Federal Register / Vol. 72, No. 48 / Tuesday, March 13, 2007 / Notices 11383 

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 

requestors/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/requestor 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 

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 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 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

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(c)(1)(i)–(viii). 

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 David T. Conley, Associate 

General Counsel II—Legal Department, 

Progress Energy Service Company, LLC, 

Post Office Box 1551, Raleigh, North 

Carolina 27602, attorney for the 

licensee. 

For further details with respect to this 

action, see the application for 

amendment dated February 8, 2007, 

which is available for public inspection 

at the Commission’s PDR, located at 

One White Flint North, File Public Area 

O1 F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. Publicly 

available records will be accessible from 

the Agencywide Documents Access and 

Management System’s (ADAMS) Public 

Electronic Reading Room on the Internet 

at the NRC Web site, http:// 

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

Persons who do not have access to 

ADAMS or who encounter problems in 

accessing the documents located in 

ADAMS, should contact the NRC PDR 

Reference staff by telephone at 1–800–

397–4209, 301–415–4737, or by e-mail 

to pdr@nrc.gov. 

Dated at Rockville, Maryland, this 7th day 

of March 2007. 

For the Nuclear Regulatory Commission. 

Stewart N. Bailey, 

Senior Project Manager, Plant Licensing 

Branch II–2, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E7–4517 Filed 3–12–07; 8:45 am] 

BILLING CODE 7590–01–P 

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 

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 February 

15, 2007 through March 1, 2007. The 

last biweekly notice was published on 

February 27, 2007 (72 FR 8800). 


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11384 Federal Register / Vol. 72, No. 48 / Tuesday, March 13, 2007 / Notices 

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, Rulemaking, 

Directives and Editing 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 

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 


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Federal Register / Vol. 72, No. 48 / Tuesday, March 13, 2007 / Notices 11385 

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. 

Calvert Cliffs Nuclear Power Plant, Inc., 

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

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of amendment request: February 

1, 2007. 

Description of amendment request: 

The proposed license amendment 

would revise Surveillance Requirement 

(SR) 3.5.2.8 in Technical Specification 

3.5.2, ‘‘ECCS [Emergency Core Cooling 

System]—Operating,’’ to reflect the 

replacement of the containment 

recirculation sump suction inlet trash 

racks and screens with strainers, in 

response to Nuclear Regulatory 

Commission (NRC) Generic Letter 2004–

02, ‘‘Potential Impact of Debris Blockage 

on Emergency Recirculation during 

Design Basis Accidents at PressurizedWater Reactors.’’ The proposed license 

amendment would replace ‘‘trash racks 

and screens’’ with ‘‘strainers’’ in SR 

3.5.2.8. 

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? 

The consequences of accidents evaluated 

in the Updated Final Safety Analysis Report 

[UFSAR] that could be affected by the 

proposed change are those involving the 

pressurization of Containment and associated 

flooding of the Containment and 

recirculation of this fluid within the 

Emergency Core Cooling System (ECCS) or 

the Containment Spray System (CSS) (e.g., 

loss-of-coolant accidents [LOCAs]). The 

proposed change does not impact the 

initiation or probability of occurrence of any 

accident. Although the configurations of the 

existing containment recirculation sump 

trash racks and screen and the replacement 

sump strainer cassettes are different, they 

serve the same fundamental purpose of 

passively removing debris from the sump’s 

suction supply of the supported system 

pumps. Removal of trash racks does not 

impact the adequacy of the pump net 

positive suction head assumed in the safety 

analysis. Likewise, the change does not 

reduce the reliability of any supported 

systems or introduce any new system 

interactions. The greatly increased surface 

area of the new strainer is designed to reduce 

head loss and reduce the approach velocity 

at the strainer face significantly, decreasing 

the risk of impact from large debris entrained 

in the sump flow stream. 

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? 

The containment recirculation sump 

strainers are a passive system used for 

accident mitigation. As such, they cannot be 

accident initiators. Therefore, there is no 

possibility that this change could create any 

new or different kind of accident. No new 

accident scenarios, transient precursors, or 

limiting single failures are introduced as a 

result of the proposed change. There will be 

no adverse effect or challenges imposed on 

any safety-related system as a result of the 

change. Therefore, the possibility of a new or 

different [kind] of accident is not created. 

There are no changes which would cause 

the malfunction of safety-related equipment, 

assumed to be OPERABLE in the accident 

analyses, as a result of the proposed 

Technical Specification change. No new 

equipment performance burdens are 

imposed. The possibility of a malfunction of 

safety-related equipment with a different 

result is not created. 

Therefore, the proposed change does not 

create the possibility of a new or different 

[kind of] accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

The proposed change does not affect the 

acceptance criteria for any analyzed event 

nor is there a change to any safety analysis 

limit. There will be no effect on the manner 

in which safety limits, limiting safety system 

settings, or limiting conditions for operation 

are determined nor will there be any effect 

on those plant systems necessary to assure 

the accomplishment of protection functions. 

The proposed change does not adversely 

affect the fuel, fuel cladding, Reactor Coolant 

System, or containment integrity. The 

radiological dose consequence acceptance 

criteria listed in the Updated Final Safety 

Analysis Report will continue to be 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 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Carey Fleming, 

Esquire, Senior Counsel—Nuclear 

Generation, Constellation Generation 

Group, LLC, 750 East Pratt Street, 17th 

floor, Baltimore, MD 21202. 

NRC Acting Branch Chief: John P. 

Boska. 

Carolina Power & Light Company, 

Docket Nos. 50–325 and 5–324 

Brunswick Steam Electric Plant, Units 1 

and 2, Brunswick County, North 

Carolina 

Date of amendments request: 

December 21, 2006. 

Description of amendment request: 

The proposed amendment would 


cprice-sewell on PROD1PC66 with NOTICES
11386 Federal Register / Vol. 72, No. 48 / Tuesday, March 13, 2007 / Notices 

modify technical specification (TS) 

requirements of TS 3.4.1, ‘‘Recirculation 

Loops Operating,’’ to require the 

recirculation loops be operated with 

matched flows versus recirculation 

pump speeds as currently required. This 

change affects the Limiting Condition 

for Operation (LCO) requirements and 

Surveillance Requirements (SRs) of TS 

3.4.1. 

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. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment implements 

more conservative requirements associated 

with recirculation loop operation. 

Specifically, the LCO requirements of TS 

3.4.1 and SR 3.4.1.1 are being revised to 

directly monitor recirculation loop jet pump 

flows versus recirculation pump speed, 

eliminating potential non-conservatism 

associated with relating recirculation loop jet 

pump flow to recirculation pump speed. 

These requirements assure that the mismatch 

between recirculation loop jet pump flows 

are bounded by the existing design bases 

analyses. As a result, the proposed change 

ensures that the consequences of a design 

bases LOCA [loss-of-coolant accident] remain 

within the existing evaluation. 

The proposed change does not involve a 

physical change to the Reactor Recirculation 

system, nor does it alter the assumptions of 

the accident analyses. Therefore the 

probability of an accident previously 

evaluated is not affected. 

Based on the above, the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve a 

physical change to the Reactor Recirculation 

system, nor does it alter the assumptions of 

the accident analyses. 

The implementation of more conservative 

requirements associated with recirculation 

loop operation does not introduce any new 

failure modes. Therefore, the proposed 

change does not create the possibility of a 

new or different kind of accident from any 

accident previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed amendment implements 

more conservative requirements associated 

with recirculation loop operation. These 

requirements ensure that the Reactor 

Recirculation system is operated consistent 

with the initial conditions of the existing 

design bases analyses. Since the design bases 

analyses assumptions are unchanged, the 

proposed change does not involve 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: 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. 

Dominion Energy Kewaunee, Inc. Docket 

No. 50–305, Kewaunee Power Station, 

Kewaunee County, Wisconsin 

Date of amendment request: 

December 15, 2006. 

Description of amendment request: 

The amendment would incorporate 

changes to the Technical Specifications 

(TS) associated with previously 

approved industry initiatives. The first 

change would relocate the Safety Limit 

Violation specifications from the 

administrative controls TS section to the 

safety limit TS sections as approved by 

TSTF–05–A, ‘‘Deletion of Safety Limit 

Violation Requirements.’’ The second 

change would incorporate generic 

position titles, as approved by TSTF–

65–A, ‘‘Use of Generic Titles for Utility 

Positions,’’ and incorporates changes 

approved by NRC Administrative Letter 

(AL) 95–06, ‘‘Relocation of Technical 

Specification Administrative Controls 

Related to Quality Assurance.’’

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The proposed amendment consists of 

changes to and relocation of administrative 

TS requirements that were previously 

generically approved by the NRC. The 

proposed amendment would not change any 

of the previously evaluated accidents in the 

updated safety analysis report (USAR). The 

administrative controls that are affected by 

the proposed amendment do not have any 

function related to preventing or mitigating 

any of these previously evaluated accidents. 

The proposed amendment does not affect any 

systems, structures, or components (SSCs) 

that have the function of preventing or 

mitigating any of these previously evaluated 

accidents. The proposed amendment does 

not increase the likelihood of the 

malfunction of an SSC, thus the potential 

impact on analyzed accidents need not be 

considered. 

Because the proposed amendment is a 

relocation of administrative requirements 

that are not associated with preventing or 

mitigating the consequences of any 

previously evaluated accidents, there is no 

affect on the probability or consequences of 

an accident previously evaluated. 

Therefore, the proposed amendment does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

No. The proposed amendment consists of 

changes to and relocation of administrative 

TS requirements previously generically 

approved by the NRC. This amendment will 

not change the design function of any SSC or 

the manner that any SSC is operated. Because 

this amendment does not change the design 

function or operation of any SSC, the 

amendment would not create the possibility 

of a new or different kind of accident due to 

credible new failure mechanisms, 

malfunctions, or accident initiators not 

considered in the design and licensing bases. 

Therefore, the proposed amendment does 

not create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

No. The proposed amendment consists of 

changes to and relocation of administrative 

TS requirements previously generically 

approved by the NRC. The amendment does 

not alter any design basis safety limit and no 

safety margins are affected. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Bradley D. 

Jackson, Esq., Foley and Lardner, P.O. 

Box 1497, Madison, WI 53701–1497. 

NRC Acting Branch Chief: P. Milano. 

Duke Power Company LLC, et al., 

Docket No. 50–413, Catawba Nuclear 

Station, Unit 1 (Catawba), York County, 

South Carolina 

Date of amendment request: 

November 22, 2006. 

Description of amendment request: 

The amendment would revise the 

Catawba Unit 1 Facility Operating 

License (FOL) to provide for an 

extension of the time limit to complete 

the required modification to the 

Emergency Core Cooling System (ECCS) 

sump. 


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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 license amendment does 

not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

The proposed license amendment 

delineates a new Unit 1 FOL condition to 

implement a completion date associated with 

the ECCS sump strainer modification. The 

proposed license amendment is 

administrative in nature and is being 

submitted to fulfill a commitment made in 

previous Duke licensing correspondence. 

Therefore, the proposed license amendment 

has no effect upon either the probability or 

consequences of an accident previously 

evaluated. 

2. The proposed license amendment does 

not create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

As stated above, the proposed license 

amendment is administrative in nature and 

does not change the manner in which Unit 

1 is designed or operated. Therefore, the 

proposed license amendment cannot create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. The proposed license amendment does 

not involve a significant reduction in a 

margin of safety. 

Margin of safety is related to the 

confidence in the ability of the fission 

product barriers to perform their intended 

functions. These barriers include the fuel 

cladding, the reactor coolant system, and the 

containment. The performance of these 

barriers will not be affected by the addition 

of the proposed FOL condition. Being 

administrative in nature, the proposed 

license amendment therefore does not 

involve a significant reduction in any safety 

margin. 

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: Ms. Lisa F. 

