17944 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices 

Dated: April 5, 2007. 

Rochelle C. Bavol, 

Office of the Secretary. 

[FR Doc. 07–1795 Filed 4–6–07; 11:49 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 March 16, 

2007 to March 29, 2007. The last 

biweekly notice was published on 

March 27, 2007 (72 FR 14303). 

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 


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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 

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. 

AmerGen Energy Company, LLC, Docket 

No. 50–461, Clinton Power Station, Unit 

No. 1, DeWitt County, Illinois 

Date of amendment request: 

November 13, 2006. 

Description of amendment request: 

The proposed amendment changes the 

technical specification (TS) testing 

frequency for the surveillance 

requirement (SR) in TS 3.1.4, ‘‘Control 

Rod Scram Times.’’ The proposed 

change revises the test frequency of SR 

3.1.4.2, control rod scram time testing, 

from ‘‘120 days cumulative operation in 

MODE 1’’ to ‘‘200 days cumulative 

operation in Mode 1.’’

AmerGen has reviewed the proposed 

no significant hazards consideration 

determination published in the Federal 

Register on August 23, 2004 (69 FR 

51864), as part of the consolidated line 

item improvement process (CLIIP) and 

has concluded that the proposed 

determination presented in the notice is 

applicable to Clinton Power Station, 

Unit No. 1. 

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. Does the change involve a significant 

increase in the probability or consequences 

of an accident previously evaluated? 

Response: No. 

The proposed change extends the 

frequency for testing control rod scram time 

testing from every 120 days of cumulative 

Mode 1 operation to 200 days of cumulative 

Mode 1 operation. The frequency of 

surveillance testing is not an initiator of any 

accident previously evaluated. The frequency 

of surveillance testing does not affect the 

ability to mitigate any accident previously 

evaluated, as the tested component is still 

required to be operable. Therefore, the 

proposed change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

Response: No. 

The proposed change extends the 

frequency for testing control rod scram time 

testing from every 120 days of cumulative 

Mode 1 operation to 200 days of cumulative 

Mode 1 operation. The proposed change does 

not result in any new or different modes of 

plant operation. 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 proposed change extends the 

frequency for testing control rod scram time 

testing from every 120 days of cumulative 

Mode 1 operation to 200 days of cumulative 

Mode 1 operation. The proposed change 

continues to test the control rod scram time 

to ensure the assumptions in the safety 

analysis are protected. Therefore, the 

proposed change does not involve a 

significant reduction in a margin of safety. 

Based on the above, the proposed 

change 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. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves no significant hazards 

consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Assistant General Counsel, 

Exelon Generation Company, LLC, 200 

Exelon Way, Kennett Square, PA 19348. 

NRC Branch Chief: Russell Gibbs. 

Detroit Edison Company, Docket No. 

50–341, Fermi 2, Monroe County, 

Michigan 

Date of amendment request: January 

26, 2007. 

Description of amendment request: 

The proposed amendment would revise 

technical specifications (TS) 

requirements for unavailable barriers by 


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adding limiting condition for operation 

(LCO) 3.0.9. This would establish 

conditions under which TS systems 

would remain operable when required 

physical barriers are not capable of 

providing their related support function. 

Also, the proposed amendment would 

make editorial changes to LCO 3.0.8 to 

be consistent with the terminology in 

LCO 3.0.9. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration by a reference to a generic 

analysis published in the Federal 

Register on October 3, 2006 (71 FR 

58444), which is presented below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of any accident previously 

evaluated? 

Response: No. 

The proposed change allows a delay time 

for entering a supported system technical 

specification (TS) when the inoperability is 

due solely to an unavailable barrier if risk is 

assessed and managed. The postulated 

initiating events which may require a 

functional barrier are limited to those with 

low frequencies of occurrence, and the 

overall TS system safety function would still 

be available for the majority of anticipated 

challenges. Therefore, the probability of an 

accident previously evaluated is not 

significantly increased, if at all. The 

consequences of an accident while relying on 

the allowance provided by proposed LCO 

3.0.9 are no different than the consequences 

of an accident while relying on the TS 

required actions in effect without the 

allowance provided by proposed LCO 3.0.9. 

Therefore, the consequences of an accident 

previously evaluated are not significantly 

affected by this change. The addition of a 

requirement to assess and manage the risk 

introduced by this change will further 

minimize possible concerns. 

Therefore, this change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

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 alteration of the plant (no new or 

different type of equipment will be installed). 

Allowing delay times for entering supported 

system TS when inoperability is due solely 

to an unavailable barrier, if risk is assessed 

and managed, will not introduce new failure 

modes or effects and will not, in the absence 

of other unrelated failures, lead to an 

accident whose consequences exceed the 

consequences of accidents previously 

evaluated. The addition of a requirement to 

assess and manage the risk introduced by this 

change will further minimize possible 

concerns. Thus, this change does not create 

the possibility of a new or different kind of 

accident from an accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change allows a delay time 

for entering a supported system TS when the 

inoperability is due solely to an unavailable 

barrier, if risk is assessed and managed. The 

postulated initiating events which may 

require a functional barrier are limited to 

those with low frequencies of occurrence, 

and the overall TS system safety function 

would still be available for the majority of 

anticipated challenges. The risk impact of the 

proposed TS changes was assessed following 

the three-tiered approach recommended in 

RG 1.177. A bounding risk assessment was 

performed to justify the proposed TS 

changes. This application of LCO 3.0.9 is 

predicated upon the licensee’s performance 

of a risk assessment and the management of 

plant risk. The net change to the margin of 

safety is insignificant as indicated by the 

anticipated low levels of associated risk 

(ICCDP and ICLERP) as shown in Table 1 of 

Section 3.1.1 in the Safety Evaluation. 

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 G. 

Pettinari, Legal Department, 688 WCB, 

Detroit Edison Company, 2000 2nd 

Avenue, Detroit, Michigan 48226–1279. 

NRC Branch Chief: L. Raghavan. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–286, Indian Point 

Nuclear Generating Unit No. 3, 

Westchester County, New York 

Date of amendment request: January 

18, 2007. 

Description of amendment request: 

The proposed amendment would revise 

the expiration limit for the reactor 

coolant system Pressure/Temperature 

(P/T) limit graphs in Technical 

Specifications (TS); revise the adjusted 

reference temperature for the reactor 

vessel; and revise the Low Temperature 

Overpressure Protection (LTOP) arming 

temperature value specified in TSs. It 

would also make editorial changes in 

the use of inequality signs in TSs 

associated with the LTOP arming 

temperature in order to make them 

consistent. 

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 does not affect the 

accident initiators or mitigation assumptions 

associated with any of the accidents 

previously evaluated. Operating restrictions 

on pressure-temperature conditions for the 

reactor pressure vessel provide assurance 

that reactor vessel integrity will be 

maintained under accident or transient 

conditions. The proposed change uses 

approved criteria and analysis methods to 

update the time period for which the current 

operating limits remain valid. 

The LTOP system performs an automatic 

function by opening relief valves if reactor 

coolant system pressure reaches a 

temperature-dependent limit. The proposed 

change includes establishing a more 

restrictive temperature limit for when this 

system must be in service, to reflect the 

material condition of the reactor vessel at the 

new EFPY limit proposed for the pressuretemperature graphs. The mitigation function 

and capability of the LTOP system is not 

being changed by this request. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No 

There are no new accident initiators being 

introduced by this proposed change. The 

proposed change does not involve 

installation of new plant equipment, 

modification of existing equipment, or 

changes in the way that plant equipment is 

operated. Pressure-temperature operating 

limits depicted by graphs in the technical 

specifications will not be changed and will 

continue to be used by plant operators. A 

change in the LTOP system arming 

temperature will assure that the graphs 

remain valid for the proposed new operating 

period of 27.2 EFPY [effective full power 

years]. 

Therefore, the proposed changes do 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 

Operating limits on pressure and 

temperature conditions for the reactor 

coolant system (RCS) are important to assure 

that the RCS pressure boundary stresses are 

within analyzed limits. Margins of safety are 

inherent in the analysis methods, 

assumptions, and limits specified in 

regulations and guidance documents. The 

proposed change is based on NRC-accepted 

methods, assumptions and limits and 

maintains the required margin of safety. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 


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review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. John Fulton, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc., 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Acting Branch Chief: Douglas V. 

Pickett. 

Entergy Nuclear Operations, Inc., 

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

Point Nuclear Generating Unit Nos. 2 

(IP2) and 3 (IP3), Westchester County, 

New York 

Date of amendment request: March 

13, 2007. 

Description of amendment request: 

The amendment would revise License 

Condition 2.K for IP2 and License 

Condition 2.H for IP3, which require the 

implementation and maintenance of an 

approved Fire Protection Program for 

each unit. 

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 changes are strictly an 

administrative relocation of the specific fire 

protection SER [safety evaluation report] 

references and do not modify any 

requirements of the fire protection programs. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes are strictly an 

administrative relocation of the specific fire 

protection SER references and do not modify 

any requirements of the fire protection 

programs. 

