible. The changes do not result in 

adverse conditions or result in any increase 

in the challenges to safety systems. 

Therefore, the proposed change does not 

create the possibility of a new or different 

type of accident from any accident 

previously evaluated. 

3. Operation of the Point Beach Nuclear 

Plant in accordance with the proposed 

amendments does not result in a significant 

reduction in a margin of safety. 

The steam generators (SGs) are an integral 

part of the reactor coolant pressure boundary 

and, as such, are relied upon to maintain the 

primary system’s pressure and inventory. 

They are also relied upon to remove residual 

heat from the primary system. The safety 

function of an SG is maintained by ensuring 

the integrity of its tubes. Steam generator 

tube integrity is a function of the design, 

environment, and the physical condition of 

the tube. The proposed change redefines the 

PBNP Unit 1 primary pressure boundary 

from the tube end weld to 17 inches below 

the top of the tubesheet and incorporates 

revisions to the inspection criteria for SG 

tube inspection in the tubesheet. The SG 

operating environment is not affected by the 

change. The proposed change maintains the 

required structural margins of the SG tubes 

for both normal and accident conditions. 

For cracking located within the tubesheet, 

steam generator tube rupture is constrained 

by the tubesheet. For circumferentially 

oriented cracking, the associated analysis for 

the proposed change validates that 17 inches 

of degradation free expanded tubing provides 

the necessary resistance to tube pullout with 

applicable safety factors applied. 

The revised inspection criteria continue to 

verify SG tube integrity. The safety function 

of the affected components will be 

maintained with the redefined primary 

pressure boundary. 

There are no new or significant changes to 

the initial conditions contributing to accident 

severity or consequences. The proposed 

amendment will not otherwise affect the 

plant protective boundaries, will not cause a 

release of fission products to the public, nor 

will it degrade the performance of any other 

structures, systems or components (SSCs) 

important to safety. Therefore, the requested 

change will not result in a significant 

reduction in the margin of safety. 

The NRC staff has reviewed the licensee’s 

analysis and, based on this review, it appears 

that the three standards of 10 CFR 50.92(c) 

are satisfied. Therefore, the NRC staff 


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proposes to determine that the amendment 

request involves no significant hazards 

consideration. 

Attorney for licensee: Jonathan Rogoff, 

Esquire, Vice President, Counsel & Secretary, 

Nuclear Management Company, LLC, 700 

First Street, Hudson, WI 54016. 

NRC Acting Branch Chief: Martin Murphy. 

PPL Susquehanna, LLC, Docket Nos. 50–387 

and 50–388, Susquehanna Steam Electric 

Station, Units 1 and 2 (SSES 1 and 2), 

Luzerne County, Pennsylvania 

Date of amendment request: October 13, 

2005. 

Description of amendment request: The 

proposed amendment would modify the 

licensing bases of SSES 1 and 2 by adopting 

the Alternative Source Term (AST) 

methodology which replaces the current 

accident source term with an AST. The AST 

is characterized by the composition and 

magnitude of the radioactive material, the 

chemical and physical form of the 

radionuclides, and the timing of the releases 

of these radionuclides. The exceptions would 

be that the current Technical Information 

Document (TID)14844 accident source term 

would remain the licensing basis for (1) 

equipment qualification, (2) NUREG–0737 

evaluations other than Control Room 

Habitability Envelope (CRHE) doses, and (3) 

Final Safety Analysis Report (FSAR) 

accidents not included in Regulatory Guide 

1.183. 

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 of 

occurrence or consequences of an accident 

previously evaluated? 

Response: No. 

Adoption of the AST and pursuant TS 

[Technical Specification] changes, changes to 

the TS’s to address NRC Generic Letter 2003–

01 (Reference 12.1) [see application dated 

October 13, 2005] and the changes to the 

atmospheric dispersion factors, have no 

impact to the initiation of DBAs [design basis 

accidents]. Once the occurrence of an 

accident has been postulated, the new 

accident source term and atmospheric 

dispersion factors are an input to analyses 

that evaluate the radiological consequences. 

Some of the proposed changes do affect the 

design or manner in which the facility is 

operated following an accident; however, the 

proposed changes do not involve a revision 

to the design or manner in which the facility 

is operated that could increase in the 

probability of an accident previously 

evaluated of a DBA discussed in Chapter 15 

of the FSAR. 