Vaughn, Associate General Counsel and 

Managing Attorney, Duke Energy 

Carolinas, LLC, 526 South Church 

Street, EC07H, Charlotte, NC 28202. 

NRC Branch Chief: Evangelos C. 

Marinos. 

Duke Power Company LLC, Docket Nos. 

50–369 and 50–370, McGuire Nuclear 

Station, Units 1 and 2, Mecklenburg 

County, North Carolina 

Date of amendment request: April 11, 

2006. 

Description of amendment request: 

The proposed amendments would 

revise the Technical Specifications 

(TSs) related to the organizational 

description in TS 5.2.1 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided it’s analysis of the 

issue of no significant hazards 

consideration. The NRC staff has 

reviewed the licensee’s analysis against 

the standards of 10 CFR 50.92(c). The 

NRC staff’s review 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 revises an 

organizational description in TS 5.2.1 to 

reflect the change of the title of the Vice 

President Nuclear Generation. The 

change is solely administrative in nature 

and has no impact on any accident 

probabilities or consequences. The 

change does not affect structures or 

components in the plant. The change 

has no affect on any accident previously 

evaluated. Therefore the proposed 

change does not involve a significant 

increase in the probability or 

consequences of an accident previously 

evaluated. 

Criterion 2—The Proposed Change Does 

Not Create the Possibility of a New or 

Different Kind of Accident From any 

Accident Previously Evaluated 

There are no new accident causal 

mechanisms created as a result of this 

proposed change. No changes are being 

made to the plant that will introduce 

any new accident causal mechanisms. 

The change is solely administrative in 

nature and does not impact any plant 

systems that are accident initiators. 

Therefore, no new accidents or a 

different accident than previously 

evaluated is being created. 

Criterion 3—The Proposed Change Does 

Not Involve a Significant Reduction in 

a Margin of Safety. 

Margin of safety is related to 

confidence in the ability of the fission 

product barriers to perform their design 

functions during and following an 

accident situation. The proposed change 

is solely administrative in nature and 

does not affect the performance of the 

barriers. Consequently, no safety 

margins will be impacted. Therefore, the 

proposed change does not involve a 

significant reduction in a margin of 

safety. 

Based on this review, it appears that 

the three standards of 10 CFR 50.92(c) 

are satisfied, therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ms. Lisa F. 

Vaughn, Duke Power Company LLC, 

422 South Church Street, Charlotte, 

North Carolina 28201–1006. 

NRC Branch Chief: Evangelos C. 

Marinos. 

Exelon Generation Company, LLC 

(EGC), Docket Nos. STN 50–454 and 

STN 50–455, Byron Station, Unit Nos. 1 

and 2, Ogle County, Illinois 

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and 

2, Will County, Illinois. 

Docket Nos. 50–237 and 50–249, 

Dresden Nuclear Power Station, Units 2 

and 3, Grundy County, Illinois. 

Docket Nos. 50–373 and 50–374, 

LaSalle County Station, Units 1 and 2, 

LaSalle County, Illinois. 

Docket Nos. 50–277 and 50–278, 

Peach Bottom Atomic Power Station, 

Units 2 and 3, York and Lancaster 

Counties, Pennsylvania. 

Docket Nos. 50–254 and 50–265, 

Quad Cities Nuclear Power Station, 

Units 1 and 2, Rock Island County, 

Illinois. 

Date of amendment request: 

December 15, 2006. 

Description of amendment request: 

The proposed amendment would 

modify the technical specifications 

(TSs) by replacing the term ‘‘plantspecific titles’’ with ‘‘generic titles’’ in 

TS Section 5.2.1.a, ensuring the TS 

description is consistent with the EGC 

Quality Assurance Topical Report 

(QATR). The proposed amendment will 

also revise the Peach Bottom TS Section 

5.2.1.a, to replace the reference to the 

Updated Final Safety Analysis Report 

with reference to the EGC QATR. This 

will align the Peach Bottom TS wording 

with the rest of the EGC fleet. 

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 is a word 

replacement in TS 5.2.1, ‘‘Onsite and Offsite 

Organizations.’’ The proposed change 

involves no changes to plant systems or 

accident analyses. The proposed change is 

administrative in nature and, as such, does 

not affect initiators of analyzed events or 

assumed mitigation of accidents or 

transients. 

Therefore, the proposed change does not 

involve any increase in the probability or 


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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 would require 

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 a physical alteration 

of the plant, add any new equipment, or 

allow any existing equipment to be operated 

in a manner different from the present 

method of operation. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change is administrative in 

nature and has no impact on equipment 

design or method of operation. There are no 

changes being made to safety limits or safety 

system allowable values that would 

adversely affect plant safety as a result of the 

proposed change. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, 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, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Michael L. 

Marshall, Jr. 

Exelon Generation Company, LLC (EGC) 

Docket Nos. 50–254 and 50–265, Quad 

Cities Nuclear Power Station, Unit 1, 

Rock Island County, Illinois 

Date of amendment request: January 

16, 2007. 

Description of amendment request: 

The proposed amendment revises the 

values of the safety limit minimum 

critical power ratio (SLMCPR) in the 

Quad Cities Nuclear Power Station 

(QCNPS), Unit 1, Technical 

Specification (TS) Section 2.1.1, 

‘‘Reactor Core SLs [Safety Limits].’’

Specifically, the proposed change 

would require that for QCNPS, Unit 1, 

the minimum critical power ratio shall 

be greater than 1.11 for two 

recirculation loop operation, or greater 

than 1.13 for single recirculation loop 

operation. This change is needed to 

support the next cycle of operation for 

QCNPS, Unit 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. 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-approved methods to 

ensure that fuel performance during normal, 

transient, and accident conditions is 

acceptable. The proposed change 

conservatively establishes the SLMCPR for 

QCNPS, Unit 1, 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 QCNPS, Unit 1, 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. 

Based upon the above, EGC concludes that 

the proposed amendment presents no 

significant hazards consideration under the 

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

accordingly, a finding of no significant 

hazards consideration is justified. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Michael L. 

Marshall, Jr. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–440, 

Perry Nuclear Power Plant, Unit 1, Lake 

County, Ohio 

Date of amendment request: December 

29, 2006. 

Description of amendment request: 

The proposed amendment revises 

Technical Specification (TS) 3.1.8, 

‘‘Scram Discharge Volume (SDV) Vent 

and Drain Valves,’’ to allow a vent or 

drain line with one inoperable valve to 

be isolated instead of requiring the valve 

to be restored to operable status within 

7 days. 

The NRC staff issued a notice of 

opportunity for comment in the Federal 

Register on February 24, 2003 (68 FR 

8637), on possible amendments 

concerning the consolidated line item 

implement process (CLIIP), including a 

model safety evaluation and a model no 

significant hazards consideration 


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(NSHC) determination. The NRC staff 

subsequently issued a notice of 

availability of the models for referencing 

in license amendment applications in 

the Federal Register on April 15, 2003 

(68 FR 18294), as part of the CLIIP. In 

its application dated December 29, 

2006, the licensee affirmed the 

applicability of the following 

determination. 

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: 

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

A change is proposed to allow the affected 

SDV vent and drain line to be isolated when 

there are one or more SDV vent or drain lines 

with one valve inoperable instead or 

requiring the valve to be restored to operable 

status within 7 days. With one SDV vent or 

drain valve inoperable in one or more lines, 

the isolation function would be maintained 

since the redundant valve in the affected line 

would perform its safety function of isolating 

the SDV. Following the completion of the 

required action, the isolation function is 

fulfilled since the associated line is isolated. 

The ability to vent and drain the SDVs is 

maintained and controlled through 

administrative controls. This requirement 

assures the reactor protection system is not 

adversely affected by the inoperable valves. 

With the safety functions of the valves being 

maintained, the probability or consequences 

of an accident previously evaluated are not 

significantly increased. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. Thus, this change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. The proposed change does not involve 

a significant reduction in the margin of 

safety. 

The proposed change ensures that the 

safety functions of the SDV vent and drain 

valves are fulfilled. The isolation function is 

maintained by redundant valves and by the 

required action to isolate the affected line. 

The ability to vent and drain the SDVs is 

maintained through administrative controls. 

In addition, the reactor protection system 

will prevent filling of an SDV to the point 

that it has insufficient volume to accept a full 

scram. Maintaining the safety functions 

related to isolation of the SDV and insertion 

of control rods ensures that the proposed 

change does not involve a significant 

reduction in the margin of safety. 

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: Michael L. 

Marshall, Jr. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–440, 

Perry Nuclear Power Plant, Unit 1, Lake 

County, Ohio, and Docket Nos. 50–334 

and 50–412, Beaver Valley Power 

Station, Unit Nos. 1 and 2, Beaver 

County, Pennsylvania 

Date of amendment request: January 

11, 2007. 

Description of amendment request: 

The proposed license amendments 

would modify technical specification 

(TS) requirements for inoperable 

snubbers by adding Limiting Condition 

for Operation (LCO) 3.0.8. The proposed 

license amendments also modify LCO 

3.0.1 to incorporate the addition of LCO 

3.0.8. This change is based on the TS 

Task Force (TSTF) Traveler, TSTF–372, 

Revision 4. A notice of availability for 

this TS improvement using the 

consolidated line item improvement 

process was published in the Federal 

Register on May 4, 2005. 

The Nuclear Regulatory Commission 

(NRC) staff issued a notice of 

availability of a model no significant 

hazards consideration (NSHC) 

determination for referencing license 

amendment applications in the Federal 

Register on November 24, 2004 (69 FR 

68412), and May 4, 2005 (70 FR 23252). 

The licensee affirmed the applicability 

of the model NSHC determination in its 

application dated January 11, 2007. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. The Proposed Change Does Not Involve 

a Significant Increase in the Probability or 

Consequences of an Accident Previously 

Evaluated. 

The proposed change allows a delay time 

for entering a supported system technical 

specification (TS) when the inoperability is 

due solely to an inoperable snubber if risk is 

assessed and managed. The postulated 

seismic event requiring snubbers is a lowprobability occurrence and the overall TS 

system safety function would still be 

available for the vast majority of anticipated 

challenges. Therefore, the probability of an 

accident previously evaluated is not 

significantly increased, if at all. The 

consequences of an accident while relying on 

allowance provided by proposed LCO 3.0.8 

are no different than the consequences of an 

accident while relying on the TS required 

actions in effect without the allowance 

provided by proposed LCO 3.0.8. Therefore 

the consequences of an accident previously 

evaluated are not significantly affected by 

this change. The addition of a requirement to 

assess and manage the risk introduced by this 

change will further minimize possible 

concerns. Therefore, this change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. The Proposed Change Does Not Create 

the Possibility of a New or Different Kind of 

Accident From Any Previously Evaluated. 

The proposed change does not involve a 

physical alteration of the plant (no new or 

different type of equipment will be installed). 

Allowing delay times for entering supported 

system TS when inoperability is due solely 

to inoperable snubbers, if risk is assessed and 

managed, will not introduce new failure 

modes or effects and will not, in the absence 

of other unrelated failures, lead to an 

accident whose consequences exceed the 

consequences of accidents previously 

evaluated. The addition of a requirement to 

assess and manage the risk introduced by this 

change will further minimize possible 

concerns. Thus, this change does not create 

the possibility of a new or different kind of 

accident from an accident previously 

evaluated. 

3. The Proposed Change Does Not Involve 

a Significant Reduction in the Margin of 

Safety. 

The proposed change allows a delay time 

for entering a supported system TS when the 

inoperability is due solely to an inoperable 

snubber, if risk is assessed and managed. The 

postulated seismic event requiring snubbers 

is a low-probability occurrence and the 

overall TS system safety function would still 

be available for the vast majority of 

anticipated challenges. The risk impact of the 

proposed TS changes was assessed following 

the three-tiered approach recommended in 

RG 1.177. A bounding risk assessment was 

performed to justify the proposed TS 

changes. This application of LCO 3.0.8 is 

predicated upon the licensee’s performance 

of a risk assessment and the management of 

plant risk. The net change to the margin of 

safety is insignificant. Therefore, this change 

does not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: David W. 