Therefore, the proposed changes do 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 proposed changes are strictly an 

administrative relocation of the specific fire 

protection SER references and do not modify 

any requirements of the fire protection 

programs. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. John Fulton, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc., 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Acting Branch Chief: Douglas V. 

Pickett. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–247, Indian Point 

Nuclear Generating Unit No. 2, 

Westchester County, New York 

Date of amendment request: March 

22, 2007. 

Description of amendment request: 

The proposed amendment will revise 

the test acceptance criteria specified in 

Technical Specification Surveillance 

Requirement (SR) 3.8.1.10 for the diesel 

generator endurance test. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

No. The proposed change revises the 

acceptance criteria to be applied to an 

existing surveillance test of the facility 

emergency diesel generators (DGs). 

Performing a surveillance test is not an 

accident initiator and does not increase the 

probability of an accident occurring. The 

proposed new acceptance criteria will assure 

that the DGs are capable of carrying the peak 

electrical loading assumed in the various 

existing safety analyses which take credit for 

the operation of the DGs. Establishing 

acceptance criteria that bound existing 

analyses validates the related assumption 

used in those analyses regarding the 

capability of equipment to mitigate accident 

conditions. Therefore the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

No. The proposed change revises the test 

acceptance criteria for a specific performance 

test conducted on the existing DGs. The 

proposed change does not involve 

installation of new equipment or 

modification of existing equipment, so no 

new equipment failure modes are introduced. 

The proposed revision to the DG surveillance 

test acceptance criteria also is not a change 

to the way that the equipment or facility is 

operated and no new accident initiators are 

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? 

No. The conduct of performance tests on 

safety-related plant equipment is a means of 

assuring that the equipment is capable of 

maintaining the margin of safety established 

in the safety analyses for the facility. The 

proposed change in the DG technical 

specification surveillance test acceptance 

criteria is consistent with values assumed in 

existing safety analyses and is consistent 

with the design rating of the DGs. Therefore 

the proposed change does not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. John Fulton, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc., 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Acting Branch Chief: Douglas V. 

Pickett. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–333, James A. FitzPatrick 

Nuclear Power Plant, Oswego County, 

New York 

Date of amendment request: February 

15, 2007. 

Description of amendment request: 

The proposed changes would revise 

Technical Specification (TS) 3.10.1 to 

expand its scope to include provisions 

for reactor coolant temperature 

excursions greater than 212 °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, which is defined to be 

reactor coolant temperature less than or 

equal to 212 °F. 

This change was proposed by the 

industry’s TS Task Force (TSTF) and is 

designated TSTF–484. The NRC staff 

issued a notice of opportunity for 

comment in the Federal Register on 

August 21, 2006 (71 FR 48561), on 

possible amendments concerning 

TSTF–484, including a model safety 

evaluation and model no significant 

hazards (NSHC) determination, using 

the consolidated line item improvement 

process (CLIIP). The NRC staff 

subsequently issued a notice of 

availability of the models for referencing 

in license amendment applications in 

the Federal Register on October 27, 


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2006 (71 FR 63050). The licensee 

affirmed the applicability of the 

following NSHC determination in its 

application dated February 15, 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 °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 °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 °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. 

Based on the above, the NRC staff 

concludes that the proposed change 

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. 

Attorney for licensee: Mr. John Fulton, 

Assistant General Counsel, Entergy 

Nuclear Operations, Inc., 440 Hamilton 

Avenue, White Plains, NY 10601. 

NRC Acting Branch Chief: Douglas V. 

Pickett. 

Exelon Generation Company, LLC 

(EGC), Docket Nos. 50–373 and 50–374, 

LaSalle County Station (LSCS), Units 1 

and 2, LaSalle County, Illinois 

Date of amendment request: 

November 17, 2006. 

Description of amendment request: 

The proposed amendments would 

replace references to Section XI of the 

American Society of Mechanical 

Engineers (ASME) Boiler and Pressure 

Vessel Code (Code) with a reference to 

the ASME Code of Operation and 

Maintenance of Nuclear Power Plants 

(OM Code) in Technical Specification 

(TS) 5.5.7, ‘‘Inservice Testing Program 

[IST].’’ These proposed changes are 

consistent with the implementation of 

the LSCS, Units 1 and 2 third 10-year 

IST program in accordance with the 

requirements of Title 10 of the Code of 

Federal Regulations (10 CFR) Section 

50.55a, ‘‘Codes and standards,’’

paragraph (f), ‘‘Inservice testing 

requirements.’’ The third 10-year 

interval for LSCS, Units 1 and 2 is 

scheduled to start on October 12, 2007. 

In addition to the replacement of the 

references, EGC is also adding 

provisions in TS 5.5.7, item b, to only 

apply Surveillance Requirement (SR) 

3.0.2 to those inservice testing 

frequencies of two years or less. These 

proposed changes are based on TS Task 

Force (TSTF) Traveler No. 479–A 

(TSTF–479–A), Revision 0, ‘‘Changes to 

Reflect Revision of 10 CFR 50.55a,’’ as 

modified by TSTF–497, Revision 0, 

‘‘Limit Inservice Testing Program SR 

3.0.2 Application to Frequencies of 2 

Years or Less’’ and approved by the 

NRC in December 6, 2005, and October 

4, 2006. 

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 change involve a significant 

increase in the probability or consequences 

of an accident previously evaluated? 

Response: No. 

The proposed changes revise TS 5.5.7 for 

LSCS Units 1 and 2 to conform to the 

requirements of 10 CFR 50.55a, ‘‘Codes and 

standards,’’ paragraph (f) regarding the 

inservice testing of pumps and calves for the 

Third 10-year Interval. The current TS 

reference the [American Society of 

Mechanical Engineers] ASME Boiler and 

Pressure Vessel Code, Section XI, 

requirements for the inservice testing of 

ASME Code Class 1, 2, and 3 pumps and 

valves. The proposed changes would 

reference the ASME OM Code, which is 

consistent with 10 CFR 50.55a, paragraph (f), 

‘‘Inservice testing requirements,’’ and 

approved for use by the NRC. In addition, 

provisions modifying TS 5.5.7, item b, clarify 

that SR 3.0.2 is only applied to those 

inservice testing frequencies of two years or 

less. The definitions of the frequencies are 

not changed by this license amendment 

request. 

The proposed changes are administrative 

in nature, do not affect any accident 

initiators, do not affect the ability of LSCS to 

successfully respond to previously evaluated 

accidents and do not affect radiological 

assumptions used in the evaluations. Thus, 

the radiological consequences of any 

accident previously evaluated are not 

increased. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the change create the possibility of 

a new or different kind of accident from any 

accident previously evaluated? 

Response: No. 

The proposed changes revise TS 5.5.7 for 

LSCS Units 1 and 2 to conform to the 

requirements of 10 CFR 50.55a(f) regarding 

the inservice testing of pumps and valves for 

the Third 10-year Interval. The current TS 

reference the ASME Boiler and Pressure 

Vessel Code, Section XI, requirements for the 

inservice testing of ASME Code Class 1, 2, 

and 3 pumps and valves. The proposed 

changes would reference the ASME OM 

Code, which is consistent with the 10 CFR 

50.55a(f) and approved for use by the NRC. 

In addition, provisions modifying TS 5.5.7, 

item b, clarify that SR 3.0.2 is only applied 

to those inservice testing frequencies of two 

years or less. The definitions of the 

frequencies are not changed by this license 

amendment request. 

The proposed changes to TS Section 5.5.7 

do not affect the performance of any LSCS 

structure, system, or component credited 

with mitigating any accident previously 

evaluated and do not introduce any new 

modes of system operation or failure 

mechanisms. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the change involve a significant 

reduction in a margin of safety? 

Response: No. 

The proposed changes revise TS 5.5.7 for 

LSCS Units 1 and 2 to conform to the 

requirements of 10 CFR 50.55a(f) regarding 

the inservice testing of pumps and valves for 

the Third 10-Year Interval. The current TS 

reference the ASME Boiler and Pressure 

Vessel Code, Section XI, requirements for the 

inservice testing of ASME Code Class 1, 2, 


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and 3 pumps and valves. The proposed 

changes would reference the ASME OM 

Code, which is consistent with the 10 CFR 

50.55a(f) and approved for use by the NRC. 

In addition, provisions modifying TS 5.5.7, 

item b, clarify that SR 3.0.2 is only applied 

to those inservice testing frequencies of two 

years or less. The definitions of the 

frequencies are not changed by this license 

amendment request. 

The proposed changes do not modify the 

safety limits setpoints at which protective 

actions are initiated and do not change the 

requirements governing operation or 

availability of safety equipment assumed to 

operate to preserve the margin of safety. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

requested amendments involve no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Assistant General Counsel, 

Exelon Generation Company, LLC, 200 

Exelon Way, Kennett Square, PA 19348. 

NRC Branch Chief: Russell Gibbs. 

Exelon Generation Company, LLC, 

Docket Nos. 50–352 and 50–353, 

Limerick Generating Station, Units 1 

and 2, Montgomery County, 

Pennsylvania 

Date of amendment request: June 2, 

2006. 