Therefore, the proposed change does not 

involve an increase in the probability of an 

accident previously evaluated. 

The structures, systems and components 

affected by the proposed changes act as 

mitigators to the consequences of accidents. 

Based on the revised analyses, the proposed 

changes do revise certain performance 

requirements; however, the proposed 

changes do not involve a revision to the 

parameters or conditions that could 

contribute to the initiation of a DBA 

discussed in Chapter 15 of the FSAR. 

Plant-specific radiological analyses have 

been performed using the AST methodology 

and new atmospheric dispersion factors. 

Based on the results of these analyses, it had 

been demonstrated that the CRHE dose 

consequences of the limiting events 

considered in the analyses meet the 

regulatory guidance provided for use with 

the AST, and the offsite doses are well within 

acceptable limits. This guidance is presented 

in 10 CFR 50.67, RG [Regulatory Guide] 

1.183, and Standard Review Plan [SRP] 

Section 15.0.1. 

Therefore, the proposed amendment does 

not result in a significant increase in the 

consequences of any previously evaluated 

accident. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Implementation of AST and the associated 

proposed TS changes and new atmospheric 

dispersion factors do not alter or involve any 

design basis accident initiators. These 

changes do not affect the design function or 

mode of operation of structures, systems and 

components in the facility prior to a 

postulated accident. Since structures, 

systems and components are operated 

essentially no differently after the AST 

implementation, no new failure modes are 

created by this proposed change. 

Licensing basis changes are proposed and 

justified to credit use of the SLC [Standby 

Liquid Control] System to buffer suppression 

pool pH to prevent iodine re-evolution 

following a postulated design basis loss of 

coolant accident. There are new required 

manual operator actions associated with the 

SLC System that are not currently considered 

in the SSES design basis. Operator training 

will be updated to reflect the new manual 

operator actions for the pH control function 

of the of the SLC System as defined in the 

TS Section 3.1.7. These changes are not 

significant because the operators are already 

trained for the operation of the SLC System. 

Procedural changes are mostly limited to the 

timing of SLC initiation and termination. In 

addition, no new hardware changes are 

necessary to use SLC in this new functional 

mode. 

Licensing basis changes are proposed and 

justified for the operation of the CREOASS 

[control room emergency outside air supply 

system] to respond to NRC Generic Letter 

2003–01 and TSTF [Technical Specification 

Task Force] 448. No new hardware changes 

are necessary to implement these changes. 

Since CREOASS will not be operated 

differently as a result of these changes, no 

new failure modes are created by these 

changes. 

Therefore, the proposed license 

amendment will 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 results of the accident analyses revised 

in support of the proposed change are subject 

to the acceptance criteria in 10 CFR 50.67. 

The analyzed events have been carefully 

selected, and the analyses supporting these 

changes have been performed using approved 

methodologies to ensure that analyzed events 

are bounding and safety margin has not been 

reduced. The dose consequences of these 

limiting events are within the acceptance 

criteria presented in 10 CFR 50.67, RG 1.183, 

and SRP 15.0.1. Thus, by meeting the 

applicable regulatory limits for AST, there is 

no significant reduction in a margin of safety. 

Changes to the SLC System to credit use of 

the Standby Liquid Control (SLC) System to 

buffer suppression pool pH to prevent iodine 

re-evolution and the CREOASS to address 

NRC Generic Letter 2003–01 and TSTF–448 

[to] improve the margin of safety. 

New offsite and Control Room atmospheric 

dispersion factors (X/Qs) based on site 

specific meteorological data, calculated in 

accordance with the guidance of RGs 1.145 

and 1.194, utilizes more recent data and 

improved calculational methodologies. 

Therefore, because the proposed changes 

continue to result in dose consequences 

within the applicable regulatory limits, the 

changes are considered to not result in 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: Bryan A. Snapp, 

Esquire, Assoc. General Counsel, PPL 

Services Corporation, 2 North Ninth St., 

GENTW3, Allentown, PA 18101–1179. 

NRC Branch Chief: Richard J. Laufer. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of amendment request: June 7, 

2006. 