Jenkins, Attorney, FirstEnergy 

Corporation, 76 South Main Street, 

Akron, OH 44308. 

NRC Branch Chief: Michael L. 

Marshall, Jr. 


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Nine Mile Point Nuclear Station 

(NMPNS), LLC, Docket No. 50–220, Nine 

Mile Point Nuclear Station Unit No. 1 

(NMP1), Oswego County, New York 

Date of amendment request: 

December 14, 2006. 

Description of amendment request: 

The proposed license amendment 

would revise the accident source term 

used in the NMP1 design basis 

radiological consequence analyses in 

accordance with 10 CFR 50.67. The 

revised accident source term replaces 

the current methodology that is based 

on TID–14844, ‘‘Calculation of Distance 

Factors for Power and Test Reactor 

Sites,’’ with the alternative source term 

(AST) methodology described in 

Regulatory Guide (RG) 1.183, 

‘‘Alternative Source Terms for 

Evaluating Design Basis Accidents at 

Nuclear Power Reactors.’’ The 

amendment request is for full 

implementation of the AST as described 

in RG 1.183, with the exception that 

TID–14844 will continue to be used as 

the radiation dose basis for equipment 

qualification and vital area access. 

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. 

Adoption of the AST and those plant 

systems affected by implementing AST do 

not initiate DBAs [design-basis accidents]. 

The AST does not affect the design or 

manner in which the facility is operated; 

rather, for postulated accidents, the AST is 

an input to calculations that evaluate the 

radiological consequences. The AST does not 

by itself affect the post-accident plant 

response or the actual pathway of the 

radiation released from the fuel. It does, 

however, better represent the physical 

characteristics of the release, so that 

appropriate mitigation techniques may be 

applied. Implementation of the AST has been 

incorporated in the analyses for the limiting 

DBAs at NMP1. 

The structures, systems and components 

affected by the proposed change mitigate the 

consequences of accidents after the accident 

has been initiated. Application of the AST 

does result in changes to NMP1 Updated 

Final Safety Analysis Report (UFSAR) 

functions (e.g., Liquid Poison system). As a 

condition of the application of AST, NMPNS 

is proposing to use the Liquid Poison system 

to control the suppression pool pH following 

a LOCA [loss-of-coolant accident]. The 

proposed changes also revise operability 

requirements for the secondary containment 

and certain post-accident filtration systems 

while handling irradiated fuel that has 

decayed for greater than 24 hours and during 

core alterations. These changes have been 

included within the AST evaluations. These 

changes do not require any physical changes 

to the plant. As a result, 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 XV of the NMP1 

UFSAR. Since design basis accident initiators 

are not being altered by adoption of the AST, 

the probability of an accident previously 

evaluated is not affected. 

Plant-specific AST radiological analyses 

have been performed and, based on the 

results of these analyses, it has been 

demonstrated that the dose consequences of 

the limiting events considered in the 

analyses are within the acceptance criteria 

provided by the NRC for use with the AST. 

These criteria are presented in 10 CFR 50.67 

and Regulatory Guide 1.183. Even though the 

AST dose limits are not directly comparable 

to the previously specified whole body and 

thyroid dose guidelines of General Design 

Criterion 19 and 10 CFR 100.11, the results 

of the AST analyses have demonstrated that 

the 10 CFR 50.67 limits are satisfied. 

Therefore, it is concluded that adoption of 

the AST does not involve a significant 

increase in the consequences of an accident 

previously evaluated. 

Based on the above discussion, it is 

concluded that the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Implementation of AST and the proposed 

changes do not alter or involve any design 

basis accident initiators. These changes do 

not involve any physical changes to the plant 

and do not affect the design function or mode 

of operations of systems, structures, or 

components in the facility prior to a 

postulated accident. Since systems, 

structures, and components are operated 

essentially no differently after the AST 

implementation, no new failure modes are 

created by this proposed change. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The changes proposed are associated with 

a new licensing basis for analysis of NMP1 

DBAs. Approval of the licensing basis change 

from the original source term to the AST is 

being requested. The results of the accident 

analyses performed in support of the 

proposed changes are subject to revised 

acceptance criteria. The limiting DBAs have 

been analyzed using conservative 

methodologies, in accordance with the 

guidance contained in Regulatory Guide 

1.183, to ensure that analyzed events are 

bounding and that safety margin has not been 

reduced. The dose consequences of these 

limiting events are within the acceptance 

criteria presented in 10 CFR 50.67 and 

Regulatory Guide 1.183. Thus, the proposed 

changes continue to ensure that the doses at 

the exclusion area boundary and low 

population zone boundary, as well as in the 

control room, are within corresponding 

regulatory criteria. 

Therefore, by meeting the applicable 

regulatory criteria for AST, it is concluded 

that the proposed changes do not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mark J. 

Wetterhahn, Esquire, Winston & Strawn, 

1700 K Street, NW., Washington, DC 

20006. 

NRC Acting Branch Chief: John P. 

Boska. 

Nine Mile Point Nuclear Station 

(NMPNS), LLC, Docket No. 50–410, Nine 

Mile Point Nuclear Station Unit No. 2 

(NMP2), Oswego County, New York 

Date of amendment request: January 

4, 2007. 

Description of amendment request: 

The proposed license amendment 

would revise Technical Specification 

(TS) 3.7.1, ‘‘Service Water (SW) System 

and Ultimate Heat Sink (UHS),’’ as 

follows: Revise the existing Limiting 

Condition for Operation (LCO) 

statement to require four operable SW 

pumps to be in operation when SW 

subsystem supply header water 

temperature is ≤82 °F; add a 

requirement that five operable SW 

pumps be in operation when SW 

subsystem supply header water 

temperature is >82 °F and ≤84 °F; delete 

Condition G and the associated 

Required Actions and Completion 

Times; revise Surveillance Requirement 

3.7.1.3 to increase the maximum 

allowed SW subsystem supply header 

water temperature from 82 °F to 84 °F; 

and modify the requirements for 

increasing the surveillance frequency as 

the temperature approaches the limit. 

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. 


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The proposed change eliminates the 

requirement to perform temperature 

averaging when the UHS temperature is 

>82 °F, establishes 84 °F as the design limit 

for UHS water temperature for operation on 

a continuous basis, and revises the frequency 

for verifying that the UHS temperature is 

within the prescribed limit. The TS currently 

allow operation with the UHS water 

temperature temporarily exceeding 82 °F, up 

to a maximum of 84 °F. The UHS 

temperature itself is not an initiator of 

accidents analyzed in the Updated Safety 

Analysis Report (USAR). Raising the 

maximum temperature limit and revising the 

associated surveillance requirement 

frequency do not involve any plant hardware 

changes or new operator actions that could 

serve to initiate an accident. Continuous 

operation with the elevated UHS temperature 

may result in a few balance-of-plant 

equipment high temperature alarms. 

Operator response to these alarms would be 

in accordance with established alarm 

response procedures. In all cases, trip 

setpoints leading to a reactor scram or a 

power runback will not be reached, and the 

likelihood of component failures that could 

initiate an accident will not be significantly 

increased. 

The potential impact of the proposed 

change on the ability of the plant to mitigate 

postulated accidents has been evaluated. 

These evaluations demonstrate that safetyrelated systems and components that rely on 

the UHS as the cooling medium or as a pump 

suction source are capable of performing 

their intended safety functions at the higher 

UHS temperature, and that containment 

integrity and equipment qualification are 

maintained. The calculated post-accident 

dose consequences reflected in the USAR do 

not directly utilize UHS temperature as an 

input and thus are not impacted by the 

proposed change. 

Based on the above, the proposed change 

will have no adverse effect on plant 

operation or the availability or operation of 

any accident mitigation equipment. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change will not alter the 

current plant configuration (no new or 

different type of equipment will be installed) 

or require any new or unusual operator 

actions. The proposed change will not alter 

the way any structure, system, or component 

functions and will not cause an adverse effect 

on plant operation or accident mitigation 

equipment. The response of the plant and the 

operators following a design-basis accident is 

unaffected by the change. The proposed 

change does not introduce any credible new 

failure mechanisms, malfunctions, or 

accident initiators not considered in the 

design and licensing bases. Analyses have 

shown that the design basis heat removal 

capability of the affected safety-related 

components is maintained at the increased 

UHS water temperature limit. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The margin of safety is determined by the 

design and qualification of the plant 

equipment, the operation of the plant within 

analyzed limits, and the point at which 

protective or mitigative actions are initiated. 

The proposed change does not impact these 

factors. An evaluation of the safety systems 

has been performed to ensure their safety 

functions can be met for operation with a 

UHS water temperature of 84 °F on a 

continuous basis. Operation with the UHS 

water temperature temporarily exceeding 

82 °F, up to a maximum of 84 °F, is currently 

allowed. Operating on a continuous basis at 

the higher UHS temperature represents a 

slight reduction in design margins in terms 

of the ability of affected systems to remove 

accident heat loads. However, the evaluation 

has demonstrated that the proposed change 

does not have a significant impact on the 

capability of the affected systems to perform 

their safety-related post-accident functions 

and to mitigate accident consequences. The 

design limits for the containment and fuel 

cladding will not be exceeded, and 

equipment qualification will be maintained. 

No protection setpoints are affected by the 

proposed change. The revised frequency for 

performing the TS surveillance to verify that 

the UHS temperature is within the prescribed 

limit will continue to assure that plant 

operators are aware of and are monitoring 

increasing UHS temperature trends prior to 

reaching a value of 82 °F, when a fifth SW 

pump must be placed in operation. This 

action is no different than that required by 

the current TS. 

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: Mark J. 

Wetterhahn, Esquire, Winston & Strawn, 

1700 K Street, NW., Washington, DC 

20006. 

NRC Acting Branch Chief: John P. 

Boska. 

Nuclear Management Company, LLC, 

Docket No. 50–263, Monticello Nuclear 

Generating Plant, Wright County, 

Minnesota 

Date of amendment request: January 

29, 2007. 

Description of amendment request: 

The proposed amendment would revise 

Table 3.3.5.1–1, ‘‘Emergency Core 

cooling System Instrumentation,’’ of the 

Technical Specifications (TSs) to extend 

the quarterly surveillance interval from 

quarterly to a nominal 24-month 

interval for three low pressure coolant 

injection loop select logic functions. 

Consistent with the extended test 

interval, the licensee also proposed to 

change the allowable values associated 

with each of the three logic functions 

(i.e., response time in seconds). The 

licensee stated that the quarterly 

surveillance requirement was 

inappropriately introduced when the 

TSs was converted from its previous 

custom format to the current Improved 

Technical Specification format by 

Amendment No. 146. Before the 

conversion, there was no such quarterly 

surveillance requirement. Furthermore, 

the plant was not designed to have these 

three logic functions tested while online. 

Basis for proposed no significant 

hazards consideration determination: 

As required by Title 10 of the Code of 

Federal Regulations (10 CFR) Part 

50.91(a), the licensee has provided its 

analysis of the issue of no significant 

hazards consideration (NSHC). The NRC 

staff reviewed the licensee’s analysis, 

and has performed its own analysis as 

follows: 

(1) Does the proposed amendment 

involve a significant increase in the 

probability or consequences of an 

accident previously evaluated? 

No. The proposed amendment would 

extend the performance interval from 

quarterly to a 24-month interval, and 

change the associated allowable values 

for the three logic functions. The 

performance of these surveillances, or 

the failure to perform, as well as the 

surveillance finding (i.e., response time 

in seconds) are not precursors to, and do 

not affect the probability of, an accident. 