Description of amendment request: 

The proposed amendments incorporates 

revised 10 CFR Part 20 requirements for 

Limerick Generating Station Units 1 and 

2 technical specifications (TSs). 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

Updating the Technical Specifications (TS) 

to be consistent with 10 CFR Part 20 has no 

impact on plant structures, systems, or 

components, does not affect any accident 

initiators, and does not change any safety 

analysis. Therefore, the proposed changes do 

not involve an 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. 

Updating the TS to be consistent with 10 

CFR Part 20 will not change any equipment, 

require new equipment to be installed, or 

change the way current equipment operates. 

No credible new failure mechanisms, 

malfunctions, or accident initiators are 

created by the proposed changes. 

Therefore, the 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. 

Updating the TS to be consistent with 10 

CFR Part 20 does not adversely affect existing 

plant safety margins or the reliability of 

equipment assumed to operate in the safety 

analysis. As such, there are no changes being 

made to safety analysis assumptions, safety 

limits or limiting safety system settings that 

would adversely affect plant safety as a result 

of the proposed changes. Therefore, the 

proposed changes do not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Brad 

Fewell, Assistant General Counsel, 

Exelon Generation Company, LLC, 200 

Exelon Way, Kennett Square, PA 19348. 

NRC Branch Chief: Harold K. 

Chernoff. 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–346, 

Davis-Besse Nuclear Power Station, Unit 

No. 1, Ottawa County, Ohio 

Date of amendment request: February 

12, 2007. 

Description of amendment request: 

The proposed license amendment 

would revise Technical Specification 

(TS) Limiting Condition for Operation 

3.9.4, ‘‘Containment Penetrations’’, to 

allow penetrations included under TS 

3.9.4(c) to be opened during core 

alterations or movement of irradiated 

fuel, under administrative controls. This 

change is based on the TS Task Force 

Traveler No. 312–A, Revision 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 amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change would allow 

containment penetrations identified under 

Technical Specification 3.9.4(c) to remain 

open during fuel movement and core 

alterations. These penetrations are normally 

closed during this time period to prevent the 

release of radioactive material in the event of 

a Fuel Handling Accident inside 

containment. These penetrations are not 

initiators of any accident. The probability of 

a Fuel Handling Accident is unaffected by 

the status of these penetrations. 

The Fuel Handling Accident analyses 

demonstrate that the maximum offsite dose is 

well [within] the acceptance limits specified 

in SRP [Standard Review Plan] 15.7.4, and 

the control room dose is within the 

acceptance criteria specified in GDC [General 

Design Criterion] 19. Furthermore, the 

existing analysis results are independent of 

the containment release path, and therefore 

are unaffected by the proposed change. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change does not involve the 

addition or modification of any plant 

equipment. Also, the proposed change will 

not alter the design, configuration, or method 

of operation of the plant beyond the standard 

functional capabilities of the equipment. The 

proposed change involves a Technical 

Specification change that will allow 

containment penetrations identified under 

Technical Specification 3.9.4(c) to remain 

open during fuel movement and core 

alterations. Open penetrations are not 

accident initiators, and will not create the 

possibility of a new kind of accident. 

Administrative controls will be implemented 

to ensure the capability to close the affected 

containment penetrations in the event of a 

Fuel Handling Accident. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change has the potential to 

slightly increase the post-Fuel Handling 

Accident dose at the site boundary and in the 

control room. However, the existing analyses 

take no credit for containment of the release, 

so that the existing analysis results will 

remain bounding. 

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, Mail Stop A–GO–18, 76 

South Main Street, Akron, OH 44308. 

NRC Branch Chief: Russell Gibbs. 


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17950 Federal Register / Vol. 72, No. 68 / Tuesday, April 10, 2007 / Notices 

FirstEnergy Nuclear Operating 

Company, et al., Docket No. 50–440, 

Perry Nuclear Power Plant, Unit No. 1, 

Lake County, Ohio 

Date of amendment request: January 

19, 2007. 

Description of amendment request: 

The proposed amendment would 

modify Technical Specification (TS) 

5.5.9, ‘‘Diesel Fuel Oil Testing 

Program,’’ by relocating a reference to a 

specific American Society for Testing 

and Materials (ASTM) international 

standard for fuel oil testing to licenseecontrolled documents, and by adding an 

alternate criteria to the ‘‘clear and 

bright’’ acceptance test for new fuel oil, 

per the consolidated line item 

improvement process (CLIIP). 

The U.S. Nuclear Regulatory 

Commission (NRC) staff issued a notice 

of opportunity for comment in the 

Federal Register on February 22, 2006 

(71 FR 9179), on possible amendments 

concerning the CLIIP, including a model 

safety evaluation and a model no 

significant hazards consideration 

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 21, 2006 

(71 FR 20735), as part of the CLIIP. 

In its application dated January 19, 

2007, 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. Does the proposed change involve a 

significant increase in the probability or 

consequences of any accident previously 

evaluated? 

Response: No. 

The proposed changes relocate the specific 

ASTM standard references from the 

Administrative Controls Section of TS to a 

licensee-controlled document. Requirements 

to perform testing in accordance with 

applicable ASTM standards are retained in 

the TS as are requirements to perform 

surveillances of both new and stored diesel 

fuel oil. Future changes to the licenseecontrolled document will be evaluated 

pursuant to the requirements of 10 CFR 

50.59, ‘‘Changes, tests and experiments,’’ to 

ensure that such changes do not result in 

more than a minimal increase in the 

probability or consequences of an accident 

previously evaluated. In addition, the ‘‘clear 

and bright’’ test used to establish the 

acceptability of new fuel oil for use prior to 

addition to storage tanks has been expanded 

to recognize more rigorous testing of water 

and sediment content. Relocating the specific 

ASTM standard references from the TS to a 

licensee-controlled document and allowing a 

water and sediment content test to be 

performed to establish the acceptability of 

new fuel oil will not affect nor degrade the 

ability of the emergency diesel generators 

(DGs) to perform their specified safety 

function. Fuel oil quality will continue to 

meet ASTM requirements. 

The proposed changes do not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, and 

configuration of the facility or the manner in 

which the plant is operated and maintained. 

The proposed changes do not adversely affect 

the ability of structures, systems, and 

components (SSCs) to perform their intended 

safety function to mitigate the consequences 

of an initiating event within the assumed 

acceptance limits. The proposed changes do 

not affect the source term, containment 

isolation, or radiological release assumptions 

used in evaluating the radiological 

consequences of any accident previously 

evaluated. Further, the proposed changes do 

not increase the types and amounts of 

radioactive effluent that may be released 

offsite, nor significantly increase individual 

or cumulative occupational/public radiation 

exposures. 

Therefore, the changes do not involve a 

significant increase in the probability or 

consequences of any 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 changes relocate the specific 

ASTM standard references from the 

Administrative Controls Section of TS to a 

licensee-controlled document. In addition, 

the ‘‘clear and bright’’ test used to establish 

the acceptability of new fuel oil for use prior 

to addition to storage tanks has been 

expanded to allow a water and sediment 

content test to be performed to establish the 

acceptability of new fuel oil. The changes do 

not involve a physical alteration of the plant 

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

will be installed) or a change in the methods 

governing normal plant operation. The 

requirements retained in the TS continue to 

require testing of the diesel fuel oil to ensure 

the proper functioning of the DGs. 

Therefore, the changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes relocate the specific 

ASTM standard references from the 

Administrative Controls Section of TS to a 

licensee-controlled document. Instituting the 

proposed changes will continue to ensure the 

use of applicable ASTM standards to 

evaluate the quality of both new and stored 

fuel oil designated for use in the emergency 

DGs. 

Changes to the licensee-controlled 

document are performed in accordance with 

the provisions of 10 CFR 50.59. This 

approach provides an effective level of 

regulatory control and ensures that diesel 

fuel oil testing is conducted such that there 

is no significant reduction in a margin of 

safety. 

The ‘‘clear and bright’’ test used to 

establish the acceptability of new fuel oil for 

use prior to addition to storage tanks has 

been expanded to allow a water and 

sediment content test to be performed to 

establish the acceptability of new fuel oil. 

The margin of safety provided by the DGs is 

unaffected by the proposed changes since 

there continue to be TS requirements to 

ensure fuel oil is of the appropriate quality 

for emergency DG use. The proposed changes 

provide the flexibility needed to improve fuel 

oil sampling and analysis methodologies 

while maintaining sufficient controls to 

preserve the current margins 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: Russell A. Gibbs 

Florida Power Corporation, et al., 

Docket No. 50–302, Crystal River Unit 3 

Nuclear Generating Plant (CR–3), Citrus 

County, Florida 

Date of amendment request: 

December 14, 2006, as supplemented by 

letter dated March 14, 2007. 

Description of amendment request: 

The proposed amendment would 

modify the technical specification (TS) 

requirements for inoperable snubbers by 

adding Limiting Condition for 

Operation (LCO) 3.0.8. The changes are 

consistent with NRC approved Industry/ 

Technical Specification Task Force 

(TSTF) standard TS change TSTF–372, 

Revision 4. 