Description of amendment request: 

The proposed amendment would delete 

Technical Specification (TS) 3/4.6.4, 

‘‘Combustible Gas Control,’’ TS 3/ 

4.6.4.1, ‘‘Hydrogen Analyzers,’’ and TS 

3/4.6.4.2, ‘‘Electric Hydrogen 

Recombiners.’’ The changes would be 

consistent with NRC-approved TS Task 

Force (TSTF) Standard Technical 

Specification (STS) Change Traveler 

number TSTF–447, Revision 1, 

‘‘Elimination of Hydrogen Recombiners 

and Change to Hydrogen and Oxygen 

Monitors,’’ as part of the Consolidated 

Line Item Improvement Process (CLIIP). 

The NRC staff issued a notice of 

availability of ‘‘Model Application 

Concerning Technical Specification 

Improvement To Eliminate Hydrogen 

Recombiner Requirement, and Relax the 

Hydrogen and Oxygen Monitor 

Requirements for Light Water Reactors 


jlentini on PROD1PC65 with NOTICES
51232 Federal Register / Vol. 71, No. 167 / Tuesday, August 29, 2006 / Notices 

Using the Consolidated Line Item 

Improvement Process (CLIIP)’’, in the 

Federal Register on September 25, 2003 

(68 FR 55416). The notice included a 

model safety evaluation (SE), a model 

no significant hazards consideration 

(NSHC) determination, and a model 

application. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee presented an analysis of NSHC 

by endorsing the model NSHC 

determination for TSTF–447 which was 

published in the Federal Register on 

September 25, 2003 (68 FR 55416) as 

follows: 

Criterion 1—The Proposed Change Does 

Not Involve a Significant Increase in the 

Probability or Consequences of an Accident 

Previously Evaluated. 

The revised 10 CFR 50.44 no longer defines 

a design-basis loss-of-coolant accident 

(LOCA) hydrogen release, and eliminates 

requirements for hydrogen control systems to 

mitigate such a release. The installation of 

hydrogen recombiners and/or vent and purge 

systems required by 10 CFR 50.44(b)(3) was 

intended to address the limited quantity and 

rate of hydrogen generation that was 

postulated from a design-basis LOCA. The 

Commission has found that this hydrogen 

release is not risk-significant because the 

design basis LOCA hydrogen release does not 

contribute to the conditional probability of a 

large release up to approximately 24 hours 

after the onset of core damage. In addition, 

these systems were ineffective at mitigating 

hydrogen releases from risk-significant 

accident sequences that could threaten 

containment integrity. 

With the elimination of the design basis 

LOCA hydrogen release, hydrogen monitors 

are no longer required to mitigate designbasis accidents and, therefore, the hydrogen 

monitors do not meet the definition of a 

safety-related component as defined in 10 

CFR 50.2. RG [Regulatory Guide] 1.97 

Category 1, is intended for key variables that 

most directly indicate the accomplishment of 

a safety function for design-basis accident 

events. The hydrogen monitors no longer 

meet the definition of Category 1 in RG 1.97. 

As part of the rulemaking to revise 10 CFR 

50.44 the Commission found that Category 3, 

as defined in RG 1.97, is an appropriate 

categorization for the hydrogen monitors 

because the monitors are required to 

diagnose the course of beyond design-basis 

accidents. 

The regulatory requirements for the 

hydrogen monitors can be relaxed without 

degrading the plant emergency response. The 

emergency response, in this sense, refers to 

the methodologies used in ascertaining the 

condition of the reactor core, mitigating the 

consequences of an accident, assessing and 

projecting offsite releases of radioactivity, 

and establishing protective action 

recommendations to be communicated to 

offsite authorities. Classification of the 

hydrogen monitors as Category 3, and 

removal of the hydrogen monitors from TS 

will not prevent an accident management 

strategy through the use of the SAMGs 

[Severe Accident Management Guidelines], 

the emergency plan (EP), the emergency 

operating procedures (EOP), and site survey 

monitoring that support modification of 

emergency plan protective action 

recommendations (PARs). 

Therefore, the elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, does not involve a significant 

increase in the probability or the 

consequences of any accident previously 

evaluated. 

Criterion 2—The Proposed Change Does 

Not Create the Possibility of a New or 

Different Kind of Accident From Any 

Previously Evaluated. 

The elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, will not result in any failure mode 

not previously analyzed. The hydrogen 

recombiner and hydrogen monitor equipment 

was intended to mitigate a design-basis 

hydrogen release. The hydrogen recombiner 

and hydrogen monitor equipment are not 

considered accident precursors, nor does 

their existence or elimination have any 

adverse impact on the pre-accident state of 

the reactor core or post accident confinement 

of radionuclides within the containment 

building. 