There is no design or operation change 

associated with the proposed 

amendment. Therefore, the proposed 

amendment does not increase the 

probability of an accident previously 

evaluated. 

A delay in performing these 

surveillances would not result in a 

system being unable to perform its 

required function. The extended 

surveillance and associated changed 

allowable values will not affect the three 

logic functions to operate as designed. 

Therefore, the plant systems required to 

mitigate accidents will remain capable 

of performing their design function. As 

a result, the proposed amendment will 

not lead to any significant change in the 

consequences of any accident. 

(2) Does the proposed amendment 

create the possibility of a new or 

different kind of accident from any 

accident previously evaluated? 


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No. The proposed amendment does 

not involve a physical alteration of any 

system, structure, or component (SSC) 

or a change in the way any SSC is 

operated. The proposed amendment 

does not involve operation of any SSCs 

in a manner or configuration different 

from those previously recognized or 

evaluated. No new failure mechanisms 

will be introduced by the extended 

surveillance interval and associated 

allowable values. Thus, 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? 

No. The proposed amendment only 

changes the surveillance interval and 

associated allowable values for the three 

logic functions. There will be no 

modification of any TSs limiting 

condition for operation, no change to 

any limit on previously analyzed 

accidents, no change to how previously 

analyzed accidents or transients would 

be mitigated, no change in any 

methodology used to evaluate 

consequences of accidents, and no 

change in any operating procedure or 

process. The instrumentation and 

components involved in this proposed 

amendment have exhibited reliable 

operation based on the results of their 

performance during past periodic 

emergency core cooling system 

functional testing. Therefore, the 

proposed amendment does not involve 

a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on the 

NRC staff’s own analysis above, it 

appears that the three standards of 10 

CFR 50.92(c) are satisfied. Therefore, the 

NRC staff proposes to determine that the 

proposed amendment 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 Branch Chief: L. Raghavan. 

Nuclear Management Company, LLC, 

Docket Nos. 50–282 and 50–306, Prairie 

Island Nuclear Generating Plant, Units 

1 and 2, Goodhue County, Minnesota 

Date of amendment request: January 

29, 2007. 

Description of amendment request: 

The proposed amendments would 

revise technical specification (TS) 3.5.3, 

‘‘ECCS (Emergency Core Cooling 

Systems)—Shutdown’’ operability 

requirements for the Safety Injection (SI) 

subsystem. These revisions will allow 

the required SI pump to be rendered 

incapable of injecting into the Reactor 

Coolant System (RCS) during low 

temperature (MODE 4) operations due to 

a single action or automatic signal. The 

capability of the plant operators to 

initiate SI flow on a timely basis will be 

maintained. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This license amendment request proposes 

to add a new Note to Technical Specification 

3.5.3, ‘‘Emergency Core Cooling System—

Shutdown’’. This Note will allow the Safety 

Injection system to be considered operable 

within the Limiting Condition for Operation 

requirements while the system is not capable 

of automatic injection provided it is capable 

of being manually aligned for injection. 

This Emergency Core Cooling System is 

not an accident initiator, thus the proposed 

changes do not increase the probability of an 

accident. The current licensing basis, 

Technical Specifications and Bases do not 

require automatic initiation instrumentation 

for the Emergency Core Cooling System in 

Mode 4, but rather assume operator action to 

mitigate an accident. With the proposed 

Technical Specification and Bases changes, 

the Emergency Core Cooling System will 

continue to be operable for manual initiation. 

Since the changes proposed in this license 

amendment request do not impact the 

performance of the system, these changes do 

not involve a significant increase in the 

consequences of an accident previously 

evaluated. 

The changes proposed in this license 

amendment do not involve a significant 

increase in the probability or consequences 

of an accident previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

This license amendment request proposes 

to add a new Note to Technical Specification 

3.5.3, ‘‘Emergency Core Cooling System—

Shutdown’’. This Note will allow the Safety 

Injection system to be considered operable 

within the Limiting Condition for Operation 

requirements while the system is not capable 

of automatic injection provided it is capable 

of being manually aligned for injection. 

The changes proposed for the Emergency 

Core Cooling System Technical 

Specifications do not change any system 

operations, maintenance activities or testing 

requirements. The Limiting Condition for 

Operation will continue to be met, no new 

failure modes or mechanisms are created and 

no new accident precursors are generated by 

this change. The Technical Specification 

changes proposed in this license amendment 

do not create the possibility of a new or 

different kind of accident from any 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

This license amendment request proposes 

to add a new Note to Technical Specification 

3.5.3, ‘‘Emergency Core Cooling System—

Shutdown’’. This Note will allow the Safety 

Injection system to be considered operable 

within the Limiting Condition for Operation 

requirements while the system is not capable 

of automatic injection provided it is capable 

of being manually aligned for injection. 

The current licensing basis, Technical 

Specifications and Bases rely upon operator 

actions to initiate safety injection to mitigate 

an accident in Mode 4 and do not require 

operability of any process instrumentation 

capable of automatically initiating the 

Emergency Core Cooling System. With the 

changes proposed in this license amendment 

request, the safety injection system will 

continue to be operable and the plant will 

continue to rely on operator actions for safety 

injection initiation. Thus, the Technical 

Specification changes proposed in this 

license amendment request do not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment requests involve 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: P. Milano. 

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 

11, 2006, as supplemented on October 

25, November 21, and December 4, 

2006. 

Description of amendment request: 

The proposed amendments would 

increase the SSES 1 and 2 licensed 

thermal power to 3952 Mega-watts 

thermal (MWt), which is 20% above the 

original rated thermal power (RTP) of 

3293 MWt, and approximately 13% 

above the current RTP of 3489 MWt. 

The proposed amendments would 

revise the SSES 1 and 2 Operating 

License and Technical Specifications 

necessary to implement the increased 

power level. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 


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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? 

Extended Power Uprate 

Response: No. 

The probability (frequency of occurrence) 

of Design Basis Accidents occurring is not 

affected by the increased power level, 

because Susquehanna continues to comply 

with the regulatory and design basis criteria 

established for plant equipment. A 

probabilistic risk assessment demonstrates 

that the calculated core damage frequencies 

do not significantly change due to Constant 

Pressure Power Uprate (CPPU). Scram 

setpoints (equipment settings that initiate 

automatic plant shutdowns) are established 

such that there is no significant increase in 

scram frequency due to CPPU. No new 

challenges to safety-related equipment result 

from CPPU. 

The changes in consequences of postulated 

accidents, which would occur from 102% of 

the CPPU (rated thermal power) RTP 

compared to those previously evaluated, are 

acceptable. The results of CPPU accident 

evaluations do not exceed the NRC-approved 

acceptance limits. The spectrum of 

postulated accidents and transients has been 

investigated, and are shown to meet the 

plant’s currently licensed regulatory criteria. 

In the area of fuel and core design, for 

example, the Safety Limit Minimum Critical 

Power Ratio (SLMCPR) and other applicable 

Specified Acceptable Fuel Design Limits 

(SAFDLS) are still met. Continued 

compliance with the SLMCPR and other 

SAFDLs will be confirmed on a cycle specific 

basis consistent with the criteria accepted by 

the NRC. 

Challenges to the Reactor Coolant Pressure 

Boundary were evaluated at CPPU conditions 

(pressure, temperature, flow, and radiation) 

were found to meet their acceptance criteria 

for allowable stresses and overpressure 

margin. 

Challenges to the containment have been 

evaluated, and the containment and its 

associated cooling systems continue to meet 

10 CFR [Part] 50, Appendix A, Criterion 16, 

Containment Design; Criterion 38, 

Containment Heat Removal; and Criterion 50, 

Containment Design Basis. The increase in 

the calculated post LOCA [loss-of-coolant 

accident] suppression pool temperature 

above the currently assumed peak 

temperature was evaluated and determined 

to be acceptable. 

Radiological release events (accidents) 

have been evaluated, and shown to meet the 

guidelines of 10 CFR 50.67. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

LPRM [Local Power Range Monitor] 

Calibration Interval Technical Specification 

SR [Surveillance Requirement] Frequency 

Change 

Response: No. 

The revised surveillance interval continues 

to ensure that the LPRM signal is adequately 

calibrated. This change will not alter the 

basic operation of process variables, 

structures, systems, or components as 

described in the SSES FSAR [final safety 

analysis report], and no new equipment is 

introduced by the change in LPRM 

surveillance interval. The performance of the 

APRM [average power range monitor] and 

RBM [rod block monitor] systems is not 

significantly affected by the proposed LPRM 

surveillance interval increase. Therefore, the 

probability of accidents previously evaluated 

is unchanged. 

The proposed change results in no change 

in radiological consequences of the design 

basis LOCA as currently analyzed for SSES. 

The consequences of an accident can be 

affected by the thermal limits existing at the 

time of the postulated accident, but LPRM 

chamber exposure has no significant effect on 

the calculated thermal limits because LPRM 

accuracy does not significantly deviate with 

exposure. For the extended calibration 

interval, the assumption in the safety limit 

analysis remains valid, maintaining the 

accuracy of the thermal limit calculation. 

Therefore, the thermal limit calculation is not 

significantly affected by LPRM calibration 

frequency and the consequences of an 

accident previously evaluated are 

unchanged. 

The change does not affect the initiation of 

any event, nor does it negatively impact the 

mitigation of any event. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

RHR [Residual Heat Removal] Service Water 

System and Ultimate Heat Sink Technical 

Specification and Methods Change 

Response: No. 

The proposed changes do not involve any 

new initiators for any accidents nor do they 

increase the likelihood of a malfunction of 

any Structures, Systems or Components 

(SSCs). Implementation of the subject 

changes reduces the probability of adverse 

consequences of accidents previously 

evaluated, because inclusion of the manual 

spray array bypass isolation valves and the 

small spray array isolation valves in the 

Technical Specifications (TS) increases their 

reliability to function for safe shutdown. The 

use of the ANS/ANSI–5.1–1979 decay heat 

model in the UHS [ultimate heat sink] 

performance analysis is not relevant to 

accident initiation, but rather, pertains to the 

method used to evaluate currently postulated 

accidents. Its use does not, in any way, alter 

existing fission product boundaries, and 

provides a conservative prediction of decay 

heat. Therefore, the change in decay heat 

calculational method does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

Containment Analysis Methods Change 

Response: No. 

The use of passive heat sinks, and the 

ANS/ANSI–5.1–1979 decay heat model are 

not relevant to accident initiation, but rather, 

pertain to the method used to evaluate 

postulated accidents. The use of these 

elements does not, in any way, alter existing 

fission product boundaries, and provides a 

conservative prediction of the containment 

response to DBA [design-basis accident]- 

LOCAs. Therefore[,] the Containment 

Analysis Method Change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

Feedwater Pump/Condensate Pump Trip 

Change 

Response: No. 

Feedwater pump trips and condensate 

pump trips rarely occur. A low water level 

SCRAM on loss of one feedwater pump or 

one condensate pump is bounded by the loss 

of all feedwater transient in Final Safety 

Analysis Report (FSAR) Appendix 15E. A 

trip of one feedwater pump or a trip of one 

condensate pump does not result in the loss 

of all feedwater. The Feedwater Pump / 

Condensate Pump Trip Change is included in 

the CPPU Probabilistic Risk Assessment 

(PRA). The best estimate for the Susquehanna 

Steam Electric Station (SSES) Core Damage 

Frequency (CDF) risk increase due to the 

CPPU is 6E–08 for Unit 1 and 7E–08 for Unit 

2 which are in the lower left corner of Region 

III of Regulatory Guide [sic] (Reference 15) 

(i.e., very small risk changes). The best 

estimate for the Large Early Release 

Frequency (LERF) increase is 1.0E–09/yr for 

both units which is also in the lower left 

corner of the Region III range of Regulatory 

Guide 1.174. Therefore, the Feedwater Pump/ 

Condensate Pump Trip Change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

Main Turbine Pressure Regulation System 

Response: No. 