The proposed amendment includes an 

administrative change to LCO 3.0.1 that 

will clarify that LCO 3.0.7 allows 

specified TS requirements to be 

suspended during physics tests 

performed in accordance with TSs 3.1.8 

and 3.1.9. This administrative change 

will make the CR–3 TSs more consistent 

with the standard TSs and with TSTF–

372, Revision 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 

May 4, 2005 (70 FR 23252). The licensee 

affirmed the applicability of the model 

NSHC determination in its application 

dated April 26, 2006. 

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: 


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1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change allows a delay time 

before declaring supported TS systems 

inoperable when the associated snubber(s) 

cannot perform its required safety function. 

Entrance into Actions or delaying entrance 

into Actions is not an initiator of any 

accident previously evaluated. Consequently, 

the probability of an accident previously 

evaluated is not significantly increased. The 

consequences of an accident while relying on 

the delay time allowed before declaring a TS 

supported system inoperable and taking its 

Conditions and Required Actions are no 

different than the consequences of an 

accident under the same plant conditions 

while relying on the existing TS supported 

system Conditions and Required Actions. 

Therefore, the consequences of an accident 

previously evaluated are not significantly 

increased by this change. Therefore, this 

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 allows a delay time 

before declaring supported TS systems 

inoperable when the associated snubber(s) 

cannot perform its required safety function. 

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 operations. Thus, this 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 allows a delay time 

before declaring supported TS systems 

inoperable when the associated snubber(s) 

cannot perform its required safety function. 

The proposed change restores an allowance 

in the pre-ISTS [improved Standard 

Technical Specifications] conversion TS that 

was unintentionally eliminated by the 

conversion. The pre-ISTS TSs were 

considered to provide an adequate margin of 

safety for plant operation, as does the postISTS conversion TS. Therefore, this 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: 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: Thomas H. Boyce. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of amendment request: March 

16, 2007. 

Description of amendment request: 

The proposed amendment would add 

new Technical Specification (TS) 

requirements for the response times 

associated with a steam generator 

feedwater pump (SGFP) trip and 

feedwater isolation valve (FIV) closure. 

The amendment would also revise the 

TS requirements for the containment fan 

cooler unit (CFCU) cooling water flow 

rate. These changes are associated with 

a revised containment response analysis 

that credits a SGFP trip and FIV closure 

(on a feedwater regulator valve failure) 

to reduce the mass/energy release to the 

containment during a main steam line 

break (MSLB). The containment analysis 

also credits a reduced heat removal 

capability for the CFCUs, allowing a 

reduction in the required service water 

(SW) flow to the CFCUs. 

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 does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. The proposed change establishes 

response time requirements for feedwater 

isolation and reduced CFCU flow rates to 

support containment analyses to 

accommodate reduced CFCU heat removal 

capacity. The changes in analysis input 

assumptions affect plant response to an 

accident and are not accident initiators; 

therefore, they have no bearing on the 

probability of an accident. The Salem FSAR 

[Final Safety Analysis Report] Chapter 15 

accidents which are impacted by a change in 

the CFCU modeling parameters are LOCA 

[loss-of-coolant accident] and MSLB mass 

and energy release Containment analyses. 

The consequences of these postulated 

accidents are shown to be acceptable using 

assumptions consistent with the proposed 

changes. 

For the LOCA transients, the containment 

cooling systems are considered for three 

aspects: core response, containment response 

and dose. The core response is most limiting 

when the containment conditions minimize 

back pressure since this increases the 

blowdown and reduces the effectiveness of 

the ECCS [emergency core cooling system]. 

The LOCA core response (10 CFR 50.46 

[Section 50.46 of Title 10 of the Code of 

Federal Regulations]—PCT [peak cladding 

temperature]) is conservatively biased to 

minimize the containment backpressure such 

that any safety injection effectiveness is 

minimized (the core becomes the highest 

resistance flow path). Thus, any reduction in 

the accident capability of the CFCUs has no 

bearing on the LOCA core response. 

The bounding containment integrity 

analyses are the LBLOCA [large-break LOCA] 

and the MSLB Inside Containment events. 

The containment integrity analysis relies on 

two heat removal paths to maintain 

containment pressure and temperature 

conditions. The CFCU air-to-water heat 

exchangers reject containment energy to the 

SW System and the Containment Spray 

System removes containment energy by using 

spray droplet direct contact heat exchange to 

transfer the energy from the containment 

ambient to the containment sump, where it 

is transferred out of containment via the RHR 

[residual heat removal] heat exchanger and 

CCW [component cooling water]/SW 

Systems. Containment integrity analyses for 

both LOCA and MSLB, using input 

assumptions consistent with the proposed 

changes, show that containment integrity is 

maintained with reduced CFCU heat removal 

capacity. 

The potential dose impacts due to reduced 

CFCU heat removal capacity are bounded as 

the design basis assumptions concerning the 

number of operating CFCUs (three of five), 

and the thermal-hydraulic transient 

operation of the Containment Spray System 

are unchanged. The Salem design basis only 

credits Containment Spray iodine removal 

effectiveness during the LOCA injection and 

recirculation phases based on a single failure 

of an entire ESF [engineered safety features] 

train. This assumption results in 3 of 5 

CFCUs being available to ensure adequate 

mixing of the containment ambient air as 

well as operation of a single Containment 

Spray Train, which controls containment 

spray droplet size and pH, as described in 

UFSAR [updated FSAR] Section 6.2.3. As a 

further conservatism, the current LOCA 

Alternate Source Term (AST) analysis 

(Calculation S–C–ZZ–MDC–1945, an interim 

revision of which was sent to the NRC 

[Nuclear Regulatory Commission] staff for 

review via letter dated September 16, 2004) 

only credits two CFCUs for mixing. The 

Containment Building and Auxiliary 

Building leakage rates are unaffected by the 

revised containment analysis as the peak 

containment pressure and temperatures are 

less than the design basis values described in 

the Salem UFSAR. Therefore, there is no 

impact on offsite dose rates due to the 

reduced CFCU heat removal capacity. 

One other high energy line break for 

consideration is the rupture of a feedwater 

line break. From a containment response 

aspect, this event is bounded by the MSLB 

event, so it is not explicitly analyzed (or even 

discussed in the Salem UFSAR). 

A review of the Salem design basis for AST 

dose calculations shows that the revised 

Containment Integrity Analysis, WCAP–

16503, does not challenge any of the 

assumptions that are part of the AST design 

basis. 

Section 6.2 of the UFSAR indicates that the 

Appendix J Type A containment leak rate test 


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pressure is based on the containment design 

pressure of 47.0 psig, not the calculated 

accident pressure. Since the design pressure 

value bounds the peak pressure calculated in 

WCAP–16503 and is not being changed, the 

Appendix J testing requirements are not 

impacted. 

Thus, in conclusion, the proposed change 

does not involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. The proposed change 

modifies response time requirements for 

feedwater isolation, and reduces CFCU flow 

rates and heat removal requirements 

consistent with the new containment 

analysis. 

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 create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. The proposed changes support 

revised containment analysis to 

accommodate the reduced CFCU heat 

removal capacity. 

The response time-related changes impose 

new surveillance acceptance criteria to 

existing plant equipment that actuates to 

isolate feedwater following a safety injection 

signal. There is no change in actuation logic 

associated with the addition of response time 

criteria; therefore no new accident sequences 

would result from the imposition of response 

time test criteria to existing plant equipment. 

The reduction in minimum service water 

system flow to the CFCUs is supported by 

analyses demonstrating acceptable system 

performance and containment integrity 

following a demand for system operation. 

The post-accident conditions resulting from 

the proposed reduction in flow do not 

adversely impact the environmental 

qualification of equipment, such that no new 

consequential failures are introduced to any 

design basis accident scenario. CFCU 

operation with the proposed reduction in 

minimum required accident flow would not 

result in the progression of any design basis 

event into a previously unanalyzed accident. 

Therefore, no new accident scenarios are 

created from the CFCU flowrate reduction. 

3. Does the proposed change involve a 

significant reduction in [a] margin of safety? 

Response: No. 

The proposed change does not involve a 

significant reduction in the margin of safety. 

The revised containment analyses 

accommodate reduced CFCU heat removal 

capacity using input assumptions consistent 

with the proposed changes. 

The proposed change involves the addition 

of feedwater isolation response time 

surveillance criteria and reduction in 

minimum service water system flows to 

CFCUs. These changes affect input to the 

analyses of mass/energy releases and 

containment response to a design basis main 

steam line break or loss of coolant accident. 

The analyses, consistent with the proposed 

changes, demonstrate that the acceptance 

criteria continue to be met, and the postaccident conditions do not adversely affect 

containment integrity or otherwise challenge 

any safety limit. The margin of safety with 

respect to containment pressure is preserved 

by demonstrating that the calculated 

pressures do not exceed the design limit of 

47 psig. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Jeffrie J. Keenan, 

Esquire, Nuclear Business Unit—N21, 

P.O. Box 236, Hancocks Bridge, NJ 

08038. 

NRC Branch Chief: Harold K. 