Therefore, this 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 the 

Margin of Safety. 

The elimination of the hydrogen 

recombiner requirements and relaxation of 

the hydrogen monitor requirements, 

including removal of these requirements 

from TS, in light of existing plant equipment, 

instrumentation, procedures, and programs 

that provide effective mitigation of and 

recovery from reactor accidents, results in a 

neutral impact to the margin of safety. 

The installation of hydrogen recombiners 

and/or vent and purge systems required by 

10 CFR 50.44(b)(3) was intended to address 

the limited quantity and rate of hydrogen 

generation that was postulated from a designbasis LOCA. The Commission has found that 

this hydrogen release is not risk-significant 

because the design-basis LOCA hydrogen 

release does not contribute to the conditional 

probability of a large release up to 

approximately 24 hours after the onset of 

core damage. Category 3 hydrogen monitors 

are adequate to provide rapid assessment of 

current reactor core conditions and the 

direction of degradation while effectively 

responding to the event in order to mitigate 

the consequences of the accident. The intent 

of the requirements established as a result of 

the TMI, Unit 2 accident can be adequately 

met without reliance on safety-related 

hydrogen monitors. 

Therefore, this change does not involve a 

significant reduction in the margin of safety. 

Removal of hydrogen monitoring from TS 

will not result in a significant reduction in 

their functionality, reliability, and 

availability. 

The NRC staff has reviewed the licensee’s 

analysis and, based on this review, it appears 

that the three standards of 10 CFR 50.92(c) 

are satisfied. Therefore, the NRC staff 

proposes to determine that the amendment 

request involves no significant hazards 

consideration. 

Attorney for licensee: Jeffrie J. Keenan, 

Esquire, Nuclear Business Unit—N21, 

P.O. Box 236, Hancocks Bridge, NJ 

08038. 

NRC Acting Branch Chief: Brooke D. 

Poole. 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

of 1954, as amended (the Act), and the 

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

Systems (ADAMS) Public Electronic 

Reading Room on the internet at the 


jlentini on PROD1PC65 with NOTICES
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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@nrc.gov. 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of application for amendment: 

January 31, 2006, as supplemented by 

letter dated July 12, 2006. 

Brief description of amendment: The 

amendment changed the technical 

specifications to address issues related 

to an inconsistency that was 

inadvertently introduced during 

conversion to improved technical 

specifications when ‘‘1 per room’’

replaced ‘‘2’’ as the required channels 

per trip system for the reactor water 

cleanup area ventilation differential 

temperature—high isolation function. 

Date of issuance: August 7, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment No.: 173. 

Facility Operating License No. NPF–

43: Amendment revised the Technical 

Specifications, Surveillance 

Requirements, and License. 

Date of initial notice in Federal 

Register: March 14, 2006 (71 FR 

13171). The July 12, 2006, supplement 

contained clarifying information and 

did not change the NRC staff’s initial 

proposed finding of no significant 

hazards consideration. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated August 7, 2006. 

No significant hazards consideration 

comments received: No. 

Dominion Nuclear Connecticut, Inc., et 

al., Docket No. 50–423, Millstone Power 

Station, Unit No. 3, New London 

County, Connecticut 

Date of application for amendment: 

September 13, 2005. 

Brief description of amendment: The 

amendment revised the Millstone Power 

Station, Unit No. 3 Technical 

Specification (TSs) temperature 

requirement for the reactivity control 

system rod drop time test. 

Date of issuance: August 15, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment No.: 231. 

Facility Operating License No. NPF–

49: The amendment revised the TSs. 

Date of initial notice in Federal 

Register: October 25, 2005 (70 FR 

61656). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated August 15, 

2006. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of application for amendment: 

May 24, 2005, revised by letter dated 

May 2, 2006. 

Brief description of amendment: The 

Technical Specification amendment 

deleted the requirements for NRC 

approval of the engineering evaluation 

justifying continued reactor operation 

with safety relief valve (SRV) discharge 

pipe temperature exceeding the limit. 

Date of issuance: August 4, 2006. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 90 days. 

Amendment No.: 222. 

Facility Operating License No. DPR–

35: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 16, 2005 (70 FR 

48205). 