Technical Specification 3.7.8 does not 

directly or indirectly affect any plant system, 

equipment, component, or change the 

process used to operate the plant. Technical 

Specification 3.7.8 would ensure acceptable 

performance, since it would establish 

requirements for adhering to the appropriate 

thermal limits, depending on the operability 

of the main turbine pressure regulation 

system. Use of the appropriate limits assures 

that the appropriate safety limits will not be 

exceeded during normal or anticipated 

operational occurrences. Thus, Technical 

Specification 3.7.8 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? 

Extended Power Uprate 

Response: No. 

Equipment that could be affected by EPU 

has been evaluated. No new operating mode, 

safety-related equipment lineup, accident 

scenario, or equipment failure mode was 

identified. The full spectrum of accident 


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considerations has been evaluated and no 

new or different kind of accident has been 

identified. CPPU uses developed technology 

and applies it within capabilities of existing 

or modified plant safety related equipment in 

accordance with the regulatory criteria 

(including NRC approved codes, standards 

and methods). No new accidents or event 

precursors have been identified. 

The SSES TS require revision to 

implement EPU. The revisions have been 

assessed and it was determined that the 

proposed change will not introduce a 

different accident than that previously 

evaluated. Therefore[,] the proposed changes 

do not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

LPRM Calibration Interval Technical 

Specification SR Frequency Change 

Response: No. 

The proposed change will not physically 

alter the plant or its mode of operation. The 

performance of the APRM and RBM systems 

is not significantly affected by the proposed 

LPRM surveillance interval increase. As 

such, no new or different types of equipment 

will be installed and the basic operation of 

installed equipment is unchanged. The 

methods of governing plant operation and 

testing are consistent with current safety 

analysis assumptions. Therefore, the 

proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

RHR Service Water System and Ultimate 

Heat Sink Technical Specification and 

Methods Change 

Response: No. 

The subject changes apply Technical 

Specification controls to new UHS manual 

bypass isolation valves and the existing small 

spray array isolation valves. The design 

functions of the systems are not affected. 

The addition of manually operated valves 

in the system, operational changes and the 

Technical Specification changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

The use of the ANS/ANSI–5.1–1979 decay 

heat model is not relevant to accident 

initiation, but rather pertains to the method 

used to evaluate currently postulated 

accidents. The use of this analytical tool does 

not involve any physical changes to plant 

structures or systems, and does not create a 

new initiating event for the spectrum of 

events currently postulated in the FSAR. 

Further, it does not result in the need to 

postulate any new accident scenarios. 

Therefore[,] the decay heat calculational 

method change does not create the possibility 

of a new or different kind of accident from 

any accident previously evaluated[.] 

Containment Analysis Methods Change 

Response: No. 

The use of passive heat sinks and the ANS/ 

ANSI–5.1–1979 decay heat model are not 

relevant to accident initiation, but pertain to 

the method used to evaluate currently 

postulated accidents. The use of these 

analytical tools does not involve any physical 

changes to plant structures or systems, and 

does not create a new initiating event for the 

spectrum of events currently postulated in 

the FSAR. Further, they do not result in the 

need to postulate any new accident 

scenarios. Therefore, the Containment 

Analysis Method Change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

Feedwater Pump/Condensate Pump Trip 

Change 

Response: No. 

The occurrence of a reactor SCRAM is 

already considered in the current licensing 

basis and is not an accident. A SCRAM 

resulting from the trip of a feedwater pump 

or a condensate pump is bounded by a loss 

of all feedwater event. The loss of all 

feedwater transient is already considered in 

the plant licensing basis. The SCRAM due to 

the feedwater or condensate pump trip does 

not change the results of the loss of all 

feedwater transient in any way. Therefore, 

the Feedwater Pump/Condensate Pump Trip 

Change does not create the possibility of a 

new or different kind of accident from any 

accident previously evaluated. 

Main Turbine Pressure Regulation System 

Response: No. 

Technical Specification 3.7.8 will not 

directly or indirectly affect any plant system, 

equipment, or component and therefore does 

not affect the failure modes of any of these 

items. Thus, Technical Specification 3.7.8 

does not create the possibility of a previously 

unevaluated operator error or a new single 

failure. 

Therefore, Technical Specification 3.7.8 

does not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Extended Power Uprate 

Response: No. 

The CPPU affects only design and 

operational margins. Challenges to the fuel, 

reactor coolant pressure boundary, and 

containment were evaluated for CPPU 

conditions. Fuel integrity is maintained by 

meeting existing design and regulatory limits. 

The calculated loads on affected structures, 

systems and components, including the 

reactor coolant pressure boundary, will 

remain within their design allowables for 

design basis event categories. No NRC 

acceptance criterion is exceeded. Because the 

SSES configuration and responses to 

transients and postulated accidents do not 

result in exceeding the presently approved 

NRC acceptance limits, the proposed changes 

do not involve a significant reduction in a 

margin of safety. 

LPRM Calibration Interval Technical 

Specification Change 

Response: No. 

The proposed change has no impact on 

equipment design or fundamental operation 

and there are no changes being made to 

safety limits or safety system allowable 

values that would adversely affect plant 

safety as a result of the proposed change. The 

performance of the APRM and RBM systems 

is not significantly affected by the proposed 

LPRM surveillance interval increase. The 

margin of safety can be affected by the 

thermal limits existing prior to an accident; 

however, uncertainties associated with LPRM 

chamber exposure have no significant effect 

on the calculated thermal limits. For the 

extended calibration interval, the assumption 

in the safety limit analysis remains valid, 

maintaining the accuracy of the thermal limit 

calculation. 

Since the proposed change does not affect 

safety analysis assumptions or initial 

conditions, the margin of safety in the safety 

analyses are maintained. Therefore, the 

proposed change does not involve a 

significant reduction in a margin of safety. 

RHR Service Water System and Ultimate 

Heat Sink Technical Specification and 

Methods Change 

Response: No. 

Implementation of the subject changes 

does not significantly reduce the margin of 

safety since these changes add components 

and Technical Specification controls for the 

components not currently addressed in the 

Technical Specifications. These changes 

increase the reliability of the affected 

components/systems to function for safe 

shutdown. 

Therefore[,] these changes do not involve 

a significant reduction in margin of safety. 

The ANS/ANSI–5.1–1979 model provides 

a conservative prediction of decay heat. The 

use of this element is consistent with current 

industry standards, and has been previously 

accepted by the staff for use in containment 

analysis by other licensees, as described in 

GE Nuclear Energy. ‘‘Constant Pressure 

Power Uprate,’’ Licensing Topical Report 

NEDC–33004P–A, Revision 4, dated July 

2003; and the letter to Gary L. Sozzi (GE) 

from Ashok Thandani (NRC) on the Use of 

the SHEX Computer Program and ANSI/ANS 

5.1–1979, ‘‘Decay Heat Source Term for 

Containment Long-Term Pressure and 

Temperature Analysis,’’ July 13, 1993. 

Therefore, the decay heat calculational 

method change does not involve a significant 

reduction in the margin of safety. 

Containment Analysis Methods Change 

Response: No. 

The use of passive heat sinks and the ANS/ 

ANSI–5.1–1979 decay heat model are 

realistic phenomena, and provide a 

conservative prediction of the plant response 

to DBA–LOCAs. The use of these elements is 

consistent with current industry standards, 

and has been previously accepted by the staff 

for other licensees, as described in GE 

Nuclear Energy: ‘‘Constant Pressure Power 

Uprate,’’ Licensing Topical Report NEDC–

33004P–A, Revision 4, dated July 2003; the 

letter to Gary L. Sozzi (GE) from Ashok 

Thandani (NRC) on the Use of the SHEX 

Computer Program; and ANSI/ANS 5.1–1979, 

‘‘Decay Heat Source Term for Containment 

Long-Term Pressure and Temperature 

Analysis,’’ July 13, 1993. Therefore the 

Containment Analysis Method Change does 

not involve a significant reduction in [a] 

margin of safety. 


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Feedwater Pump/Condensate Pump Trip 

Change 

Response: No. 

A low water level SCRAM on loss of one 

feedwater pump or one condensate pump is 

bounded by the loss of all feedwater transient 

in FSAR Appendix 15E. The loss of all 

feedwater transient is a non-limiting event 

that does not contribute to the setting of the 

fuel safety limits. Consequently, a SCRAM 

resulting from a feedwater pump or 

condensate pump trip does not reduce the 

margin to fuel safety limits. Therefore, the 

potential for a SCRAM resulting from a 

feedwater pump trip or a condensate pump 

trip does not involve a significant reduction 

in [a] margin of safety. 

Main Turbine Pressure Regulation System 

Since Technical Specification 3.7.8 does 

not alter any plant system, equipment, 

component, or processes used to operate the 

plant, the proposed change will not 

jeopardize or degrade the function or 

operation of any plant system or component 

governed by Technical Specifications. 

Technical Specification 3.7.8 preserves the 

margin of safety by establishing requirements 

for adhering to the appropriate thermal 

limits. 

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 (Acting) Branch Chief: Douglas 

V. Pickett. 

Southern Nuclear Operating Company, 

Inc., Georgia Power Company, 

Oglethorpe Power Corporation, 

Municipal Electric Authority of Georgia, 

City of Dalton, Georgia, Docket Nos. 50–

321 and 50–366, Edwin I. Hatch Nuclear 

Plant, Units 1 and 2, Appling County, 

Georgia 

Date of amendment request: February 

2, 2007. 

Description of amendment request: 

The proposed amendments would 

revise Technical Specification (TS) LCO 

3.10.1 to expand its scope to include 

provisions for temperature excursions 

greater than 212 degrees F as a 

consequence of scram time testing 

initiated in conjunction with an 

inservice leak or hydrostatic test. During 

these tests and with temperature greater 

than 212 degrees F, operational 

conditions are considered to be in Mode 

4. 

The NRC staff issued a notice of 

availability of a model safety evaluation 

and model no significant hazards 

consideration (NSHC) determination for 

referencing in license amendment 

applications in the Federal Register on 

October 27, 2006 (71 FR 63050). The 

licensee affirmed the applicability of the 

model NSHC determination in its 

application dated February 2, 2007. 

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 

Technical Specifications currently allow 

for operation at greater than 212 deg F while 

imposing MODE 4 requirements in addition 

to the secondary containment requirements 

required to be met. Extending the activities 

that can apply this allowance will not 

adversely impact the probability or 

consequences of an accident previously 

evaluated. Therefore, the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

Criterion 2—The Proposed Change Does Not 

Create the Possibility of a New or Different 

Kind of Accident From Any Accident 

Previously Evaluated 

Technical Specifications currently allow 

for operation at greater than 212 deg F while 

imposing MODE 4 requirements in addition 

to the secondary containment requirements 

required to be met. No new operational 

conditions beyond those currently allowed 

by LCO 3.10.1 are introduced. The changes 

do not involve a physical alteration of the 

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

equipment will be installed) or a change in 

the methods governing normal plant 

operation. In addition, the changes do not 

impose any new or different requirements or 

eliminate any existing requirements. The 

changes do not alter assumptions made in the 

safety analysis. The proposed changes are 

consistent with the safety analysis 

assumptions and current plant operating 

practice. Therefore, the proposed change 

does not create the possibility of a new or 

different kind of accident from any accident 

previously evaluated. 

Criterion 3—The Proposed Change Does Not 

Involve a Significant Reduction in a Margin 

of Safety 

Technical Specifications currently allow 

for operation at greater than 212 deg F while 

imposing MODE 4 requirements in addition 

to the secondary containment requirements 

required to be met. Extending the activities 

that can apply this allowance will not 

adversely impact any margin of safety. 