Chernoff. 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Units 1 and 

2, Somervell County, Texas 

Date of amendment request: 

December 19, 2006. 

Brief description of amendments: The 

amendments requested would revise 

Technical Specifications (TS) 

requirement 3.7.5, ‘‘Auxiliary Feedwater 

(AFW) System,’’ TS 3.8.1, ‘‘AC 

Sources—Operating,’’ TS 3.8.9, 

‘‘Distribution Systems—Operating,’’ and 

TS Example 1.3–3. 

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. D[o] the proposed change[s] involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed changes eliminate certain 

Completion Times from the Technical 

Specifications. Completion Times are not an 

initiator to any accident previously 

evaluated. As a result, the probability of an 

accident previously evaluated is not affected. 

The consequences of an accident during the 

revised Completion Time are no different 

than the consequences of the same accident 

during the existing Completion Times. As a 

result, the consequences of an accident 

previously evaluated are not affected by this 

change. The proposed changes do not alter or 

prevent the ability of structures, systems, and 

components from performing their intended 

function to mitigate the consequences of an 

initiating event within the assumed 

acceptance limits. The proposed changes do 

not affect the source term, containment 

isolation, or radiological release assumptions 

used in evaluating the radiological 

consequences of an accident previously 

evaluated. Further, the proposed changes do 

not increase the types or amounts of 

radioactive effluent that may be released 

offsite, nor significantly increase individual 

or cumulative occupational/public radiation 

exposures. The proposed changes are 

consistent with the safety analysis 

assumptions and resultant consequences. 

Therefore, the proposed change[s] d[o] not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. D[o] the proposed change[s] create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The changes do not involve a physical 

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

different type of equipment will be installed) 

or a change in the methods governing normal 

plant operation. The changes do not alter any 

assumptions made in the safety analysis. 

Therefore, the proposed change[s] d[o] not 

create the possibility of a new or different 

accident from any accident previously 

evaluated. 

3. D[o] the proposed change[s] involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed change to delete the second 

Completion Time does not alter the manner 

in which safety limits, limiting safety system 

settings or limiting conditions for operation 

are determined. The safety analysis 

acceptance criteria are not affected by this 

change. The proposed changes will not result 

in plant operation in a configuration outside 

of the design basis. 

Therefore, the proposed change[s] d[o] 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: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Branch Chief: David Terao 

TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station, Units 1 and 

2, Somervell County, Texas 

Date of amendment request: January 

18, 2007. 

Brief description of amendments: The 

amendments requested would revise 

Technical Specifications (TS) 

requirement 3.8.1, ‘‘AC Sources—

Operating,’’ Extension of Completion 

Times for Diesel Generators. 

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: 


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1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed Technical Specification (TS) 

changes do not significantly increase the 

probability of occurrence of a previously 

evaluated accident because the Diesel 

Generators (DGs) are not initiators of 

previously evaluated accidents involving a 

loss of offsite power (LOOP). The proposed 

changes to the TS Required Actions and 

Completion Times (CT) do not affect any of 

the assumptions used in the deterministic or 

the Probabilistic Safety Assessment (PSA) 

analysis. Implementation of the proposed 

changes does not result in a risk significant 

impact. The onsite AC [alternating current] 

power sources will remain highly reliable 

and the proposed changes will not result in 

a significant increase in the risk of plant 

operation. This is demonstrated by showing 

that the impact on plant safety as measured 

by the increase in core damage frequency 

(CDF) is less than 1E–06 per year and the 

increase in large early release frequency 

(LERF) is less than 1E–07 per year. In 

addition, for the CT changes, the incremental 

conditional core damage probabilities 

(ICCDP) and incremental conditional large 

early release probabilities (ICLERP) are less 

than 5E–07 and 5E–08, respectively. These 

changes meet the acceptance criteria in 

Regulatory Guides 1.174 and 1.177. 

Therefore, since the onsite AC power sources 

will continue to perform their functions with 

high reliability as originally assumed and the 

increase in risk as measured by DCDF, 

DLERF, ICCDP, and ICLERP risk metrics is 

within the acceptance criteria of existing 

regulatory guidance, there will not be a 

significant increase in the consequences of 

any accidents. 

The proposed changes do not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, or 

configuration of the facility or the manner in 

which the plant is operated and maintained. 

The proposed changes do not alter or 

prevent the ability of structures, systems, and 

components (SSCs) from performing their 

intended function to mitigate the 

consequences of an initiating event within 

the assumed acceptance limits. The proposed 

changes do not affect the source term, 

containment isolation, or radiological release 

assumptions used in evaluating the 

radiological consequences of an accident 

previously evaluated. The proposed changes 

are consistent with safety analysis 

assumptions and resultant consequences. 

The proposed TS changes will continue to 

ensure the DGs perform their function when 

called upon. Extending the TS CT to 14 days, 

when an AACPS [alternate AC power source] 

is available, does not affect the design, the 

operational characteristics, the function, or 

the reliability of the DGs. Additionally, the 

CT extension to 14 days does not affect the 

interfaces between the DGs and other plant 

systems. Conversely, in the absence of an 

AACPS, the DG 72-hour CT will be applied. 

The availability of the onsite AC power 

system to perform its accident mitigation 

function is not affected by the proposed 

activity and thus there is no impact to the 

radiological consequences of any accident 

analysis. 

To fully evaluate the effect of the changes 

to the CT, PSA methods were utilized. The 

results of this analysis show no significant 

increase in the CDF and LERF. 

The Configuration Risk Management 

Program (CRMP) in TS 5.5.18 is an 

administrative program that assesses risk 

based on plant status. The risk-informed CT 

will be implemented consistent with the 

CRMP and approved plant procedures. When 

utilizing the 14-day extension, requirements 

of the CRMP per TS 5.5.18 call for the 

consideration of other measures to mitigate 

the consequences of an accident occurring 

while a DG is inoperable. Furthermore, 

administrative controls will be applied when 

exercising the 14-day CT extension and are 

adequate to maintain defense-in-depth and 

sufficient safety margins. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed changes do not result in a 

change in the manner in which the electrical 

distribution subsystems provide plant 

protection. The changes to the CT do not 

change any existing accident scenarios, nor 

create any new or different accident 

scenarios. 

In addition, the changes do not impose any 

new or different accident mitigation 

requirements or eliminate any existing 

requirements. 

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 previously 

evaluated. 

3. Do the proposed changes involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed changes do not alter the 

manner in which safety limits, limiting safety 

system settings or limiting conditions for 

operation are determined. Neither the safety 

analyses nor the safety analysis acceptance 

criteria are impacted by these changes. The 

proposed changes will not result in plant 

operation in a configuration outside the 

current design basis. The proposed activities 

only involve changes to certain TS CTs. 

Therefore 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: George L. Edgar, 

Esq., Morgan, Lewis and Bockius, 1800 

M Street, NW., Washington, DC 20036. 

NRC Branch Chief: David Terao. 

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. 

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: March 8, 

2007. 

Brief description of amendment 

request: The proposed amendments 

would revise the McGuire Nuclear 

Station, Units 1 and 2, Technical 

Specification 3.5.2.8, and the associated 

Bases and authorizes changes to the 

Updated Final Safety Analysis Reports 

concerning modifications to the 

emergency core cooling system sump. 

Date of publication of individual 

notice in Federal Register: March 19, 

2007. 

Expiration date of individual notice: 

Comments April 18, 2007; Hearing May 

18, 2007. 

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, 


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and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

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

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, 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. 

Carolina Power & Light Company, et al., 

Docket No. 50–400, Shearon Harris 

Nuclear Power Plant, Unit 1, Wake and 

Chatham Counties, North Carolina 

Date of application for amendment: 

May 23, 2006, as supplemented by 

letters dated October 3, 2006, and 

October 24, 2006. 

Brief description of amendment: This 

amendment revises Technical 

Specification by modifying the steam 

generator tube surveillance program at 

Shearon Harris Nuclear Power Plant, 

Unit 1. 

Date of issuance: March 16, 2007. 

Effective date: This amendment is 

effective as of the date of issuance and 

shall be implemented within 90 days of 

issuance. 

Amendment No. 124. 

Facility Operating License No. NPF–

63: Amendment revises the Technical 

Specifications. 

Date of initial notice in Federal 

Register: December 19, 2006 (71 FR 

75991). The supplemental letters 

provided additional information that 

was within the scope of the initial 

notice and did not change the initial 

proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendment is contained in the 

Safety Evaluation dated: March 16, 

2007. 

No significant hazards consideration 

comments received: No. 

Dominion Nuclear Connecticut, Inc., 

Docket No. 50–423, Millstone Power 

Station, Unit No. 3 New London County, 

Connecticut 

Date of application for amendment: 

July 19, 2006. 

Brief description of amendment: The 

proposed amendment changed the 

Millstone Power Station, Unit No. 3 

(MPS3) reactor core safety limits 

Technical Specification (TS) and 

relocated the reactor core safety limit 

figure to the Core Operating Limits 

Report in the MPS3 Technical 

Requirements Manual. 

Date of issuance: March 14, 2007 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days from the date of 

issuance. 