The licensee originally requested for 

deletion of TS 3.6.D.3 in their submittal 

dated May 24, 2005. The NRC staff 

found this unacceptable. Therefore, the 

licensee revised the original application 

by letter dated May 2, 2006, reducing 

the scope of the application as originally 

noticed. Hence, there is no change to the 

NRC staff’s original proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated August 4, 2006. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: July 21, 

2005, as supplemented on February 15, 

May 3, and June 2, 2006. 

Brief description of amendment: The 

amendment modified the Waterford 3 

Technical Specification (TS) to revise 

the existing steam generator (SG) tube 

surveillance program to be consistent 

with TS Task Force (TSTF) Change 

TSTF–449, ‘‘Steam Generator Tube 

Integrity,’’ Revision 4, and the model 

safety evaluation prepared by the NRC 

and published in the Federal Register 

notice on March 2, 2005 (70 FR 10298). 

In this regard, the scope of the 

application includes changes to the 

definition of leakage, changes to the 

primary-to-secondary leakage 

requirements, changes to the SG tube 

surveillance program, changes to the SG 

reporting requirements, and associated 

changes to the TS Bases. 

Date of issuance: July 31, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 90 

days from the date of issuance. 

Amendment No.: 204. 

Facility Operating License No. NPF–

38: The amendment revised the 

Technical Specifications and the 

Facility Operating License. 

Date of initial notice in Federal 

Register: October 25, 2005 (70 FR 

61659). The February 15, May 3, and 

June 2, 2006, supplemental letters 

contained additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the NRC 

staff’s original proposed finding of no 

significant hazards consideration. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 31, 2006. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: October 

25, 2005. 

Brief description of amendment: The 

amendment modified the Waterford 3 

Technical Specification 3.1.1.4, 

‘‘Minimum Temperature for Criticality,’’

to raise the minimum temperature for 

criticality from the current value of ≥520 

°F to ≥533 °F. Changes were also 

proposed to the associated Action 

statement to reflect the increase in 

temperature and to replace the current 

statement in Surveillance Requirement 

4.1.1.4 with wording consistent with 

NUREG–1432, ‘‘Standard Technical 

Specifications—Combustion 

Engineering Plants.’’

Date of issuance: July 31, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 30 

days from the date of issuance. 

Amendment No.: 205. 

Facility Operating License No. NPF–

38: The amendment revised the 

Technical Specifications. 

Date of initial notice in Federal 

Register: December 6, 2005 (70 FR 

72672). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated July 31, 2006. 

No significant hazards consideration 

comments received: No. 


jlentini on PROD1PC65 with NOTICES
51234 Federal Register / Vol. 71, No. 167 / Tuesday, August 29, 2006 / Notices 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–346, 

Davis-Besse Nuclear Power Station, Unit 

1, Ottawa County, Ohio 

Date of application for amendment: 

May 5, 2004, as supplemented by letters 

dated January 17, October 10, and 

November 2, 2005 and May 30, 2006. 

Brief description of amendment: This 

amendment revised the technical 

specifications (TSs) for instrumentation 

setpoints, allowable values, and 

calibration requirements based on 

updated calculations and reviews, and 

add a definition of ‘‘annual’’ frequency 

for use in the TSs. 

Date of issuance: August 9, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days. 

Amendment No.: 275. 

Facility Operating License No. NPF–3: 

Amendment revised the Technical 

Specifications and License. 

Date of initial notice in Federal 

Register: June 8, 2004 (69 FR 32074). 

The January 17, October 10, and 

November 2, 2005 and May 30, 2006, 

supplements contained clarifying 

information and did not change the NRC 

staff’s initial proposed finding of no 

significant hazards consideration. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated August 9, 2006. 

No significant hazards consideration 

comments received: No. 

Southern California Edison Company, et 

al., Docket Nos. 50–361 and 50–362, 

San Onofre Nuclear Generating Station, 

Units 2 and 3, San Diego County, 

California 

Date of application for amendments: 

January 28, 2005, as supplemented by 

letter dated January 12, 2006. 

Brief description of amendments: The 

amendment revised Technical 

Specifications (Tss) 1.1, ‘‘Definitions,’’

3.4, ‘‘Reactor Coolant System [RCS],’’

and 5.7, ‘‘Reporting Requirements,’’ to 

relocate the RCS pressure-temperature 

curves and limits from the TSs to a 

licensee-controlled document identified 

as the Pressure and Temperature Limit 

Report. 