Allowing completion of inspections and 

testing and supporting completion of scram 

time testing initiated in conjunction with an 

inservice leak or hydrostatic test prior to 

power operation results in enhanced safe 

operations by eliminating unnecessary 

maneuvers to control reactor temperature and 

pressure. Therefore, the proposed change 

does not involve a significant reduction in a 

margin of safety. 

The NRC staff proposes to determine 

that the amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Ernest L. Blake, 

Jr., Esquire, Shaw, Pittman, Potts and 

Trowbridge, 2300 N Street, NW., 

Washington, DC 20037. 

NRC Branch Chief: Evangelos C. 

Marinos. 

Tennessee Valley Authority, Docket No. 

50–328, Sequoyah Nuclear Plant, Unit 2, 

Hamilton County, Tennessee 

Date of amendment request: January 

12, 2007. 

Description of amendment request: 

The proposed amendment would revise 

the steam generator (SG) program 

requirements in the Sequoyah (SQN) 

Unit 2 Technical Specifications (TSs) to 

allow use of an SG voltage-based repair 

criteria probability of detection (POD) 

method using plant-specific SG tube 

inspection results. The proposed POD 

method is referred to as the probability 

of prior cycle detection (POPCD) 

method. 

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 use of a revised SG 

voltage-based repair criteria POD method, the 

POPCD method, to determine the BOC 

[beginning of cycle] indication voltage 

distribution for the SQN Unit 2 operational 

assessments does not increase the probability 

of an accident. Based on industry and plantspecific bobbin detection data for ODSCC 

[outside diameter stress corrosion cracking] 

within the SG tube support plate (TSP) 

region, large voltage bobbin indications 

which individually can challenge structural 

or leakage integrity can be detected with near 

100 percent certainty. Since large voltage 

outside diameter stress corrosion cracking 

ODSCC bobbin indications within the SG 

TSP can be detected, they will not be left in 

service, and therefore these indications 

should not be included in the voltage 

distribution for the purpose of operational 

assessments. The POPCD method improves 

the estimate of potentially undetected 

indications for operational assessments, but 

does not directly affect the inspection results. 

Since large voltage indications are detected, 

they will not result in an increase in the 

probability of SG tube rupture accident or an 

increase in the consequences of a tube 

rupture or main steam line break (MSLB) 

accident. 


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2. Does the proposed change create the 

possibility of a new or different accident 

from any accident previously evaluated? 

Response: No. 

The use of the POPCD method is associated 

with numerical predictions of probabilities 

for the steam generator tube rupture (SGTR) 

accident. Since the SGTR accident is 

considered in SQN’s Updated Final Safety 

Analysis Report, there is no possibility to 

create a design basis accident that has not 

been previously evaluated. Therefore, the 

proposed change does not create the 

possibility of a new or different accident 

from any accident previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. The use of the POPCD 

method to determine the BOC voltage 

distribution for the SQN Unit 2 operational 

assessments does not involve a significant 

reduction in a margin of safety. The 

applicable margin of safety potentially 

impacted is the SG tube structural and 

leakage criteria. Based on industry and plantspecific bobbin detection data for ODSCC 

within the SG TSP region, large voltage 

bobbin indications that can individually 

challenge structural or leakage integrity can 

be detected with near 100 percent certainty 

and will not be left in service. Therefore, 

these indications should not be included in 

the voltage distribution for the purposes of 

operational assessments. Since these large 

voltage indications are detected, they will not 

result in a significant increase in the actual 

EOC [end of cycle] leakage for a MSLB 

accident or the actual EOC probability of 

burst. The POPCD method approach to POD 

considers the potential for missing 

indications that might challenge structural or 

leakage integrity by applying the POPCD data 

from successive inspections. If a large 

indication was missed in one inspection, it 

would continue to grow until detected in a 

later inspection. Accordingly, there is no 

significant increase 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: General 

Counsel, Tennessee Valley Authority, 

400 West Summit Hill Drive, ET 11A, 

Knoxville, Tennessee 37902. 

NRC Branch Chief: Brenda Mozafari 

(Acting). 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of amendment request: February 

2, 2007. 

Description of amendment request: 

The proposed amendment would revise 

Technical Specification 3.6.1.7, 

‘‘Suppression Chamber-to-Drywell 

Vacuum Breakers,’’ to allow a one-time 

extension to the current closure 

verification surveillance requirement for 

one of two redundant disks in one of 

nine vacuum breakers until reliable 

position indication can be restored in 

the main control room during the next 

refueling outage (R–18), which is 

scheduled to begin on May 12, 2007. 

Date of publication of individual 

notice in Federal Register: February 

12, 2007 (72 FR 6606). 

Expiration date of individual notice: 

February 26, 2007. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of application for amendment: 

December 27, 2006. 

Brief description of amendment: The 

proposed amendment would revise 

Limiting Condition for Operation 3.14.A 

to adopt the Technical Specification 

Task Force 484, Revision 0, ‘‘Use of 

Technical Specification 3.10.1 for Scram 

Time Testing Activities.’’

Date of publication of individual 

notice in Federal Register: February 

20, 2007 (72 FR 7776). 

Expiration date of individual notice: 

March 22, 2007 (public comments) and 

April 23, 2007 (hearing requests). 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of application for amendment: 

January 15, 2007. 

Brief description of amendment: The 

amendment request supercedes the 

previously submitted license 

amendment request dated April 12, 

2006, proposing new PressureTemperature (PT) curves and to extend 

the applicability of current PT limits 

expressed in Technical Specification 

Figures 3.6.1, 3.6.2, and 3.6.3 through 

the end of operating cycle 18. 

Date of publication of individual 

notice in Federal Register: February 

12, 2007 (72 FR 6609). 

Expiration date of individual notice: 

March 14, 2007 (public comments) and 

April 13, 2007 (hearing requests). 

PSEG Nuclear LLC, Docket No. 50–272, 

Salem Nuclear Generating Station, Unit 

No. 1, Salem County, New Jersey 

Date of amendment request: January 

18, 2007. 

Brief description of amendment 

request: The amendment request 

proposes a one-time change to the 

Technical Specifications (TSs) regarding 

the steam generator (SG) tube inspection 

and repair required for the portion of 

the SG tubes passing through the 

tubesheet region. Specifically, for Salem 

Unit No. 1 refueling outage 18 (planned 

for spring 2007) and the subsequent 

operating cycle, the proposed TS 

changes would limit the required 

inspection (and repair if degradation is 

found) to the portions of the SG tubes 

passing through the upper 17 inches of 

the approximate 21-inch tubesheet 

region. 

Date of publication of individual 

notice in Federal Register: January 25, 

2007 (72 FR 3427). 

Expiration date of individual notice: 

February 26, 2007 (public comments) 

and March 26, 2007 (hearing requests). 

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 


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Federal Register / Vol. 72, No. 48 / Tuesday, March 13, 2007 / Notices 11397 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

Systems (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. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units Nos. 1, 2, and 

3, Maricopa County, Arizona 

Date of application for amendments: 

September 28, 2006. 

Brief description of amendments: The 

amendments revised Technical 

Specification (TS) requirements for 

mode change limitations in Limiting 

Condition for Operation (LCO) 3.0.4 and 

Surveillance Requirement 3.0.4 to adopt 

the provisions of Industry/TS Task 

Force (TSTF) Traveler number TSTF–

359, ‘‘Increase Flexibility in Mode 

Restraints.’’ The amendments also 

revised TS Example 1.4–1 to reflect the 

changes made to LCO 3.0.4 and to be 

consistent with TSTF–485, which has 

been incorporated into the Standard 

Technical Specifications Revision 3.1. 

Date of issuance: February 21, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: Unit 1–165, Unit 

2—165, Unit 3—165. 

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The 

amendments revised the Operating 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: November 7, 2006 (71 FR 

65140). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

February 21, 2007. 

No significant hazards consideration 

comments received: No. 

Arizona Public Service Company, et al., 

Docket Nos. STN 50–528, STN 50–529, 

and STN 50–530, Palo Verde Nuclear 

Generating Station, Units Nos. 1, 2, and 

3, Maricopa County, Arizona 

Date of application for amendments: 

February 14, 2006. 

Brief description of amendments: The 

amendments revised Technical 

Specification (TS) requirements in the 

Limiting Condition for Operation for TS 

3.6.3, ‘‘Containment Isolation Valves,’’

and associated Actions and Surveillance 

Requirements to allow for a blind flange 

to be used for containment isolation in 

each of the two flow paths of the 42- 

inch refueling purge valves in Modes 1 

through 4, without remaining in TS 

3.6.3 Condition D. 

Date of issuance: February 22, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment Nos.: Unit 1–166, Unit 

2–166, Unit 3–166. 

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The 

amendments revised the Operating 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: March 14, 2006 (71 FR 

13171). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 22, 

2007. 

No significant hazards consideration 

comments received: No. 

Carolina Power & Light Company, 

Docket Nos. 50–325 and 50–324, 

Brunswick Steam Electric Plant, Units 1 

and 2, Brunswick County, North 

Carolina 

Date of application for amendments: 

April 26, 2006. 

Brief Description of amendments: 

Revised the Technical Specification 

(TS) requirements for inoperable 

snubbers by adding Limiting Condition 

for Operation 3.0.8. 

Date of issuance: February 15, 2007. 

Effective date: February 15, 2007, 

implement within 90 days. 

Amendment Nos.: 241 and 269. 

Renewed Facility Operating License 

Nos. DPR–71 and DPR–62: Amendments 

change the TSs. 

Date of initial notice in Federal 

Register: June 6, 2006 (71 FR 32603). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated February 15, 

2007. 

Duke Power Company LLC, Docket Nos. 

50–369 and 50–370, McGuire Nuclear 

Station, Units 1 and 2, Mecklenburg 

County, North Carolina 

Date of application for amendments: 

April 11, 2006, as supplemented 

November 29, 2006. 

Brief description of amendments: The 

amendments revised the Technical 

Specifications (TSs) related to steam 

generator tube integrity. The changes are 

consistent with the consolidated lineitem improvement process, Nuclear 

Regulatory Commission’s approved 

Technical Specification Task Force 

(TSTF) Standard Specification Change 

Traveler, TSTF–449, Revision 4, ‘‘Steam 

Generator Tube Integrity.’’

Date of issuance: March 1, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: 237, 218. 

Renewed Facility Operating License 

Nos. NPF–9 and NPF–17: Amendments 

revised the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: December 5, 2006 (71 FR 

70557) The supplement dated 

November 29, 2006, 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 March 1, 2007. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50–

368, Arkansas Nuclear One, Unit No. 2 

(ANO–2), Pope County, Arkansas 

Date of application for amendment: 

March 20, 2006. 

Brief description of amendment: The 

amendment removed ANO–2 reactor 

coolant structural integrity requirements 

contained in TS 3.4.10.1. The TS change 

is consistent with NUREG–1432, 

‘‘Standard Technical Specifications 

Combustion Engineering Plants,’’

Revision 3.1. The Bases for TS 3.4.10.1 

will be deleted and performed under the 

ANO–2 TS Bases Control Program, and 

is not included with the submittal. The 

amendment also renumbers TS pages 3/ 

4 4–22a, 23, 23a, and 23b as TS pages 

3/4 4–23, 24, 25, and 26, respectively. 

Date of issuance: March 1, 2007. 


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Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment No.: 270. 

Renewed Facility Operating License 

No. NPF–6: Amendment revised the 

Technical Specifications/license. 

Date of initial notice in Federal 

Register: May 9, 2006 (71 FR 26999). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated March 1, 2007. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., System Energy 

Resources, Inc., South Mississippi 

Electric Power Association, and Entergy 

Mississippi, Inc., Docket No. 50–416, 

Grand Gulf Nuclear Station, Unit 1, 

Claiborne County, Mississippi 

Date of application for amendment: 

November 1, 2006. 