Amendment No.: 236 

Facility Operating License No. NPF–

49: The amendment revised the TSs. 

Date of initial notice in Federal 

Register: August 29, 2006 (71 FR 

51227). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

March 14, 2007. 

No significant hazards consideration 

comments received: No. 

Duke Power Company LLC, et al., 

Docket Nos. 50–413 and 50–414, 

Catawba Nuclear Station, Units 1 and 2, 

York County, South Carolina 

Date of application for amendments: 

April 11, 2006. 

Brief description of amendments: 

(TSTF–372, Rev. 4) The amendments 

added Technical Specification (TS) 

Limiting Condition for Operation (LCO) 

3.0.8 to allow 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 with an approved Bases 

Control Program that is consistent with 

the TS Bases Control Program described 

in Section 5.5 of the applicable vendor’s 

Standard Technical Specifications. The 

amendment also made an administrative 

change, renumbering existing LCO 3.0.8 

to LCO 3.0.9. 

Date of issuance: March 19, 2007 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: 235, 231 

Renewed Facility Operating License 

Nos. NPF–35 and NPF–52: Amendments 

revised the licenses and the technical 

specifications. 

Date of initial notice in Federal 

Register: December 5, 2006 (71 FR 

70555). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

March 19, 2007. 

No significant hazards consideration 

comments received: No 

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. 

Brief description of amendments: 

(TSTF–372, Rev. 4) The amendments 

added Technical Specification (TS) 

Limiting Condition for Operation (LCO) 

3.0.8 to allow 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 with an approved Bases 

Control Program that is consistent with 

the TS Bases Control Program described 

in Section 5.5 of the applicable vendor’s 

Standard Technical Specifications. The 

amendment also made an administrative 

change, renumbering existing LCO 3.0.8 

to LCO 3.0.9. 

Date of issuance: March 29, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: 238, 220. 

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 

70556). The Commission’s related 

evaluation of the amendments is 

contained in a Safety Evaluation dated 

March 29, 2007. 

No significant hazards consideration 

comments received: No. 

Energy Northwest, Docket No. 50–397, 

Columbia Generating Station, Benton 

County, Washington 

Date of application for amendment: 

May 22, 2006, as supplemented by letter 

dated February 5, 2007. 

Brief description of amendment: The 

amendment revised Technical 

Specification Surveillance 

Requirements 3.8.1.11, 3.8.1.12, 

3.8.1.16, and 3.8.1.19 to eliminate the 

specific test-performance mode 


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restrictions for the High-Pressure Core 

Spray Division 3 diesel generator. 

Date of issuance: March 23, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 45 days from the date of 

issuance. 

Amendment No.: 203. 

Facility Operating License No. NPF–

21: The amendment revised the 

Technical Specifications and Facility 

Operating License. 

Date of initial notice in Federal 

Register: July 18, 2006 (71 FR 40745). 

The supplemental letter dated February 

5, 2007, 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 amendment is contained in a 

Safety Evaluation dated March 23, 2007. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts. 

Date of application for amendment: 

December 27, 2006. 

Brief description of amendment: The 

amendment revised the Technical 

Specification 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 issuance: March 26, 2007. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 226. 

Facility Operating License No. DPR–

35: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: February 20, 2007 (72 FR 

7776). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

March 26, 2007. 

No significant hazards consideration 

comments received: No 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts. 

Date of application for amendment: 

January 15, 2007. 

Brief description of amendment: The 

amendment revised the Technical 

Specifications (TS) to extend the use of 

the current pressure-temperature limits 

as specified in TS Figures 3.6.1, 3.6.2, 

and 3.6.3 through the end of operating 

cycle 18. 

Date of issuance: March 26, 2007. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 227. 

Facility Operating License No. DPR–

35: The amendment revised the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: February 12, 2007 (72 FR 

6609). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

March 26, 2007. 

No significant hazards consideration 

comments received: No. 

Entergy Nuclear Vermont Yankee, LLC 

and Entergy Nuclear Operations, Inc., 

Docket No. 50–271, Vermont Yankee 

Nuclear Power Station, Vernon, 

Vermont 

Date of application for amendment: 

April 22, 2006. 

Brief description of amendment: The 

amendment revised Technical 

Specification (TS) requirements for 

inoperable snubbers by relocating the 

current TS requirements Limiting 

Condition for Operation (LCO) 3.6.I and 

Surveillance Requirement (SR) 4.6.I to 

the Technical Requirements Manual and 

adding LCO 3.0.8 to the TSs. The 

associated TS Bases section has also 

been relocated. 

Date of Issuance: March 26, 2007. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 60 days. 

Amendment No.: 230. 

Facility Operating License No. DPR–

28: The amendment revised the License 

and TSs. 

Date of initial notice in Federal 

Register: June 6, 2006 (71 FR 32604). 

The Commission’s related evaluation of 

this amendment is contained in a Safety 

Evaluation dated March 26, 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: 

January 18, 2007. 

Brief description of amendment: The 

amendment revised the description of 

the control rod assemblies in Grand Gulf 

Nuclear Station, Unit 1, Technical 

Specification 4.2.2, ‘‘Control Rod 

Assemblies,’’ to allow the use of 

hafnium as an additional type of control 

material. 

Date of issuance: March 16, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days from the date of 

issuance. 

Amendment No: 174. 

Facility Operating License No. NPF–

29: The amendment revises the Facility 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: February 13, 2007 (72 FR 

6782). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

March 16, 2007. 

No significant hazards consideration 

comments received: No. 

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: 

September 26, 2006. 

Brief description of amendment: The 

amendment deleted reference to the 

containment fan cooler condensate flow 

switch from Technical Specification 

3.4.5.1, ‘‘Reactor Coolant System 

Leakage—Leakage Detection 

Instrumentation,’’ and modified or 

deleted associated actions. The Nuclear 

Regulatory Commission staff had 

determined that the remaining leak 

detection methods provided adequate 

means for detecting, and to the extent 

practical, identifying the location of the 

source of potential reactor coolant 

leakage. 

Date of issuance: March 19, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 60 

days from the date of issuance. 

Amendment No.: 212. 

Facility Operating License No. NPF–

38: The amendment revised the 

Operating License and the Technical 

Specifications. 

Date of initial notice in Federal 

Register: February 13, 2007 (72 FR 

6782). The Commission’s related 

evaluation of the amendment is 

contained in a Safety Evaluation dated 

March 19, 2007. 

No significant hazards consideration 

comments received: No. 

Indiana Michigan Power Company, 

Docket Nos. 50–315 and 50–316, Donald 

C. Cook Nuclear Plant, Units 1 and 2, 

Berrien County, Michigan 

Date of application for amendments: 

May 26, 2006, as supplemented on 

December 26, 2006, and March 14, 2007. 

Brief description of amendments: The 

amendments revised the existing steam 


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generator (SG) tube surveillance 

program. The changes are modeled after 

Technical Specifications Task Force 

(TSTF) traveler TSTF–449, Revision 4, 

‘‘Steam Generator Tube Integrity,’’ and 

the model safety evaluation prepared by 

the Nuclear Regulatory Commission 

staff and published in the Federal 

Register on March 2, 2005 (70 FR 

10298). In this regard, the scope of the 

amendments includes changes to the 

definition of leakage, changes to the 

primary-to-secondary leakage 

requirements, changes to the SG tube 

surveillance program (SG tube 

integrity), and changes to the SG 

reporting requirements. 

Date of issuance: March 14, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment Nos.: 298 and 279. 

Facility Operating License Nos. DPR–

58 and DPR–74: Amendments revise the 

Technical Specifications. 

Date of initial notice in Federal 

Register: July 5, 2006 (71 FR 38183). 

The supplemental letters 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. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 14, 2007. 

No significant hazards consideration 

comments received: No. 

Nuclear Management Company, LLC, 

Docket No. 50–263, Monticello Nuclear 

Generating Plant, Wright County, 

Minnesota 

Date of application for amendment: 

March 7, 2006, as supplemented by 

letters dated May 30, September 7, 

December 15, 2006, and January 2, 2007. 

Brief description of amendment: The 

amendment revised Section 4.3, ‘‘Fuel 

Storage,’’ of the Monticello Nuclear 

Generating Plant, technical 

specifications to allow for installation of 

an additional temporary 8x8 (64-cell) 

high-density spent fuel storage rack in 

the spent fuel pool to maintain full core 

off-load capability. 

Date of issuance: March 9, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 60 days. 

Amendment No.: 150. 

Facility Operating License No. DPR–

22. Amendment revised the Technical 

Specifications. 

Date of initial notice in Federal 

Register: April 3, 2006 (71 FR 16599). 

The supplemental letters contained 

clarifying information and did not 

change the initial no significant hazards 

consideration determination, and did 

not expand the scope of the original 

Federal Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 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: 

March 23, 2006, as supplemented on 

December 19, 2006. 

Brief description of amendments: The 

amendments revise Technical 

Specification (TS) 3.3.4, ‘‘Loss of Power 

(LOP) Diesel Generator (DG) Start and 

Load Sequence Instrumentation,’’ and 

surveillance requirement 3.3.4.3.b to 

modify the TS title and correct 

nonconservatisms in the allowable 

values for the degraded voltage time 

delay. 