Date of issuance: July 13, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment Nos.: Unit 2–203; Unit 

3–195. 

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments 

revised the Technical Specifications. 

Date of initial notice in Federal 

Register: March 1, 2005 (70 FR 9996). 

The January 12, 2006, supplemental 

letter provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated July 13, 2006. 

No significant hazards consideration 

comments received: No. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–424 and 50–425, 

Vogtle Electric Generating Plant, Units 1 

and 2, Burke County, Georgia 

Date of application for amendments: 

September 19, 2005, as supplemented 

by letter dated June 9, 2006. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications (TSs) Limiting 

Conditions for Operation 3.3.1, ‘‘Reactor 

Trip System Instrumentation,’’ and TS 

Surveillance Requirement (SR) 3.2.4.2, 

‘‘Quadrant Power Tilt Ratio (QPTR).’’

The amendments revise TS 3.3.1, 

Condition D and the note in SR 3.2.4.2 

to clarify when a flux map for QTPR is 

required. 

Date of issuance: August 15, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment Nos.: 143 and 123. 

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised 

the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: March 14, 2005 (71 FR 

13179). The supplement dated June 9, 

2006, provided clarifying information 

that did not change the scope of the 

September 19, 2005, application nor the 

initial proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated August 15, 

2006. 

No significant hazards consideration 

comments received: No. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Unit Nos. 

1 and 2, Somervell County, Texas 

Date of amendment request: August 

10, 2005. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications (TSs) to relocate diesel 

fuel oil testing methods from TS 5.5.13 

to a licensee-controlled document, 

provided clarifications, and corrected 

the format and typographical errors. 

Date of issuance: July 28, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: 127 and 127. 

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments 

revised the Technical Specifications and 

the Facility Operating Licenses. 

Date of initial notice in Federal 

Register: November 8, 2005 (70 FR 

67754). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated July 28, 2006. 

No significant hazards consideration 

comments received: No. 

Virginia Electric and Power Company, et 

al., Docket Nos. 50–280 and 50–281, 

Surry Power Station, Units 1 and 2, 

Surry County, Virginia 

Date of application for amendments: 

September 13, 2005, as supplemented 

on April 7 and May 23, 2006. 

Brief Description of amendments: 

These amendments revised Technical 

Specification 5.1, ‘‘Site,’’ to redefine the 

exclusion area boundary as the site 

boundary. 

Date of issuance: August 10, 2006. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 249, 248. 

Renewed Facility Operating License 

Nos. DPR–32 and DPR–37: Amendments 

changed the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: January 3, 2006 (71 FR 156). 

The April 7 and May 23, 2006, 

supplements contained clarifying 

information only and did not change the 

initial proposed no significant hazards 

consideration determination or expand 

the scope of the initial application. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated August 10, 

2006. 

No significant hazards consideration 

comments received: No. 

Notice of Issuance of Amendments to 

Facility Operating Licenses and Final 

Determination of No Significant 

Hazards Consideration and 

Opportunity for a Hearing (Exigent 

Public Announcement or Emergency 

Circumstances) 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application for the 

amendment complies with the 


jlentini on PROD1PC65 with NOTICES
Federal Register / Vol. 71, No. 167 / Tuesday, August 29, 2006 / Notices 51235 

standards and requirements of the 

Atomic Energy Act of 1954, as amended 

(the Act), and the Commission’s rules 

and regulations. The Commission has 

made appropriate findings as required 

by the Act and the Commission’s rules 

and regulations in 10 CFR Chapter I, 

which are set forth in the license 

amendment. 

Because of exigent or emergency 

circumstances associated with the date 

the amendment was needed, there was 

not time for the Commission to publish, 

for public comment before issuance, its 

usual Notice of Consideration of 

Issuance of Amendment, Proposed No 

Significant Hazards Consideration 

Determination, and Opportunity for a 

Hearing. 

For exigent circumstances, the 

Commission has either issued a Federal 

Register notice providing opportunity 

for public comment or has used local 

media to provide notice to the public in 

the area surrounding a licensee’s facility 

of the licensee’s application and of the 

Commission’s proposed determination 

of no significant hazards consideration. 

The Commission has provided a 

reasonable opportunity for the public to 

comment, using its best efforts to make 

available to the public means of 

communication for the public to 

respond quickly, and in the case of 

telephone comments, the comments 

have been recorded or transcribed as 

appropriate and the licensee has been 

informed of the public comments. 