Brief description of amendment: The 

amendment modified technical 

specification requirements for 

inoperable snubbers by adding Limiting 

Condition of Operation 3.0.8 using the 

Consolidated Line Item Improvement 

Process. 

Date of issuance: February 20, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days of issuance. 

Amendment No: 171. 

Facility Operating License No. NPF–

29: The amendment revises the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: December 5, 2006 (71 FR 

70558). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

February 20, 2007. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., System Energy 

Resources, Inc., South Mississippi 

Electric Power Association, and Entergy 

Mississippi, Inc., Docket No. 50–416, 

Grand Gulf Nuclear Station, Unit 1, 

Claiborne County, Mississippi 

Date of application for amendment: 

November 13, 2006. 

Brief description of amendment: The 

amendment revised Grand Gulf Nuclear 

Station, Unit 1, Technical Specification 

(TS) Limiting Condition of Operation 

3.10.1, and the associated TS Bases, to 

expand its scope to include provisions 

for temperature excursions greater than 

200 °F as a consequence of inservice 

leak and hydrostatic testing, and as a 

consequence of scram time testing 

initiated in conjunction with an 

inservice leak or hydrostatic test, while 

considering operational conditions to be 

in MODE 4. 

Date of issuance: February 21, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days of issuance. 

Amendment No: 172. 

Facility Operating License No. NPF–

29: The amendment revises the 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: December 19, 2006 (71 FR 

75993). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

February 21, 2007. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., System Energy 

Resources, Inc., South Mississippi 

Electric Power Association, and Entergy 

Mississippi, Inc., Docket No. 50–416, 

Grand Gulf Nuclear Station, Unit 1, 

Claiborne County, Mississippi 

Date of application for amendment: 

May 8, 2006, as supplemented by letter 

dated November 16, 2006. 

Brief description of amendment: The 

change added an NRC-approved topical 

report to the analytical methods 

referenced in Technical Specification 

Section 5.6.5, ‘‘Core Operating Limits 

Report (COLR).’’

Date of issuance: February 22, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

prior to Cycle 16 operation. 

Amendment No: 173. 

Facility Operating License No. NPF–

29: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: June 20, 2006 (71 FR 35458). 

The supplement dated November 16, 

2006, provided additional information 

that clarified the application, did not 

expand the scope of the application as 

originally noticed, and did not change 

the staff’s original proposed no 

significant hazards consideration 

determination as published in the 

Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 22, 

2007. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket Nos. 50–247 and 50–286, Indian 

Point Nuclear Generating Unit Nos. 2 

and 3, Westchester County, New York 

Date of application for amendment: 

May 31, 2006, as supplemented by letter 

dated August 30, 2006. 

Brief description of amendment: The 

amendments revise the Technical 

Specifications (TSs) associated with 

steam generator tube integrity consistent 

with Revision 4 to the TS Task Force 

(TSTF) Standard Technical 

Specification Change Document TSTF–

449, ‘‘Steam Generator Tube Integrity.’’

Date of issuance: February 20, 2007. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 251 and 233. 

Facility Operating License Nos. DPR–

26 and DPR–64: The amendment 

revised the License and the TSs. 

Date of initial notice in Federal 

Register: August 1, 2006 (71 FR 43531). 

The August 30, 2006, supplement 

provided 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 no significant 

hazards consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 20, 

2007. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. 50–373 and 50–374, LaSalle 

County Station, Units 1 and 2, LaSalle 

County, Illinois 

Date of application for amendments: 

April 4, 2006. 

Brief description of amendments: The 

amendments add one NRC-approved 

topical report reference to the list of 

analytical methods in Technical 

Specification (TS) Section 5.6.5, ‘‘Core 

Operating Limits Report (COLR),’’ that 

can be used to determine core operating 

limits and delete seven obsolete 

references from the same TS section. 

Date of issuance: February 15, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 181/168. 

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments 

revised the Technical Specifications and 

License. 

Date of initial notice in Federal 

Register: August 15, 2006 (71 FR 

46933). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

February 15, 2007. 

No significant hazards consideration 

comments received: No. 


cprice-sewell on PROD1PC66 with NOTICES
Federal Register / Vol. 72, No. 48 / Tuesday, March 13, 2007 / Notices 11399 

FirstEnergy Nuclear Operating 

Company, et al., Docket Nos. 50–334 

and 50–412, Beaver Valley Power 

Station, Unit Nos. 1 and 2, Beaver 

County, Pennsylvania 

Date of application for amendments: 

February 25, 2005, as supplemented by 

letters dated November 11, 2005, April 

19, July 10, 2006, September 1, October 

24, December 7, 2006, and February 1, 

2007. 

Brief description of amendments: The 

amendment converts the current 

Technical Specifications to the 

Improved Technical Specifications 

(ITSs) format and relocates certain 

requirements to other licenseecontrolled documents. The ITSs are 

based on NUREG–1431, ‘‘Standard 

Technical Specifications—

Westinghouse Plants,’’ Revision 2, with 

the Technical Specification Task Force 

changes to make the Beaver Valley 

Power Station Unit Nos. 1 and 2 (BVPS–

1 and 2) ITS more consistent with 

Revision 3; the Commission’s Final 

Policy Statement, ‘‘NRC Final Policy 

Statement on Technical Specification 

Improvements for Nuclear Power 

Reactors,’’ dated July 22, 1993 (58 FR 

39132); and 10 CFR 50.36, ‘‘Technical 

specifications.’’ The purpose of the 

conversion is to provide clearer and 

more readily understandable 

requirements in the TSs for BVPS–1 and 

2 to ensure safe operation. In addition, 

the amendment includes a number of 

issues that were considered beyond the 

scope of NUREG–1431. 

Date of issuance: February 1, 2007. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 150 days. 

Amendment Nos.: 278 and 161. 

Facility Operating License Nos. DPR–

66 and NPF–73: The amendment 

revised the License and the Technical 

Specifications. 

Date of initial notice in Federal 

Register: March 22, 2006 (71 FR 

14554). The letters dated November 11, 

2005, April 19, July 10, 2006, September 

1, October 24, December 7, 2006, and 

February 1, 2007, supplement provided 

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 no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 1, 

2007. 

No significant hazards consideration 

comments received: No. 

FPL Energy Seabrook, LLC, Docket No. 

50–443, Seabrook Station, Unit No. 1, 

Rockingham County, New Hampshire 

Date of amendment request: April 28, 

2006. 

Description of amendment request: 

The amendment revised the Seabrook 

Technical Specifications (TSs) Limiting 

Condition for Operation 3.0.4 and 

Surveillance Requirement (SR) 4.0.4 to 

adopt the provisions of Industry/TS 

Task Force (TSTF) change TSTF–359, 

Revision 9, ‘‘Increased Flexibility in 

Mode Restraints.’’ TSTF–359 is part of 

the consolidated line item improvement 

process. Specifically, the proposed 

change allows, for systems and 

components, mode changes into a TS 

condition that has a specific required 

action and completion time. 

Date of issuance: February 9, 2007. 

Effective date: As of its date of 

issuance, and shall be implemented 

within 90 days. 

Amendment No.: 114. 

Facility Operating License No. NPF–

86: The amendment revised the TSs. 

Date of initial notice in Federal 

Register: July 5, 2006 (71 FR 38182). 

The Commission’s related evaluation of 

the amendment is contained in a Safety 

Evaluation dated February 9, 2007. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket Nos. 50–266 and 50–301, Point 

Beach Nuclear Plant, Units 1 and 2, 

Town of Two Creeks, Manitowoc 

County, Wisconsin 

Date of application for amendments: 

October 23, 2006. 

Brief description of amendments: The 

amendments to the Technical 

Specifications (TSs) eliminate the use of 

the defined term CORE ALTERATIONS 

in the TSs. 

Date of issuance: February 15, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: 224 & 230. 

Renewed Facility Operating License 

Nos. DPR–24 and DPR–27: Amendments 

revised the Technical Specifications/ 

License. 

Date of initial notice in Federal 

Register: December 5, 2006 (71 FR 

70562). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

February 15, 2007. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket Nos. 50–282 and 50–306, Prairie 

Island Nuclear Generating Plant 

(PINGP), Units 1 and 2, Goodhue 

County, Minnesota 

Date of application for amendments: 

February 13, 2006. 

Brief description of amendments: The 

amendments revise Prairie Island 

Nuclear Generating Plant, Units 1 and 2, 

Technical Specifications (TS) to change 

the wording in TS 3.0, ‘‘Surveillance 

Requirement (SR) Applicability’’ and 

change format and titles in TS 5.0, 

‘‘Administrative Controls.’’ The 

proposed changes improve the TS 

usability, conformance with the 

industry standard, NUREG–1431, 

‘‘Standard Technical Specifications, 

Westinghouse Plants,’’ Revision 3.0 and 

accuracy. 

Date of issuance: February 13, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment Nos.: 176 and 166. 

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: April 11, 2006 (71 FR 18375). 

The Commission’s related evaluation of 

the amendments is contained in a Safety 

Evaluation dated February 13, 2007. 

No significant hazards consideration 

comments received: No. 

Omaha Public Power District, Docket 

No. 50–285, Fort Calhoun Station, Unit 

No. 1, Washington County, Nebraska 

Date of amendment request: 

November 13, 2006. 

Brief description of amendment: The 

amendment relocated the requirements 

of Technical Specification (TS) 2.22, 

‘‘Toxic Gas Monitors,’’ and TS Table 3–

3, Item 29, to the Fort Calhoun Station, 

Unit No. 1, Updated Safety Analysis 

Report. 

Date of issuance: February 28, 2007. 

Effective date: As of its date of 

issuance and shall be implemented 

within 120 days of issuance. 

Amendment No.: 248. 

Renewed Facility Operating License 

No. DPR–40: The amendment revised 

the Technical Specifications. 

Date of initial notice in Federal 

Register: December 19, 2006 (71 FR 

75996). The Commission’s related 

evaluation of the amendment is 

contained in a safety evaluation dated 

February 28, 2007. 

No significant hazards consideration 

comments received: No. 


cprice-sewell on PROD1PC66 with NOTICES
11400 Federal Register / Vol. 72, No. 48 / Tuesday, March 13, 2007 / Notices 

PPL Susquehanna, LLC, Docket No. 50–

387 and 50–388, Susquehanna Steam 

Electric Station, Units 1 and 2 (SSES 1 

and 2), Luzerne County, Pennsylvania 

Date of application for amendments: 

April 28, 2006. 

Brief description of amendments: The 

amendments revise the SSES 1 and 2 

Technical Specifications 3.1.7, 

‘‘Standby Liquid Control (SLC) System,’’

to modify the SLC system for single loop 

pump operation and the use of enriched 

sodium pentaborate solution. 

Date of issuance: February 28, 2007. 

Effective date: As of the date of 

issuance and to be implemented prior to 

the startup following the SSES 1 Spring 

2008 15th refueling outage and SSES 2 

Spring 2007 13th refueling outage for 

Units 1 and 2, respectively. 

Amendment Nos.: 240 and 217. 

Facility Operating License Nos. NPF–

14 and NPF–22: The amendments 

revised the TSs and license. 

Date of initial notice in Federal 

Register: August 15, 2006 (71 FR 

46936). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

February 28, 2007. 

No significant hazards consideration 

comments received: No. 

Tennessee Valley Authority, Docket 

Nos. 50–259 Browns Ferry Nuclear 

Plant, Unit 1, Limestone County, 

Alabama 

Date of application for amendment: 

May 1, 2006 (TS–455), as supplemented 

by letters dated September 1, and 

November 6, 2006. 

Brief description of amendment: The 

amendment revises the numeric values 

of the safety limit critical power ratio 

(SLMCPR) in the Technical 

Specification (TS) Section 2.1.1.2 for 

one and two reactor recirculation loop 

operation to incorporate the results of 

the Cycle 7 SLMCPR analysis. 