Date of issuance: March 21, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 45 days. 

Amendment Nos.: 225 & 231. 

Renewed Facility Operating License 

Nos. DPR–24 and DPR–27: Amendments 

revised the Technical Specifications and 

License. 

Date of initial notice in Federal 

Register: April 25, 2006 (71 FR 23958). 

The December 19, 2006, supplement, 

contained clarifying information and 

did not change the staff’s initial 

proposed finding of no significant 

hazards consideration. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 21, 2007. 

No significant hazards consideration 

comments received: No. 

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 application for amendments: 

February 16, 2006, supplemented by 

letters dated July 21, and December 27, 

2006. 

Brief description of amendments: The 

amendments consist of changes to the 

Technical Specifications (TSs) related to 

steam generator tube integrity. The 

amendments are modeled after the U.S. 

Nuclear Regulatory Commission 

approved Technical Specification Task 

Force (TSTF) Standard Technical 

Specification Change Traveler, TSTF–

449, ‘‘Steam Generator Tube Integrity,’’

Revision 4 (ML0510902003). 

Date of issuance: March 20, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days. 

Amendment Nos.: 177 and 167. 

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 18376) 

The supplemental letters contained 

clarifying information and did not 

change the initial no significant hazards 

consideration determination, and did 

not expand the scope of the original 

Federal Register notice. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 20, 2007. 

No significant hazards consideration 

comments received: No. 

Pacific Gas and Electric Company, 

Docket Nos. 50–275 and 50–323, Diablo 

Canyon Nuclear Power Plant, Unit Nos. 

1 and 2, San Luis Obispo County, 

California 

Date of application for amendments: 

May 30, 2006, as supplemented by 

letters dated November 22, 2006, and 

January 11, 2007. 

Brief description of amendments: The 

amendments revised the existing steam 

generator (SG) tube surveillance 

program. The changes were modeled 

after Technical Specification Task Force 

(TSTF) traveler TSTF–449, Revision 4, 

‘‘Steam Generator Tube Integrity,’’ and 

the model safety evaluation prepared by 

the U.S. Nuclear Regulatory 

Commission and published in the 

Federal Register on March 2, 2005 (70 

FR 10298). The scope of the application 

included changes to the definition of 

leakage, changes to the primary-tosecondary leakage requirements, 

changes to the SG tube surveillance 

program (SG tube integrity), changes to 

the SG reporting requirements, and 

associated changes to the Technical 

Specification Bases. 

Date of issuance: March 21, 2007. 

Effective date: As of its date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: Unit 1—194; Unit 

2—195. 

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments 

revised the Facility Operating Licenses 

and Technical Specifications. 

Date of initial notice in Federal 

Register: July 18, 2006 (71 FR 40751). 

The supplemental letters dated 

November 22, 2006, and January 11, 

2007, provided additional information 

that clarified the application, did not 

expand the scope of the application as 


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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 March 21, 2007. 

No significant hazards consideration 

comments received: No. 

Pacific Gas and Electric Company, 

Docket Nos. 50–275 and 50–323, Diablo 

Canyon Nuclear Power Plant, Unit Nos. 

1 and 2, San Luis Obispo County, 

California 

Date of application for amendments: 

December 14, 2006. 

Brief description of amendments: The 

amendments deleted Section 2.G of 

Facility Operating License Nos. DPR–80 

and DPR–82, which require reporting of 

violations of the requirements of 

Sections 2.C, 2.E, and 2.F of the 

operating license. This operating license 

improvement was made available by the 

U.S. Nuclear Regulatory Commission on 

November 4, 2005, as part of the 

consolidated line item improvement 

process (CLIIP). 

Date of issuance: March 19, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 90 days from the date of 

issuance. 

Amendment Nos.: Unit 1–193; Unit 

2–194. 

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments 

revised the Facility Operating Licenses. 

Date of initial notice in Federal 

Register: January 3, 2007 (72 FR 154). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 19, 2007. 

No significant hazards consideration 

comments received: No. 

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: 

November 18, 2005, as supplemented on 

November 29, 2006, December 1, 2006, 

December 15, 2006, January 9, 2007, and 

March 12, 2007 (PLA–6168 and PLA–

6169). 

Brief description of amendments: The 

amendments change the SSES 1 and 2 

Technical Specifications (TSs) to 

implement the Average Power Range 

Monitor/Rod Block Monitor/TSs/ 

Maximum Load Line Limit Analysis by 

revising TS 1.1, ‘‘Definitions,’’ TS 5.6.5, 

‘‘Core Operating Limits Report,’’ and the 

surveillance requirement sections of TS 

3.3.1.1, ‘‘Reactor Protection System 

Instrumentation,’’ and TS 3.3.2.1, 

‘‘Control Rod Block Instrumentation.’’

The amendments also delete TS 3.2.4, 

‘‘Average Power Range Monitor Gain 

and Setpoints,’’ and its associated 

references in the TSs. Additionally, the 

amendments change the method of 

evaluation for the postulated 

recirculation line break in the reactor 

pressure vessel shield annulus region. 

Date of issuance: March 23, 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 for Unit 1 

and prior to the startup following the 

SSES 2 spring 2007 13th refueling 

outage for Unit 2. 

Amendment Nos.: 242 and 220. 

Facility Operating License Nos. NPF–

14 and NPF–22: The amendments 

revised the TSs and the License. 

Date of initial notice in Federal 

Register: February 14, 2006 (71 FR 

7810). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 23, 2007. 

No significant hazards consideration 

comments received: No. 

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: 

September 7, 2006. 

Brief description of amendments: The 

amendments revise the SSES 1 and 2 

Technical Specifications (TSs) Section 

5.5.6, ‘‘Inservice Testing Program,’’ and 

TS 5.5.12, ‘‘Primary Containment 

Leakage Rate Testing Program,’’ to be 

consistent with the requirements of 

Title 10 of the Code of Federal 

Regulations (10 CFR) Section 

50.55a(f)(4) and 10 CFR 50.55a(g)(4), 

respectively. The amendments 

implement TS Task Force (TSTF)–343, 

Revision 1 and TSTF–479, Revision 0. 

Date of issuance: March 19, 2007. 

Effective date: As of the date of 

issuance and to be implemented within 

30 days. 

Amendment Nos.: 241 and 219. 

Facility Operating License Nos. NPF–

14 and NPF–22: The amendments 

revised the License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: December 19, 2006 (71 FR 

75997). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 19, 2007. 

No significant hazards consideration 

comments received: No. 

PPL Susquehanna, LLC, Docket No. 50–

388, Susquehanna Steam Electric 

Station, Unit 2 (SSES 2), Luzerne 

County, Pennsylvania 

Date of application for amendment: 

November 16, 2006, as supplemented on 

February 15, 2007. 

Brief description of amendment: The 

amendment changes the SSES 2 

Technical Specification (TS) Section 

2.1.1.2 by revising the Unit 2 Cycle 14 

Minimum Critical Power Ratio Safety 

Limit for two-loop and single-loop 

operation and the references listed in TS 

5.6.5.b. 

Date of issuance: March 19, 2007. 

Effective date: As of the date of 

issuance and to be implemented within 

30 days. 

Amendment No.: 218. 

Facility Operating License No. NPF–

22: The amendment revised the License 

and Technical Specifications. 

Date of initial notice in Federal 

Register: December 19, 2006 (71 FR 

75998). 

The supplement dated February 15, 

2007, 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 19, 2007. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of application for amendments: 

May 1, 2006. 

Brief description of amendments: The 

amendments eliminate the requirement 

for a power range neutron flux high 

negative rate trip and delete the 

references to this trip in Salem Unit 

Nos. 1 and 2 Technical Specification 

(TS) Table 2.2–1, ‘‘Reactor Trip System 

Instrumentation Trip Setpoints,’’ TS 

Table 3.3–1, ‘‘Reactor Trip System 

Instrumentation,’’ TS Table 3.3–2, 

‘‘Reactor Trip System Instrumentation 

Response Times,’’ and TS Table 4.3–1, 

‘‘Reactor Trip System Instrumentation 

Surveillance Requirements.’’ The 

amendments also incorporate 

administrative and editorial changes to 

correct miscellaneous errors in the TSs 

for Salem Units Nos. 1 and 2. 

Date of issuance: March 19, 2007. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 


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Amendment Nos.: 278 and 261 

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments 

revised the TSs and the License. 

Date of initial notice in Federal 

Register: July 18, 2006 (71 FR 40752). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 19, 2007. 

No significant hazards consideration 

comments received: No. 

PSEG Nuclear LLC, Docket Nos. 50–272 

and 50–311, Salem Nuclear Generating 

Station, Unit Nos. 1 and 2, Salem 

County, New Jersey 

Date of application for amendments: 

August 4, 2006, as supplemented by 

letter dated February 20, 2007. 