In circumstances where failure to act 

in a timely way would have resulted, for 

example, in derating or shutdown of a 

nuclear power plant or in prevention of 

either resumption of operation or of 

increase in power output up to the 

plant’s licensed power level, the 

Commission may not have had an 

opportunity to provide for public 

comment on its no significant hazards 

consideration determination. In such 

case, the license amendment has been 

issued without opportunity for 

comment. If there has been some time 

for public comment but less than 30 

days, the Commission may provide an 

opportunity for public comment. If 

comments have been requested, it is so 

stated. In either event, the State has 

been consulted by telephone whenever 

possible. 

Under its regulations, the Commission 

may issue and make an amendment 

immediately effective, notwithstanding 

the pendency before it of a request for 

a hearing from any person, in advance 

of the holding and completion of any 

required hearing, where it has 

determined that no significant hazards 

consideration is involved. 

The Commission has applied the 

standards of 10 CFR 50.92 and has made 

a final determination that the 

amendment involves no significant 

hazards consideration. The basis for this 

determination is contained in the 

documents related to this action. 

Accordingly, the amendments have 

been issued and made effective as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.12(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the application for 

amendment, (2) the amendment to 

Facility Operating License, and (3) the 

Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment, as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

System’s (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@nrc.gov. 

The Commission is also offering an 

opportunity for a hearing with respect to 

the issuance of the amendment. Within 

60 days after the date of publication of 

this notice, the licensee may file a 

request for a hearing with respect to 

issuance of the amendment to the 

subject facility operating license and 

any person whose interest may be 

affected by this proceeding and who 

wishes to participate as a party in the 

proceeding must file a written request 

for a hearing and a petition for leave to 

intervene. 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 persons should 

consult a current copy of 10 CFR 2.309, 

which is available at the Commission’s 

PDR, located at One White Flint North, 

Public File Area 01F21, 11555 Rockville 

Pike (first floor), Rockville, Maryland, 

and electronically on the Internet at the 

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

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

are problems in accessing the document, 

contact the PDR Reference staff at 1 

(800) 397–4209, (301) 415–4737, or by email to pdr@nrc.gov. If a request for a 

hearing or petition for leave to intervene 

is filed by the above date, the 

Commission or a presiding officer 

designated by the Commission or by the 

Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the petitioner/ 

requestor 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 petitioner/requestor 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 petitioner intends to 

rely in proving the contention at the 

hearing. The petitioner must also 

provide references to those specific 

sources and documents of which the 

petitioner is aware and on which the 

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 


jlentini on PROD1PC65 with NOTICES
51236 Federal Register / Vol. 71, No. 167 / Tuesday, August 29, 2006 / Notices 

1To the extent that the applications contain 

attachments and supporting documents that are not 

publicly available because they are asserted to 

contain safeguards or proprietary information, 

petitioners desiring access to this information 

should contact the applicant or applicant’s counsel 

and discuss the need for a protective order. 

material issue of law or fact.1

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 petitioner to relief. A 

petitioner/requestor who fails to satisfy 

these requirements with respect to at 

least one contention will not be 

permitted to participate as a party. 

Each contention shall be given a 

separate numeric or alpha designation 

within one of the following groups: 

1. Technical—primarily concerns/ 

issues relating to technical and/or 

health and safety matters discussed or 

referenced in the applications. 

2. Environmental—primarily 

concerns/issues relating to matters 

discussed or referenced in the 

environmental analysis for the 

applications. 

3. Miscellaneous—does not fall into 

one of the categories outlined above. 

As specified in 10 CFR 2.309, if two 

or more petitioners/requestors seek to 

co-sponsor a contention, the petitioners/ 

requestors shall jointly designate a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. If a petitioner/requestor 

seeks to adopt the contention of another 

sponsoring petitioner/requestor, the 

petitioner/requestor who seeks to adopt 

the contention must either agree that the 

sponsoring petitioner/requestor shall act 

as the representative with respect to that 

contention, or jointly designate with the 

sponsoring petitioner/requestor a 

representative who shall have the 

authority to act for the petitioners/ 

requestors with respect to that 

contention. 

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. Since the Commission has 

made a final determination that the 

amendment involves no significant 

hazards consideration, if a hearing is 

requested, it will not stay the 

effectiveness of the amendment. Any 

hearing held would take place while the 

amendment is in effect. 