Date of issuance: February 6, 2007. 

Effective date: Date of issuance, to be 

implemented within 60 days. 

Amendment No.: 267. 

Facility Operating License Nos. DPR–

33: Amendment revised the TSs. 

Date of initial notice in Federal 

Register: August 15, 2006 (71 FR 

46937). The supplements dated 

September 1, and November 6, 2006, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 6, 

2007. 

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 

22, 2005, as supplemented by letters 

dated September 18 and October 23, 

2006. 

Brief description of amendments: The 

amendments revised the Final Safety 

Evaluation Report Sections 1, 6, and 15. 

The changes reflect the licensee’s 

adoption of Nuclear Regulatory 

Commission’s Regulatory Guide 1.195, 

‘‘Methods and Assumptions for 

Evaluating Radiological Consequences 

of Design Basis Accidents at Light-Water 

Reactors,’’ for calculating radiological 

consequences and replacement of steam 

generators for Comanche Peak Steam 

Electric Station, Unit 1, in the spring of 

2007. 

Date of issuance: February 20, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: 130/130. 

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments 

revised the Final Safety Analysis Report 

and Facility Operating Licenses. 

Date of initial notice in Federal 

Register: November 8, 2005 (70 FR 

67754). The supplements dated 

September 18 and October 23, 2006, 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the staff’s 

original proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated February 20, 

2007. 

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: February 

21, 2006, as supplemented by letters 

dated September 12 and December 14, 

2006. 

Brief description of amendments: The 

amendments increased the allowable 

values (AVs) for steam generator (SG) 

water level trip setpoints and the 

required minimum SG secondary side 

water inventory in shutdown modes for 

the replacement SGs in Comanche Peak 

Steam Electric Station (CPSES), Unit 1. 

For CPSES Unit 2, the corresponding 

AVs and the SG secondary water 

inventory in the current TSs remain 

unchanged since the existing SGs in 

Unit 2 will continue to be used. 

Date of issuance: February 20, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: NPF–87—131; 

NPF–89—131. 

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in Federal 

Register: June 6, 2006 (71 FR 32609). 

The supplements dated September 12 

and December 14, 2006, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated February 20, 

2007. 

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: 

December 16, 2005, as supplemented by 

letters dated August 31 and September 

29, 2006. 

Brief description of amendments: The 

amendments revised Technical 

Specifications (TSs) 1.1 and 5.6.6 

consistent with the Nuclear Regulatory 

Commission (NRC)-approved Technical 

Specification Task Force (TSTF) 

Standard Technical Specification 

Change Traveler, TSTF–419, ‘‘Revise 

PTLR [Pressure Temperature Limits 

Report] Definition and References in 

ISTS [Improved Standard Technical 

Specification] 5.6.6. 

Date of issuance: February 22, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: NPF–87–132 and 

NPF–89–132. 

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in Federal 

Register: March 14, 2006 (71 FR 


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13182). The supplements dated August 

31 and September 29, 2006, provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated February 22, 

2007. 

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: 

December 12, 2005. 

Brief description of amendments: The 

amendments revise the Technical 

Specification (TS) Surveillance 

Requirements (SRs) 3.3.1.2 and 3.3.1.3, 

‘‘Reactor Trip System (RTS) 

Instrumentation.’’ The license 

amendment request is based on 

Technical Specification Task Force 

(TSTF) Traveler, TSTF–371–A, Revision 

1, ‘‘NIS [Nuclear Instrumentation 

System] Power Range Channel Daily SR 

TS Change to Address Low Power 

Decalibration.’’ TSTF–371–A, Revision 

1, revised the requirements for 

performing a daily surveillance 

adjustment of the power range 

channel(s) to address industry concern 

that compliance with SR 3.3.1.2 and SR 

3.3.1.3 may result in a non-conservative 

channel calibration during reducedpower operations. The changes resolved 

the issue of non-conservatism. 

Date of issuance: February 26, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: NPF–87–133, NPF–

89–133. 

Facility Operating License Nos. NPF–

87 and NPF–89: The amendments 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in Federal 

Register: March 28, 2006 (71 FR 

15490). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated February 26, 

2007. 

No significant hazards consideration 

comments received: No. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of application for amendment: 

May 30, 2006, as supplemented by 

letters dated November 22 and 

December 19, 2006. 

Brief description of amendment: The 

amendment revised Surveillance 

Requirements (SRs) 3.5.2.8 and 3.6.7.1 

due to (1) the future replacement of the 

existing containment recirculation sump 

suction inlet trash racks and screens 

with strainers, (2) the resulting 

relocation of the recirculation fluid pH 

control (RFPC) system from the sump, 

and (3) the removal of details from SR 

3.6.7.1, including the relocation of the 

name of the RFPC chemical to a license 

condition in Appendix C to the license. 

The modifications will be done in the 

refueling outage scheduled for the 

spring of 2007. The amendment also 

deleted the footnote to the frequency for 

SR 3.5.2.5 because it is no longer 

applicable. 

Date of issuance: February 21, 2007. 

Effective date: As of its date of 

issuance, and shall be implemented 

prior to entry into Mode 4 during the 

plant startup from the refueling outage 

scheduled for the spring of 2007. 

Amendment No.: 180. 

Facility Operating License No. NPF–

30: The amendment revised the 

Operating License and the Technical 

Specifications. 

Date of initial notice in Federal 

Register: August 15, 2006 (71 FR 

46940). The supplemental letters dated 

November 22 and December 19, 2006, 

did not expand the scope of the 

application as originally noticed, and 

did not change the NRC staff’s original 

proposed no significant hazards 

consideration determination published 

in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated February 21, 

2007. 

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 

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. 


cprice-sewell on PROD1PC66 with NOTICES
11402 Federal Register / Vol. 72, No. 48 / Tuesday, March 13, 2007 / 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. 

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 

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 


cprice-sewell on PROD1PC66 with NOTICES
Federal Register / Vol. 72, No. 48 / Tuesday, March 13, 2007 / Notices 11403 

1Prevention of Certain Unlawful Activities with 

Respect to Registered Investment Companies, 

Investment Company Act Release No. 11421 (Oct. 

31, 1980) (45 FR 73915 (Nov. 7, 1980)). 

2Personal Investment Activities of Investment 

Company Personnel, Investment Company Act 

Release No. 23958 (Aug. 20, 1999) (64 FR 46821–

01 (Aug. 27, 1999)). 

3 Investment Adviser Codes of Ethics, Investment 

Advisers Act Release No. 2256 (Jul. 2, 2004) (66 FR 

41696 (Jul. 9, 2004)). 

4Rule 17j–1(a)(1) defines an ‘‘access person’’ as 

‘‘Any advisory person of a Fund or of a Fund’s 

investment adviser. If an investment adviser’s 

primary business is advising Funds or other 

advisory clients, all of the investment adviser’s 

directors, officers, and general partners are 

presumed to be Access Persons of any Fund advised 

by the investment adviser. All of a Fund’s directors, 

Continued 

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). 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of amendment request: February 

2, 2007. 

Description of amendment request: 

The amendment revised Technical 

Specification 3.6.1.7, ‘‘Suppression 

Chamber-to-Drywell Vacuum Breakers,’’

to allow a one-time extension to the 

current closure verification surveillance 

requirement for one of two redundant 

disks in one of nine vacuum breakers 

until reliable position indication can be 

restored in the main control room 

during the next refueling outage (R–18), 

which is scheduled to begin on May 12, 

2007. 

Date of issuance: February 27, 2007. 

Effective date: As of its date of 

issuance and shall be implemented 

within 14 days from the date of 

issuance. 

Amendment No.: 202. 

Facility Operating License No.: NPF–

21: Amendment revises the technical 

specifications and license. 

Public comments requested as to 

proposed no significant hazards 

consideration (NSHC): Yes. 72 FR 6606, 

published February 12, 2007. The notice 

provided an opportunity to submit 

comments on the Commission’s 

proposed NSHC determination. No 

comments have been received. The 

notice also provided an opportunity to 

request a hearing within 60 days after 

the date of publication of the notice, but 

indicated that if the Commission makes 

a final NSHC determination, any such 

hearing would take place after issuance 

of the amendment. 

The Commission’s related evaluation 

of the amendment, finding of exigent 

circumstances, state consultation, and 

final NSHC determination are contained 

in a safety evaluation dated February 27, 

2007. 

Attorney for licensee: William A. 

Horin, Esq., Winston & Strawn, 1700 K 

Street, NW., Washington, DC 20006–

3817. 

NRC Branch Chief: David Terao. 

Dated at Rockville, Maryland, this 2nd day 

of March 2007. 

For the Nuclear Regulatory Commission. 

Michael C. Cheok, 

Acting Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E7–4251 Filed 3–12–07; 8:45 am] 

BILLING CODE 7590–01–P 

SECURITIES AND EXCHANGE 

COMMISSION 

Proposed Collection; Comment 

Request 

Upon Written Request, Copies Available 

From: Securities and Exchange 

Commission, Office of Filings and 

Information Services, Washington, DC 

20549. 

Extension: 

Rule 17j–1, SEC File No. 270–239, OMB 

Control No. 3235–0224. 

Notice is hereby given that, pursuant 

to the Paperwork Reduction Act of 1995 

(44 U.S.C. 350l–3520), the Securities 

and Exchange Commission (the 

‘‘Commission’’) is soliciting comments 

on the collection of information 

summarized below. The Commission 

plans to submit this existing collection 

of information to the Office of 

Management and Budget (‘‘OMB’’) for 

extension and approval. 

Conflicts of interest between 

investment company personnel (such as 

portfolio managers) and their funds can 

arise when these persons buy and sell 

securities for their own accounts 

(‘‘personal investment activities’’). 

These conflicts arise because fund 

personnel have the opportunity to profit 

from information about fund 

transactions, often to the detriment of 

fund investors. Beginning in the early 

1960s, Congress and the Securities and 

Exchange Commission (‘‘Commission’’) 

sought to devise a regulatory scheme to 

effectively address these potential 

conflicts. These efforts culminated in 

the addition of section 17(j) to the 

Investment Company Act of 1940 (the 

‘‘Investment Company Act’’) (15 U.S.C. 

80a–17(j)) in 1970 and the adoption by 

the Commission of rule 17j–1 (17 CFR 

270.17j–1) in 1980.1 The Commission 

proposed amendments to rule 17j–1 in 

1995 in response to recommendations 

made in the first detailed study of fund 

policies concerning personal investment 

activities by the Commission’s Division 

of Investment Management since rule 

17j–1 was adopted. Amendments to rule 

17j–1, which were adopted in 1999, 

enhanced fund oversight of personal 

investment activities and the board’s 

role in carrying out that oversight.2

Additional amendments to rule 17j–1 

were made in 2004, conforming rule 

17j–1 to rule 204A–1 under the 

Investment Advisers Act of 1940 (15 

U.S.C. 80b), avoiding duplicative 

reporting, and modifying certain 

definitions and time restrictions.3

Section 17(j) makes it unlawful for 

persons affiliated with a registered 

investment company(‘‘fund’’) or with 

the fund’s investment adviser or 

principal underwriter (each a ‘‘17j–1 

organization’’), in connection with the 

purchase or sale of securities held or to 

be acquired by the investment company, 

to engage in any fraudulent, deceptive, 

or manipulative act or practice in 

contravention of the Commission’s rules 

and regulations. Section 17(j) also 

authorizes the Commission to 

promulgate rules requiring 17j–1 

organizations to adopt codes of ethics. 

In order to implement section 17(j), 

rule 17j–1 imposes certain requirements 

on 17j–1 organizations and ‘‘Access 

Persons’’ 4 of those organizations. The 


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