Brief description of amendments: The 

amendments allow the use of blind 

flanges for containment isolation in the 

containment purge system supply and 

exhaust lines, and make corresponding 

changes to the Technical Specifications 

(TSs). The amendments also consolidate 

the containment isolation requirements 

by moving the requirements of TS 3/4 

6.1.7, ‘‘Containment Ventilation 

System,’’ to TS 3/4 6.3.1 (TS 3/4 6.3 for 

Unit No. 2), ‘‘Containment Isolation 

Valves.’’

Date of issuance: March 19, 2007. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment Nos.: 277 and 260. 

Facility Operating License Nos. DPR–

70 and DPR–75: The amendments 

revised the License and the TSs. 

PSEG Nuclear LLC, Docket No. 50–272, 

Salem Nuclear Generating Station, Unit 

No. 1, Salem County, New Jersey 

Date of application for amendment: 

January 18, 2007, as supplemented on 

February 23, March 9, and March 22, 

2007. 

Brief description of amendment: The 

amendment approves 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 TS 

changes 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 issuance: March 27, 2007. 

Effective date: As of the date of 

issuance, to be implemented within 60 

days. 

Amendment No.: 279. 

Facility Operating License No. DPR–

70: The amendment revised the TSs and 

the License. 

Date of initial notice in Federal 

Register: January 25, 2007 (72 FR 

3427). 

The letters dated February 23, March 

9, and March 22, 2007, provided 

clarifying information that did not 

change the initial proposed no 

significant hazards consideration 

determination or expand the application 

beyond the scope of the original Federal 

Register notice. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 27, 2007. 

No significant hazards consideration 

comments received: No. 

R.E. Ginna Nuclear Power Plant, LLC, 

Docket No. 50–244, R.E. Ginna Nuclear 

Power Plant, Wayne County, New York 

Date of application for amendment: 

March 28, 2006, as supplemented by 

letter dated October 24, 2006. 

Brief description of amendment: The 

amendment revises Technical 

Specification Surveillance Requirement 

3.5.1.4 to change the method and 

frequency for verifying emergency core 

cooling system accumulator boric acid 

concentration. 

Date of issuance: March 28, 2007. 

Effective date: As of the date of 

issuance to be implemented within 45 

days. 

Amendment No.: 101. 

Renewed Facility Operating License 

No. DPR–18: Amendment revised the 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: April 25, 2006 (71 FR 23960) 

The October 24, 2006, letter provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the staff’s original 

proposed no significant hazards 

consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 28, 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, 2006, October 23, 

2006, and February 16, 2007. 

Brief description of amendments: 

These amendments modified Technical 

Specification (TS) requirements related 

to control room envelope habitability in 

TS 3.7.10, ‘‘Control Room Emergency 

Filtration/Pressurization System 

(CREFS)’’ and TS Section 5.5, 

‘‘Administrative Controls—Programs 

and Manuals.’’

Date of issuance: March 26, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: 136/136. 

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: November 8, 2005 (70 FR 

67754). The supplemental letters dated 

September 18 and October 23, 2006, and 

February 16, 2007, 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 March 26, 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: March 

31, 2006. 

Brief description of amendments: The 

amendments revised Technical 

Specification 5.0 entitled, 

‘‘ADMINISTRATIVE CONTROLS.’’

Specifically, the change deleted the Vice 

President, Nuclear Operations, as an 

alternative to the Plant Manager for 

certain functions. 

Date of Issuance: March 20, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance. 

Amendment Nos.: Unit 1–134; Unit 

2–134. 

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: September 12, 2006 (71 FR 

53722). 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated March 20, 2007. 

No significant hazards consideration 

comments received: No. 


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TXU Generation Company LP, Docket 

Nos. 50–445 and 50–446, Comanche 

Peak Steam Electric Station (CPSES), 

Unit Nos. 1 and 2, Somervell County, 

Texas 

Date of amendment request: February 

21, 2006, as supplemented by letter 

dated March 19, 2007. 

Brief description of amendments: The 

amendments revise TS 5.6.5 entitled, 

‘‘Core Operating Limits Report (COLR),’’

by adding two reports providing Lossof-Coolant Accident (LOCA) and nonLOCA analysis methodologies for 

CPSES Unit 1. 

Date of issuance: March 26, 2007. 

Effective date: As of the date of 

issuance and shall be implemented 

within 120 days from the date of 

issuance, but no later than the entry into 

Mode 5 in the restart of Unit 1 from its 

spring 2007 refueling outage. 

Amendment Nos.: 135/135. 

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 supplemental letter dated March 

19, 2007, 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 March 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 25, 2006, as supplemented by letter 

dated March 12, 2007. 

Brief description of amendment: The 

amendment revised Technical 

Specifications 3.1.7, ‘‘Rod Position 

Indication,’’ 3.2.1, ‘‘Heat Flux Hot 

Channel Factor (FQ(Z)) (FQ

Methodology),’’ 3.2.4, ‘‘Quadrant Power 

Tilt Ratio (QPTR),’’ and 3.3.1, ‘‘Reactor 

Trip System (RTS) Instrumentation,’’ to 

allow use of the Westinghouse 

proprietary computer code, the Best 

Estimate Analyzer for Core Operations—

Nuclear (BEACON). Certain required 

actions, for when a limiting condition 

for operation is not met, and certain 

surveillance requirements are being 

changed to refer to power distribution 

measurements or measurement 

information of the core. 

Date of issuance: March 21, 2007. 

Effective date: As of its date of 

issuance and shall be implemented 

before entry into Mode 2 in the plant 

restart from the refueling outage 

scheduled for the spring of 2007. This 

includes the incorporation of the 

identified changes to the Final Safety 

Analysis Report (FSAR) in Attachment 

6 of the licensee’s application dated 

May 25, 2006, into the FSAR. 

Amendment No.: 182. 

Facility Operating License No. NPF–

30: The amendment revised the 

Operating License and Technical 

Specifications. 

Date of initial notice in Federal 

Register: July 18, 2006 (71 FR 40756) 

The supplemental letter dated March 

12, 2007, 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 published 

in the Federal Register on July 18, 2006. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated March 21, 2007. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 3rd day 

of April 2007. 

For the Nuclear Regulatory Commission. 

Catherine Haney, 

Director, Division of Operating Reactor 

Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. E7–6632 Filed 4–9–07; 8:45 am] 

BILLING CODE 7590–01–P 

OVERSEAS PRIVATE INVESTMENT 

CORPORATION 

Submission for OMB Review; 

Comment Request 

AGENCY: Overseas Private Investment 

Corporation (OPIC) 

ACTION: Request for comments. 

SUMMARY: Under the provision of the 

Paperwork Reduction Act (44 U.S.C. 

Chapter 35), agencies are required to 

publish a Notice in the Federal Register 

notifying the public that Agency is 

preparing an information collection 

request for OMB review and approval 

and to request public review and 

comment on the submission. 

Comments are being solicited on the 

need for the information, its practical 

utility, the accuracy of the Agency’s 

burden estimate, and on ways to 

minimize the reporting burden, 

including automated collection 

techniques and uses of other forms of 

technology. The proposed form under 

review is summarized below. 

DATES: Comments must be received 

within 30 calendar days of this notice. 

ADDRESSES: Copies of the subject form 

and the request for review prepared for 

submission to OMB may be obtained 

from the Agency submitting officer. 

Comments on the form should be 

submitted to the Agency Submitting 

Officer. 

FOR FURTHER INFORMATION CONTACT: 

OPIC Agency Submitting Officer: Essie 

Bryant, Record Manager, Overseas 

Private Investment Corporation, 1100 

New York Avenue, NW., Washington, 

DC 20527; 202–336–8563. 

Summary Form Under Review 

Type of Request: Revised form. 

Title: OPIC Self-Monitoring 

Questionnaire. 

Form Number: OPIC–162. 

Frequency of Use: Annually for 

duration of project. 

Type of Respondents: Business or 

other institution (except farms); 

individuals. 

Standard Industrial Classification 

Codes: All. 

Description of Affected Public: U.S. 

companies or citizens investing 

overseas. 

Reporting Hours: 6.5 hours per 

project. 

Number of Responses: 350 per year. 

Federal Cost: $35,000. 

Authority for Information Collection: 

Sections 231, 234(a), 239(d), and 240A 

of the Foreign Assistance Act of 1961, 

as amended. 

Abstract (Needs and Uses): The 

questionnaire is completed by OPICassisted investors annually. The 

questionnaire allows OPIC’s assessment 

of effects of OPIC-assisted projects on 

the U.S. economy and employment, as 

well as on the environment and 

economic development abroad. 

Dated: April 5, 2007. 

John P. Crowley, III, 

Senior Administrative Counsel, Department 

of Legal Affairs. 

[FR Doc. 07–1771 Filed 4–9–07; 8:45 am] 

BILLING CODE 3210–01–M 

PENSION BENEFIT GUARANTY 

CORPORATION 

Approval of Exemption From the Bond/ 

Escrow Requirement Relating to the 

Sale of Assets by an Employer Who 

Contributes to a Multiemployer Plan; 

Washington Nationals Baseball Club, 

LLC 

AGENCY: Pension Benefit Guaranty 

Corporation. 


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