A request for a hearing or a petition 

for leave to intervene must be filed 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; (2) courier, express 

mail, and expedited delivery services: 

Office of the Secretary, Sixteenth Floor, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland 20852, 

Attention: Rulemaking and 

Adjudications Staff; (3) E-mail 

addressed to the Office of the Secretary, 

U.S. Nuclear Regulatory Commission, 

HearingDocket@nrc.gov; or (4) facsimile 

transmission addressed to the Office of 

the Secretary, U.S. Nuclear Regulatory 

Commission, Washington, DC, 

Attention: Rulemakings and 

Adjudications Staff at (301) 415–1101, 

verification number is (301) 415–1966. 

A copy of the request for hearing and 

petition for leave to intervene should 

also be sent to the Office of the General 

Counsel, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555–

0001, and it is requested that copies be 

transmitted either by means of facsimile 

transmission to (301) 415–3725 or by email to OGCMailCenter@nrc.gov. A copy 

of the request for hearing and petition 

for leave to intervene should also be 

sent to the attorney for the licensee. 

Nontimely requests and/or petitions 

and contentions will not be entertained 

absent a determination by the 

Commission or the presiding officer or 

the Atomic Safety and Licensing Board 

that the petition, request and/or the 

contentions should be granted based on 

a balancing of the factors specified in 10 

CFR 2.309(a)(1)(i)–(viii). 

Entergy Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit No. 2, 

Westchester County, New York 

Date of amendment request: July 26, 

2006. 

Description of amendment request: 

The amendment revised Function 6 

[Containment Water Level (Containment 

Sump)] of Table 3.3.3–1 (‘‘Post Accident 

Monitoring Instrumentation’’), 

referenced in the Technical 

Specification (TS) Limiting Condition 

for Operation (LCO) 3.3.3, ‘‘Post 

Accident Monitoring Instrumentation.’’

The revision changed Function 6 to 

specify 2 required channels for the 

Containment Sump water level 

instrumentation instead of 3 channels. 

Date of issuance: July 28, 2006. 

Effective date: As of its date of 

issuance, and shall be implemented 

prior to the expiration of the current 7- 

day allowed outage time for inoperable 

containment sump water level channels, 

which was entered on July 24, 2006. 

Amendment No.: 249. 

Facility Operating License No. DPR–

26: The amendment revised the TS and 

License. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): No. The 

Commission’s related evaluation of the 

amendment, finding of emergency 

circumstances, state consultation, and 

final NSHC determination are contained 

in a Safety Evaluation dated July 28, 

2006. 

Attorney for licensee: Mr. John Fulton, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc., 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Branch Chief: Richard J. Laufer. 

Dated at Rockville, Maryland, this 21st day 

of August 2006. 

For the Nuclear Regulatory Commission. 

Cornelius F. Holden, 

Deputy Director, Division of Operating 

Reactor Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 06–7137 Filed 8–28–06; 8:45 am] 

BILLING CODE 7590–01–P 

POSTAL RATE COMMISSION 

Sunshine Act Meetings 

NAME OF AGENCY: Postal Rate 

Commission. 

TIME AND DATE: Monday, August 28, 

2006, at 3 p.m. 

PLACE: Commission conference room, 

901 New York Avenue, NW., Suite 200, 

Washington, DC 20268–0001. 

STATUS: Open. 

Matters to be Considered: Consideration 

of fiscal year 2007 budget and election 

of vice chairman. 

FOR FURTHER INFORMATION CONTACT: 

Stephen L. Sharfman, General Counsel, 

at 202–789–6820. 

Dated: August 24, 2006. 

Steven W. Williams, 

Secretary. 

[FR Doc. 06–7234 Filed 8–24–06; 4:37 pm] 

BILLING CODE 7710–FW–M 

RAILROAD RETIREMENT BOARD 

Proposed Collection; Comment 

Request 

SUMMARY: In accordance with the 

requirement of Section 3506(c)(2)(A) of 

the Paperwork Reduction Act of 1995 

which provides opportunity for public 

comment on new or revised data 

collections, the Railroad Retirement 

Board (RRB) will publish periodic 

summaries of proposed data collections. 

Comments are invited on: (a) Whether 

the proposed information collection is 


jlentini on PROD1PC65 with NOTICES